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1

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...

2

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.

3

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

4

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

5

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

6

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

7

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...

8

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

9

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

10

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

11

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

12

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

13

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":""}]}

14

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

15

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

16

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

17

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

18

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

19

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

20

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

Note: This page contains sample records for the topic "naph tha boiling" 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

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

22

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

23

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

24

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

25

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

26

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

27

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

28

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

29

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

30

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

31

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

32

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

33

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

34

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

35

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

36

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

37

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

38

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

39

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

40

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

Note: This page contains sample records for the topic "naph tha boiling" 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

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

42

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

43

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.

44

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

45

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

46

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

47

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.

48

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

49

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

50

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

51

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

52

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

53

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

54

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

55

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":""}]}

56

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

57

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

58

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

59

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

60

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

Note: This page contains sample records for the topic "naph tha boiling" 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

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

62

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

63

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

64

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

65

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

66

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

67

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

68

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

69

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

70

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

71

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

72

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

73

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

74

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

75

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

76

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

77

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

78

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

79

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

80

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

Note: This page contains sample records for the topic "naph tha boiling" 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

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

82

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

83

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

84

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

85

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

86

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

87

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

88

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

89

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

90

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

91

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.

92

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.

93

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

94

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

95

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

96

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

97

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

98

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

99

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

100

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

Note: This page contains sample records for the topic "naph tha boiling" 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

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

102

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

103

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

104

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

105

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

106

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

107

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

108

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

109

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

110

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

111

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

112

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

113

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

114

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

115

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

116

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

117

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

118

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

119

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

120

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

Note: This page contains sample records for the topic "naph tha boiling" 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

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

122

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

123

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

124

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

125

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

126

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

127

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

128

National Aeronautics and Space Administration thaMthereaeWpacS  

E-Print Network (OSTI)

data. The problems were designed to be `one-pagers' with a Teacher's Guide and Answer Key as a second system. In 30 minutes or less, they generate enough energy to power Earth's electrical systems damage and electrical power outages, so scientists try to develop means for predicting when

129

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

130

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

131

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

132

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

133

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

134

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

135

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

136

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

137

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

138

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

139

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

140

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

Note: This page contains sample records for the topic "naph tha boiling" from the National Library of EnergyBeta (NLEBeta).
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141

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

142

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

143

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

144

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

145

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

146

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

147

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

148

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

149

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

150

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

151

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

152

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

153

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

154

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

155

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

156

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

157

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

158

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

159

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

160

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

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161

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

162

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

163

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

164

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

165

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

166

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

167

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

168

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

169

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

170

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

171

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

172

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

173

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

174

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

175

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

176

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

177

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

178

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

179

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

180

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

Note: This page contains sample records for the topic "naph tha boiling" from the National Library of EnergyBeta (NLEBeta).
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181

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

182

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

183

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

184

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

185

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

186

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

187

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

188

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

189

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

190

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

191

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

192

This Opinion was filed under seal on February 26, 2010. The court requested tha  

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

This Opinion was filed under seal on February 26, 2010. The court requested that if This Opinion was filed under seal on February 26, 2010. The court requested that if 1 either party believed that the February 26, 2010 Opinion contained protected material that should be redacted before publication, that party shall, by motion filed on or before March 1, 2010, request that such protected material be redacted. The court has received no motions from either party requesting that the February 26, 2010 Opinion be redacted. The court therefore publishes the February 26, 2010 Opinion in its entirety. In the United States Court of Federal Claims No. 09-864 C (E-Filed: February 26, 2010, Under Seal) (Refiled: March 2, 2010) 1 ) Bid Protest; Statutory Interpretation; Small Business Administration; Priority of Historically Underutilized Business Zone Program over 8(a)

193

The Ideas ThaT shaped a CenTury and  

E-Print Network (OSTI)

, Maximo, Smarter Planet, Global Business Services, World Community Grid, On Demand Community, Many Eyes or transmission in any form or by any means, electronic, mechanical, photocopying, recording or likewise. ISBN-10 Seeing 258 Mapping 268 Understanding 278 Believing 296 Acting 310 Acknowledgments 328 Notes 329

194

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

195

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

196

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

197

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

Science Conference Proceedings (OSTI)

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

Sterbentz, J.W.

1994-10-01T23:59:59.000Z

198

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

199

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

200

Henn-Lecordier AVS 00 MS ThA 5 111/20/00 Integrating Process Models, Equipment  

E-Print Network (OSTI)

-tool sim. 6 : Tool scheduling method 7 : Sensitivity analysis on selected metric 7 : Run factory simulation published data (TI, CVC...) ­ Using statistical software (ECHIP) Device Interconnects TiN PVD Liner Level L & Degassing Load Lock Robot move times Lot Process Time (Makespan) Output Lot size Tool configuration

Rubloff, Gary W.

Note: This page contains sample records for the topic "naph tha boiling" from the National Library of EnergyBeta (NLEBeta).
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to obtain the most current and comprehensive results.


201

ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3. 06. 6B - transient film boiling in upflow. [PWR  

SciTech Connect

Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers.

Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

1982-05-01T23:59:59.000Z

202

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

203

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

204

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

205

EPRI Boiling Water Reactor Mitigation Performance Summary  

Science Conference Proceedings (OSTI)

This report summarizes the intergranular stress corrosion cracking (IGSCC) mitigation performance of 44 BWRs with or without noble metal chemical addition or On-Line NobleChem. Results are categorized by chemistry regime and include data from the most recently completed and current operating cycles. BWRs continue to strive for high hydrogen water chemistry (HWC) availability for IGSCC mitigation, and most plants achieve an overall mitigation performance indicator in the green (excellent) or white (satisf...

2010-03-23T23:59:59.000Z

206

Thermophysical Properties and Pool Boiling Characteristics of ...  

Science Conference Proceedings (OSTI)

... Vapor Oscillatory Flow and Heat Transfer in an Oscillating Heat Pipe,” ASME J ... [33] Kandlikar, SG, Shoji, M., and Dhir, VK, 1999, Handbook of Phase ...

2013-08-12T23:59:59.000Z

207

Subcooled Boiling Data from Rod Bundles  

Science Conference Proceedings (OSTI)

Currently, steaming rate predictions are being made by the industry in thermal hydraulic (TH) codes to predict crud thickness and perform axial offset anomaly (AOA) risk assessments for pressurized water reactors (PWRs). These TH codes use single- and two-phase heat transfer correlations that have not been validated with rod bundle data under prototypical conditions. There is a need to verify that these heat transfer correlations also can predict steaming rates that are applicable for rod bundle geometri...

2002-09-19T23:59:59.000Z

208

Sources of Radioiodine at Boiling Water Reactors  

Science Conference Proceedings (OSTI)

This report determines specific components in operating BWRs that have a potential for being emission sources of radioactive iodine. Conclusions of this study indicate the following: most radioiodine emanating from plants is from a few major areas; in most cases these releases are locally treatable; the interaction with surfaces is an important phenomenon in the behavior of the iodine; and the chemical form in the plant varies according to the circumstances of release. The models developed provide an imp...

1977-12-01T23:59:59.000Z

209

The prediction of low quality boiling voids  

E-Print Network (OSTI)

Slug flow theory is used to predict the density in heated channels of various shapes. In order to make this calculation possible, measurements are made of the bubble rise velocity in annuli, tube bundles, and channels. It ...

Griffith, P.

1963-01-01T23:59:59.000Z

210

Boiling water reactor uranium utilization improvement potential  

Science Conference Proceedings (OSTI)

This report documents the results of design and operational simulation studies to assess the potential for reduction of BWR uranium requirements. The impact of the improvements on separative work requirements and other fuel cycle requirements also were evaluated. The emphasis was on analysis of the improvement potential for once-through cycles, although plutonium recycle also was evaluated. The improvement potential was analyzed for several design alternatives including axial and radial natural uranium blankets, low-leakage refueling patterns, initial core enrichment distribution optimization, reinsert of initial core discharge fuel, preplanned end-of-cycle power coastdown and feedwater temperature reduction, increased discharge burnup, high enrichment discharge fuel rod reassembly and reinsert, lattice and fuel bundle design optimization, coolant density spectral shift with flow control, reduced burnable absorber residual, boric acid for cold shutdown, six-month subcycle refueling, and applications of a once-through thorium cycle design and plutonium recycle.

Wei, P.; Crowther, R.L.; Fennern, L.E.; Savoia, P.J.; Specker, S.R.; Tilley, R.M.; Townsend, D.B.; Wolters, R.A.

1980-06-01T23:59:59.000Z

211

Experimental Boiling Water Reactor decontamination and decommissioning project  

SciTech Connect

The author begins by discussing the problems encountered during decontamination and decommissioning. Next, he discusses waste packaging and recycling. His last topic of lessons learned is subdivided into prevention and early detection, recovery issues, management issues, and noteworthy practices.

Fellhauer, C. [Argonne National Lab., IL (United States). Technology Development Div.

1995-08-01T23:59:59.000Z

212

EVALUATION OF FORCED CONVECTION NUCLEATE BOILING DETECTION BY ACOUSTIC EMISSION  

E-Print Network (OSTI)

The difference between the electrical power input to theby dividing the electrical input power by the surface Theheating of the power cables, these electrical connections

Wells, R.P.

2010-01-01T23:59:59.000Z

213

Boiling point? The skills gap in US manufacturing  

Science Conference Proceedings (OSTI)

... respondents what they considered to be the most serious skill deficiencies in their current employees, inadequate problem-solving skills topped the ...

2013-07-31T23:59:59.000Z

214

Modeling acid-gas generation from boiling chloride brines  

E-Print Network (OSTI)

of the unsaturated zone at Yucca Mountain, Nevada. J. ofwaste emplacement tunnels at Yucca Mountain, Nevada. J. ofScale Heater Test at Yucca Mountain. International Journal

Zhang, Guoxiang

2010-01-01T23:59:59.000Z

215

Fast photo-switchable surfaces for boiling heat transfer applications  

SciTech Connect

Several milligrams of the ruthenium-centered organometallic complex, ruthenium bis-4,4 Prime -di(thiomethyl)-2,2 Prime -bipyridine, mono-2 -(2-pyridyl)-1,3-oxathiane ([Ru{l_brace}(HS-CH{sub 2}){sub 2}-bpy{r_brace}{sub 2}{l_brace}pox{r_brace}](PF{sub 6}){sub 2}) were synthesized and used to produce a self assembled monolayer film on a gold substrate. X-ray photoelectron spectroscopy analysis of the film detected the presence of bound thiolate, which is an indication of a chemisorbed film. Water contact angle measurements were performed before and after 5 min of visible light irradiation using an ozone-free 1000 W Xe(Hg) arc source with a 425-680 nm long pass mirror. The contact angle changed from 52 Degree-Sign pre-irradiation (hydrophilic state) to 95 Degree-Sign post-irradiation (hydrophobic state).

Hunter, C. N.; Glavin, N. R.; Voevodin, A. A. [Air Force Research Laboratory, Materials and Manufacturing Directorate, 2941 Hobson Way, Wright-Patterson Air Force Base, Ohio 45433 (United States); Turner, D. B.; Check, M. H. [Universal Technology Corporation, 1270 North Fairfield Road, Dayton, Ohio 45532 (United States); Jespersen, M. L.; Borton, P. T. [University of Dayton Research Institute, 300 College Park, Dayton, Ohio 45469 (United States)

2012-11-05T23:59:59.000Z

216

Modeling acid-gas generation from boiling chloride brines  

E-Print Network (OSTI)

first drop" of condensate from superheated geothermal steam.from the steam flowing above the condensate. Incompletecondensate that forms at dew-point temperature through expansion of superheated geothermal steam,

Zhang, Guoxiang

2010-01-01T23:59:59.000Z

217

Scoping Study of Moisture Carryover in Boiling Water Reactors  

Science Conference Proceedings (OSTI)

Several BWRs have recently experienced higher than expected shutdown dose rates in steam-affected components/areas. The dose rate increases appear to be associated with increases in reactor water Co-60 activity and moisture carryover (MCO), particularly in the latter portions of the operating cycle. In addition to mechanical carryover, it has been suggested that volatile impurities such as hydrochloric acid may be transported with the BWR steam and concentrate in condensate on surfaces, such as the low p...

2010-12-21T23:59:59.000Z

218

Simulation of sodium boiling experiments with THERMIT sodium version  

E-Print Network (OSTI)

Natural and forced convection experiments(SBTF and French) are simulated with the sodium version of the thermal-hydraulic computer code THERMIT. Simulation is done for the test secti- -on with the pressure-velocity boundary ...

Huh, Kang Yul

1982-01-01T23:59:59.000Z

219

Visual observation of boiling in high power liquid target  

Science Conference Proceedings (OSTI)

A top pressurized, batch style, 3.15 mL total volume (2.5 mL fill volume) water target with transparent viewing windows was operated on an IBA 18/9 cyclotron at 18 MeV proton energy and beam power up to 1.1 kW. Video recordings documented bubble formation and transport, and blue light from de-excitation of water molecules produced images of proton beam stopping geometry including location of the Bragg peak.

Peeples, J. L.; Stokely, M. H.; Poorman, M. C.; Magerl, M.; Wieland, B. W. [Bruce Technologies Inc., 1939 Evans Rd. Cary, NC 27513 (United States); IBA Molecular, 801 Forestwood Dr. Romeoville, IL 60446 (United States); Bruce Technologies Inc., 1939 Evans Rd. Cary, NC 27513 (United States)

2012-12-19T23:59:59.000Z

220

Apparatus for pumping liquids at or below the boiling point  

SciTech Connect

A pump comprises a housing having an inlet and an outlet. An impeller assembly mounted for rotation within the housing includes a first impeller piece having a first mating surface thereon and a second impeller piece having a second mating surface therein. The second mating surface of the second impeller piece includes at least one groove therein so that at least one flow channel is defined between the groove and the first mating surface of the first impeller piece. A drive system operatively associated with the impeller assembly rotates the impeller assembly within the housing.

Bingham, Dennis N. (Idaho Falls, ID)

2002-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "naph tha boiling" 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

Sodium boiling in LMFBR fuel assemblies. Progress report  

Science Conference Proceedings (OSTI)

Objective is to improve current understanding of sodium voiding behavior under postulated LMFBR accident conditions. Multi-dimensional computer models are being developed under low flow and low power conditions. The following computer codes are being developed and assessed: NATOF-2D, THERMIT-S-6E, and THERMIT-S-4E. The effect of virtual mass on the characteristics and numerical stability in two-phase flows was studied. (DLC)

Not Available

1981-04-30T23:59:59.000Z

222

Enhanced Pool-Boiling Heat Transfer Using Nanostructured ...  

Many devices, appliances, and systems -- such as advanced power electronics with high-power computer chips, high-power lasers and radars, and HVAC ...

223

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

224

Enhanced boiling heat transfer by submerged, vibration induced jets .  

E-Print Network (OSTI)

??In this analysis, the efficacy of cavitation jets for heat transfer enhancement was demonstrated. The cavitation jet was formed from a cluster of cavitation bubbles… (more)

Tillery, Steven W.

2005-01-01T23:59:59.000Z

225

Nanocomposite Coatings to Eliminate Bridging Oxidation on Boiling ...  

Science Conference Proceedings (OSTI)

The Plasma Enhanced Magnetron Sputtering (PEMS) process was utilized to deposit Cr, CrN, CrSiCN, and TiSiCN onto Stellite 6B and Stellite 21 coupons.

226

Modeling acid-gas generation from boiling chloride brines  

E-Print Network (OSTI)

when steam flows through a condenser with incomplete or pooral. [6]). A water-cooled condenser was attached at the topin both the flask and condenser. Towards the end of the

Zhang, Guoxiang

2010-01-01T23:59:59.000Z

227

REFLECTOR CONTROL OF A BOILING-WATER REACTOR  

DOE Patents (OSTI)

A line connecting the reactor with a spent steam condenser contains a valve set to open when the pressure in the reactor exceeds a predetermined value and an orifice on the upstream side of the valve. Another line connects the reflector with this line between the orifice and the valve. An excess steam pressure causes the valve to open and the flow of steam through the line draws water out of the reflector. Provision is also made for adding water to the reflector when the steam pressure drops. (AEC)

Treshow, M.

1962-05-22T23:59:59.000Z

228

Why sequence Great Boiling Spring sediment and water microbial...  

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

organisms are of interest to biofuels researchers. One of the hurdles in commercializing biofuel production is that the treatment process requires the application of high...

229

Boils and Turbulence in a Weakly Stratified Shallow Tidal Sea  

Science Conference Proceedings (OSTI)

Measurements of turbulence are made in a weakly but variably stratified region of tidal straining in the eastern Irish Sea using turbulence sensors profiling vertically through the water column on the Fast Light Yo-yo (FLY) profiler and ...

S. A. Thorpe; J. A. M. Green; J. H. Simpson; T. R. Osborn; W. A. M. Nimmo Smith

2008-08-01T23:59:59.000Z

230

Modeling acid-gas generation from boiling chloride brines  

E-Print Network (OSTI)

distillation of a calcium-chloride-dominant brine was simulateddistillation of a calcium-chloride-dominated brine is then simulated

Zhang, Guoxiang

2010-01-01T23:59:59.000Z

231

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 ...

232

Separate effects of surface roughness, wettability and porosity on boiling heat transfer and critical heat flux and optimization of boiling surfaces  

E-Print Network (OSTI)

The separate effects of surface wettability, porosity, and roughness on critical heat flux (CHF) and heat transfer coefficient (HTC) were examined using carefully-engineered surfaces. All test surfaces were prepared on ...

O'Hanley, Harrison Fagan

2012-01-01T23:59:59.000Z

233

Deliverable D6.1 Requirements and scenarios for  

E-Print Network (OSTI)

2011 Contributing WP : WP6 Author(s) : Pierre Chatel (THA), Antoine Leger (THA), James Lockerbie (CITY.2 Atomic steps and attached requirements defined for scenario 1. Atomic steps and choreographies defined (THA), Antoine Leger (THA) 1.3 Added information from CITY, based on information gathered during

Paris-Sud XI, Université de

234

APPARATUS FOR CONTROL OF A BOILING REACTOR RESPONSIVE TO STEAM DEMAND  

DOE Patents (OSTI)

A method of controlling a fuel-rod-in-tube-type boilingwater reactor having nozzles at the point of water entry into the tube is described. Water is pumped into the nozzles by an auxiliary pump operated by steam from an interstage position of the associated turbine, so that the pumping speed is responsive to turbine demand. (AEC)

Treshow, M.

1963-07-23T23:59:59.000Z

235

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

E-Print Network (OSTI)

the effects of depleted (DU) or enriched uranium (EU) and of 137 Cs on vitamin D3 biosynthetic pathway, Paquet F, Voisin P, Aigueperse J and Souidi M. Effects of acute administration of depleted uranium, Dublineau I, Paquet F, Voisin P, Aigueperse J, Gourmelon P and Souidi M. In vivo effects of depleted uranium

Demazière, Christophe

236

An Analysis on the Characteristics of Boiling Liquid Expanding Vapor Explosion Accidents in Marine Transportation  

Science Conference Proceedings (OSTI)

BLEVE is a kind of disaster that may cause serious consequences in the process of maritime transportation of liquefied petroleum gas, liquefied natural gas. To analyze the accident characteristics of both the external environment and the internal causes ... Keywords: BLEVE, boiler, characteristics analysis, liquefied gas storage tank

Sining Chen; Yinquan Duo; Lijun Wei

2010-01-01T23:59:59.000Z

237

Accident source terms for boiling water reactors with high burnup cores.  

Science Conference Proceedings (OSTI)

The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

2007-11-01T23:59:59.000Z

238

BoileR Tube Life Estimator (BRTLE) Version 1.0  

Science Conference Proceedings (OSTI)

The Boiler Tube Life Estimator (BRTLE) software is a stand-alone Windowsapplication that estimates the remaining life of boiler return bends experiencing fatigue cracking using the methodology in Section 9 of API 579-1/ASME FFS-1, 2007 Edition. The calculation of the crack tip stress intensity and reference stress is based on the Level 2 analysis approach in Part 9 of the standard, but conservative assumptions were used for toughness and other variables to minimize the user input required. Therefore, the...

2011-12-19T23:59:59.000Z

239

Feasibility of breeding in hard spectrum boiling water reactors with oxide and nitride fuels  

E-Print Network (OSTI)

This study assesses the neutronic, thermal-hydraulic, and fuel performance aspects of using nitride fuel in place of oxides in Pu-based high conversion light water reactor designs. Using the higher density nitride fuel ...

Feng, Bo, Ph. D. Massachusetts Institute of Technology

2011-01-01T23:59:59.000Z

240

Dispersed-Flow Film Boiling Heat Transfer Data near Spacer Grids in a Rod Bundle  

Science Conference Proceedings (OSTI)

Technical Paper / Radiation Effects and Their Relationship to Geological Repository / Heat Transfer and Fluid Flow

Graydon L. Yoder; Jr.; David G. Morris; Charles B. Mullins; Larry J. Ott

Note: This page contains sample records for the topic "naph tha boiling" 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

Deep Under the Sea, Boiling Founts of Life Itself Page 1 September 9, 2003  

E-Print Network (OSTI)

program. Its draft report was completed in December 1992 (SEAB December 1992). #12;loading strategy waste management program (SEAB December 1992). Recent initiatives on the part of the DOE suggest

Lovley, Derek

242

A high converter concept for fuel management with blanket fuel assemblies in boiling water reactors  

SciTech Connect

Studies on the natural Uranium saving and waste reduction potential of a multiple-plant BWR system were performed. The BWR High Converter system should enable a multiple recycling of MOX fuel in current BWR plants by introducing blanket fuel assemblies and burning Uranium and MOX fuel separately. The feasibility of Uranium cores with blankets and full-MOX cores with Plutonium qualities as low as 40% were studied. The power concentration due to blanket insertion is manageable with modern fuel and acceptable values for the thermal limits and reactivity coefficients were obtained. While challenges remain, full-MOX cores also complied with the main design criteria. The combination of Uranium and Plutonium burners in appropriate proportions could enable obtaining as much as 40% more energy out of Uranium ore. Moreover, a proper adjustment of blanket average stay and Plutonium qualities could lead to a system with nearly no Plutonium left for final disposal. The achievement of such goals with current light water technology makes the BWR HC concept an attractive option to improve the fuel cycle until Gen-IV designs are mature. (authors)

Martinez-Frances, N.; Timm, W.; Rossbach, D. [AREVA, AREVA NP, Erlangen (Germany)

2012-07-01T23:59:59.000Z

243

Source term attenuation by water in the Mark I boiling water reactor drywell  

Science Conference Proceedings (OSTI)

Mechanistic models of aerosol decontamination by an overlying water pool during core debris/concrete interactions and spray removal of aerosols from a Mark I drywell atmosphere are developed. Eighteen uncertain features of the pool decontamination model and 19 uncertain features of the model for the rate coefficient of spray removal of aerosols are identified. Ranges for values of parameters that characterize these uncertain features of the models are established. Probability density functions for values within these ranges are assigned according to a set of rules. A Monte Carlo uncertainty analysis of the decontamination factor produced by water pools 30 and 50 cm deep and subcooled 0--70 K is performed. An uncertainty analysis for the rate constant of spray removal of aerosols is done for water fluxes of 0.25, 0.01, and 0.001 cm{sup 3} H{sub 2}O/cm{sup 2}-s and decontamination factors of 1.1, 2, 3.3, 10, 100, and 1000.

Powers, D.A. [Sandia National Labs., Albuquerque, NM (United States)

1993-09-01T23:59:59.000Z

244

Method for controlling boiling point distribution of coal liquefaction oil product  

SciTech Connect

The relative ratio of heavy distillate to light distillate produced in a coal liquefaction process is continuously controlled by automatically and continuously controlling the ratio of heavy distillate to light distillate in a liquid solvent used to form the feed slurry to the coal liquefaction zone, and varying the weight ratio of heavy distillate to light distillate in the liquid solvent inversely with respect to the desired weight ratio of heavy distillate to light distillate in the distillate fuel oil product. The concentration of light distillate and heavy distillate in the liquid solvent is controlled by recycling predetermined amounts of light distillate and heavy distillate for admixture with feed coal to the process in accordance with the foregoing relationships.

Anderson, Raymond P. (Overland Park, KS); Schmalzer, David K. (Englewood, CO); Wright, Charles H. (Overland Park, KS)

1982-12-21T23:59:59.000Z

245

DESIGN AND HAZARDS SUMMARY REPORT, BOILING REACTOR EXPERIMENT V (BORAX V)  

SciTech Connect

Design data for BORAX V are presented along with results of hazards evaluation studies. Considcration of the hazards associated with the operation of BORAX V was based on the following conditions: For normal steady-state power and experimental operation, the reactor and plant are adequately shielded and ventilated to allow personnel to be safely stationed in the turbine building and on the main floor of the reactor building. The control building is located one- half mile distant from the reactor building. For special, hazardous experiments, personnel are withdrawn from the reactor area. (M.C.G.)

1961-05-01T23:59:59.000Z

246

Apparatus for draining lower drywell pool water into suppresion pool in boiling water reactor  

DOE Patents (OSTI)

An apparatus which mitigates temperature stratification in the suppression pool water caused by hot water drained into the suppression pool from the lower drywell pool. The outlet of a spillover hole formed in the inner bounding wall of the suppression pool is connected to and in flow communication with one end of piping. The inlet end of the piping is above the water level in the suppression pool. The piping is routed down the vertical downcomer duct and through a hole formed in the thin wall separating the downcomer duct from the suppression pool water. The piping discharge end preferably has an elevation at or near the bottom of the suppression pool and has a location in the horizontal plane which is removed from the point where the piping first emerges on the suppression pool side of the inner bounding wall of the suppression pool. This enables water at the surface of the lower drywell pool to flow into and be discharged at the bottom of the suppression pool.

Gluntz, Douglas M. (San Jose, CA)

1996-01-01T23:59:59.000Z

247

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

248

BOILING WATER REACTOR TRANSIENT INSTABILITY STUDIES OF RINGHALS 1 REACTOR USING TRACE COUPLED WITH PARCS.  

E-Print Network (OSTI)

??Reactor plant design often incorporates data and insight ascertained from computer code simulations of plant dynamics and reactor core behavior. Increasing utilization of data gathered… (more)

Walls, Robert

2009-01-01T23:59:59.000Z

249

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

E-Print Network (OSTI)

small passages of compact heat-exchangers, Int J Therm Sci,and refrigeration and compact heat exchangers [4], chemical

Lu, Ming-Chang

2010-01-01T23:59:59.000Z

250

Heat transfer characteristics of a three-phase volume boiling direct contact heat exchanger  

DOE Green Energy (OSTI)

The advantages of direct contact heat transfer over heat transfer utilizing conventional metallic heat exchangers are listed. The performance characteristics of a three-phase direct contact heat exchanger in near counterflow operation were evaluated using water as the continuous phase fluid and refrigerant 113 as the dispersed phase fluid. Conclusions are drawn from the results having to do with refrigerant injection technique, vessel operating height, mass flow rate of refrigerant, water inlet temperature, operation at pinch point temperature differences below 13 to 20/sup 0/C, and operation with a dispersed phase fluid less dense than water. (MHR)

Blair, C.K.; Boehm, R.F.; Jacobs, H.R.

1976-03-01T23:59:59.000Z

251

The effect of surface conditions on nuceate pool boiling heat transfer to sodium  

E-Print Network (OSTI)

A simplified theoretical model for bubble nucleation stability has been proposed, and an approximate stability criterion has been developed. This criterion contains both fluid and surfqce properties, and it predicts that ...

Marto, P. J.

1965-01-01T23:59:59.000Z

252

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

E-Print Network (OSTI)

Heat Transfer Using Micro/Nano Structures by Ming-Chang Lu AHeat Transfer Using Micro/Nano Structures Copyright 2010 byHeat Transfer Using Micro/Nano Structures by Ming-Chang Lu

Lu, Ming-Chang

2010-01-01T23:59:59.000Z

253

Method for controlling boiling point distribution of coal liquefaction oil product  

DOE Patents (OSTI)

The relative ratio of heavy distillate to light distillate produced in a coal liquefaction process is continuously controlled by automatically and continuously controlling the ratio of heavy distillate to light distillate in a liquid solvent used to form the feed slurry to the coal liquefaction zone, and varying the weight ratio of heavy distillate to light distillate in the liquid solvent inversely with respect to the desired weight ratio of heavy distillate to light distillate in the distillate fuel oil product. The concentration of light distillate and heavy distillate in the liquid solvent is controlled by recycling predetermined amounts of light distillate and heavy distillate for admixture with feed coal to the process in accordance with the foregoing relationships. 3 figs.

Anderson, R.P.; Schmalzer, D.K.; Wright, C.H.

1982-12-21T23:59:59.000Z

254

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

255

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

256

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

257

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

E-Print Network (OSTI)

and condensation processes in heat transfer equipment, 2ndand condensation processes in heat transfer equipment, in,the convection process, the heat transfer coefficient of

Lu, Ming-Chang

2010-01-01T23:59:59.000Z

258

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

E-Print Network (OSTI)

a porous wick between the condenser and the evaporation. Theat the evaporator and condenser sections.. 26 Fig. 3.3a heat pipe Length of the condenser section in a heat pipe

Lu, Ming-Chang

2010-01-01T23:59:59.000Z

259

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

E-Print Network (OSTI)

can be migrated by heat conduction or stored through the53-55] rather than heat conduction (see Section 2.3), andis referred to the lateral heat conduction within the wafer

Lu, Ming-Chang

2010-01-01T23:59:59.000Z

260

May 13, 1998 Gas Frac. Mol.Wt. Density Speci c Ht. Boil. Pt.  

E-Print Network (OSTI)

Argon 30 39.95 1.784 0.125 Butane 8 58.12 2.6 0.389 -0.5 HFC-134a 62 102.0 4.5 0.20 -26.3 Table 1-pressure for every 1 m height. Gas is non- ammable. Butane and HFC-134a must be heated during winter 1 #12;RPC drop across one layer less than 5 mmH2O at 10 cc=min ow rate. 2 #12;(Outside) Ar Butane Scale Thermal

Llope, William J.

Note: This page contains sample records for the topic "naph tha boiling" 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

BWR (boiling-water reactor) radiation control: In-plant demonstration at Vermont Yankee: Final report  

Science Conference Proceedings (OSTI)

Results of the RP1934 program, which was established by EPRI in 1981 to demonstrate the adequacy of BRAC program (RP819) principles for BWR radiation control at Vermont Yankee, are presented. Evaluations were performed of the effectiveness of optimization of purification system performance, control of feedwater dissolved oxygen concentrations, minimization of corrosion product and ionic transport, and improved startup, shutdown, and layup practices. The impact on shutdown radiation levels of these corrective actions was assessed based on extensive primary system radiation survey and component gamma scan data. Implementation of the BRAC recommendations was found to be insufficient to reduce the rate of activity buildup on out-of-core surfaces at Vermont Yankee, and additional corrective actions were found necessary. Specifically, replacement of cobalt-bearing materials in the control rod drive pins and rollers and feedwater regulating valves was pursued as was installation of electropolished 316 stainless steel during a recirculation piping replacement program. Aggressive programs to further reduce copper concentrations in the reactor water by improving condensate demineralizer efficiency and to minimize organic ingress to the power cycle by reducing organic concentrations in recycled radwaste also were undertaken. Evaluations of the impact on activity buildup of several pretreatment processes including prefilming in moist air, preexposure to high temperature water containing zinc, and electropolishing also were performed in a test loop installed in the reactor water cleanup system. A significant beneficial impact of electropolishing was shown to be present for periods up to 6000 hours.

Palino, G.F.; Hobart, R.L.; Sawochka, S.G.

1987-10-01T23:59:59.000Z

262

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

E-Print Network (OSTI)

8) (2006) [9] A. Faghri, Heat pipe science and technology,investigations on micro heat pipes, Int J Energ Res, 31(6-investigation of a high flux heat pipe heat sink, J Electron

Lu, Ming-Chang

2010-01-01T23:59:59.000Z

263

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

264

Pressure drop and heat transfer characteristics of boiling water in sub-hundred micron channel  

SciTech Connect

The current work focuses on the pressure drop, heat transfer and stability in two phase flow in microchannels with hydraulic diameter of less than one hundred microns. Experiments were conducted in smooth microchannels of hydraulic diameter of 45, 65 {mu}m, and a rough microchannel of hydraulic diameter of 70 {mu}m, with deionised water as the working fluid. The local saturation pressure and temperature vary substantially over the length of the channel. In order to correctly predict the local saturation temperature and subsequently the heat transfer characteristics, numerical techniques have been used in conjunction with the conventional two phase pressure drop models. The Lockhart-Martinelli (liquid-laminar, vapour-laminar) model is found to predict the two phase pressure drop data within 20%. The instability in two phase flow is quantified; it is found that microchannels of smaller hydraulic diameter have lesser instabilities as compared to their larger counterparts. The experiments also suggest that surface characteristics strongly affect flow stability in the two phase flow regime. The effect of hydraulic diameter and surface characteristics on the flow characteristics and stability in two phase flow is seldom reported, and is of considerable practical relevance. (author)

Bhide, R.R.; Singh, S.G.; Sridharan, Arunkumar; Duttagupta, S.P.; Agrawal, Amit [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400 076 (India)

2009-09-15T23:59:59.000Z

265

Technology, safety and costs of decommissioning a reference boiling water reactor power station. Appendices. Volume 2  

SciTech Connect

Appendices are presented concerning the evaluations of decommissioning financing alternatives; reference site description; reference BWR facility description; radiation dose rate and concrete surface contamination data; radionuclide inventories; public radiation dose models and calculated maximum annual doses; decommissioning methods; generic decommissioning information; immediate dismantlement details; passive safe storage, continuing care, and deferred dismantlement details; entombment details; demolition and site restoration details; cost estimating bases; public radiological safety assessment details; and details of alternate study bases.

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

1980-06-01T23:59:59.000Z

266

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

E-Print Network (OSTI)

investigated for tube bundles consisting of long tubes (H/D>4) [Grant et al. (1979), Schrage et al. (1988 studies provide valuable knowledge on the two-phase pressure drop in conventional tube bundles. However-phase pressure multipliers, Pressure drops Exit quality, Void fractions Inlet and outlet fluid temperarures

Yanikoglu, Berrin

267

NUMERICAL SIMULATION OF BOILING HEAT TRANSFER Ying He, Shigeo Maruyama and Masahiro Shoji  

E-Print Network (OSTI)

-rich macro-layer immediately adjacent to the heater surface. Historically, macro-layer evaporation model-layer on the heater surface itself. Another representative model was reported by Dhir & Liaw(1989). They deduced an area and time-averaged model from the experimental measurements of void fraction close to the heater

Maruyama, Shigeo

268

Technology for Examination of Boiling Water Reactor Bottom Head Drain Lines  

Science Conference Proceedings (OSTI)

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

2006-11-29T23:59:59.000Z

269

Abstracts EuroDendro 2004 [P] Poster [L] Lecture  

E-Print Network (OSTI)

%. The greatest effect of fertilization by double phosphogypsum dose (10 t/ha) ­ increment rise by 39-47%, compared to the control. Having fertilized with 5 t/ha of phosphogypsum, pine annual radial increment has grown by 11-30%, while with a mixture of phosphogypsum (5 t/ha) and superphosphate (100 kg

270

Identifying the medical practice after total hip arthroplasty using an integrated hybrid approach  

Science Conference Proceedings (OSTI)

A critical option of total hip arthroplasty (THA) is considered only when tried more conservative treatments but continued to have pain, stiffness, or problems with the function of ones hip. THA plays one of major concerns under the waves of the rapid ... Keywords: Expert knowledge, Global discretization, Imbalanced class data, Rough set theory (RST), Total hip arthroplasty (THA)

You-Shyang Chen; Ching-Hsue Cheng

2012-08-01T23:59:59.000Z

271

Control of nitrogen-16 in BWR (boiling water reactor) main steam lines under hydrogen water chemistry conditions: Final report  

Science Conference Proceedings (OSTI)

The primary aim of this work was to attempt to identify methods to limit or control the N-16 main steam increases which occur as a result of plant operation under hydrogen water chemistry. The hydrogen water chemistry test data for 8 plants, N-13 chemistry measurements performed at three plants and N-16 main steam concentration measurements made at five plants were analyzed and correlations established. As a result of this study, potential chemical and physical control methods were identified. The test data compilations for the eight plants are included in this report. 6 figs.

Ruiz, C.P.; Lin, C.C.; Wong, T.L.

1989-07-01T23:59:59.000Z

272

Measurement of near-surface void fraction and macrolayer thickness in boiling water and silica-based nanofluid  

E-Print Network (OSTI)

Nanofluids are engineered fluids that contain a suspension of nanoparticles in a pure substance. Nanoparticles can be any variety of metals, metal oxides, or ceramics. They have been shown to increase heat transfer properties ...

Lerch Andrew (Andrew J.)

2008-01-01T23:59:59.000Z

273

The Quality Monitoring Technology in the Process of the Pulping Papermaking Alkaline Steam Boiling Based on Neural Network  

Science Conference Proceedings (OSTI)

On the status quo that being lack of the testing equipment which gives reliable and direct parameters on measuring the quality of pulp in the cooking process, this article focus on the lignin value soft-measurement technology in the pulp and papermaking ... Keywords: Neural network, Pulp and papermaking, Soft-measurement model

Jianjun Su; Yanmei Meng; Chaolin Chen; Funing Lu; Sijie Yan

2008-09-01T23:59:59.000Z

274

Program on Technology Innovation: Technical Support for GE Economic Simplified Boiling Water Reactor (ESBWR)-Radwaste System Design  

Science Conference Proceedings (OSTI)

EPRI has undertaken a review of advanced nuclear plant (ANP) radioactive waste system designs. This work updates EPRI's Utility Requirements Document (URD) for the design of Advanced Light Water Reactor plants. The goal is to capture the radwaste elements that have led to the dramatic improvement in radioactive waste processing in terms of technology advances and improved operating strategy seen in today's operating U.S. plants. This work will form the basis for radioactive waste processing systems in th...

2006-11-21T23:59:59.000Z

275

An assessment of BWR (boiling water reactor) Mark III containment challenges, failure modes, and potential improvements in performance  

Science Conference Proceedings (OSTI)

This report describes risk-significant challenges posed to Mark III containment systems by severe accidents as identified for Grand Gulf. Design similarities and differences between the Mark III plants that are important to containment performance are summarized. The accident sequences responsible for the challenges and the postulated containment failure modes associated with each challenge are identified and described. Improvements are discussed that have the potential either to prevent or delay containment failure, or to mitigate the offsite consequences of a fission product release. For each of these potential improvements, a qualitative analysis is provided. A limited quantitative risk analysis is provided for selected potential improvements. 21 refs., 5 figs., 46 tabs.

Schroeder, J.A.; Pafford, D.J.; Kelly, D.L.; Jones, K.R.; Dallman, F.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1991-01-01T23:59:59.000Z

276

Engineering development of advanced coal-fired low emission boil systems. Quarterly technical progress report, October 1993--December 1993  

Science Conference Proceedings (OSTI)

The first test run of the Toroidal Vortex Combustor (TVC) was completed on December 6. Riley was unable to witness or set up independent sampling equipment for NO{sub x} and precursor measurement for this run. A second run which we witnessed, but did not sample, was completed December 17. This was conducted almost entirely near SR = 1.0 while Textron investigated temperature-load relationships to address concerns from Run 1. A third run was completed over the December holiday break on Dorchester coal to address concerns Textron had about the Illinois test coal. All subsequent tests will use the Illinois coal. Boiler, firing system design. Elevation drawings were developed for dry wall-fired, conventional U-fired slagging, and TVC fired slagging units. We are investigating the feasibility of modifying a conventional U-fired design for low-NOx operation as an alternative to the TVC. The approach taken to I date for NOx reduction in existing U-fired units is to retrofit with delayed-mixing burners with staging air at various places, similar to the approach with dry fired units. The concept of staged fuel addition or reburning for the U-fired system is being examined as a potential combustion NOx control approach. This concept has high potential due to the high temperature and long residence time available in the stagger. Some field trials with coke oven gas reburn produced very low NOx results. Modeling of this concept was identified as a priority task. The model development will include matching field data for air staging on slagging units to the predictions. Emissions control. Selection of an SO2 control process continues to be a high priority task. Sargent & Lundy completed a cost comparison of several regenerable processes, most of which have NOx control potential as well: Active coke, NOXSO, copper oxide, SNOX, ammonia (for SO only, ammonium sulfate byproduct), and a limestone scrubber for comparison.

Not Available

1993-12-31T23:59:59.000Z

277

Departure from nucleate boiling and pressure drop prediction for tubes containing multiple short-length twisted-tape swirl promoters  

E-Print Network (OSTI)

Previous studies conducted at MIT showed that the power performance of an inverted pressurized water reactor (IPWR) conceptual design, i.e. the coolant and moderator are inverted such that the fuel is the continuous medium ...

Arment, Tyrell W. (Tyrell Wayne), 1988-

2012-01-01T23:59:59.000Z

278

The effect of thermal aging and boiling water reactor environment on Type 316L stainless steel welds  

E-Print Network (OSTI)

The thermal aging and consequent embrittlement of materials are ongoing issues in cast stainless steels and duplex stainless steels. Spinodal decomposition is largely responsible for the well known "475°C" embrittlement ...

Lucas, Timothy R

2011-01-01T23:59:59.000Z

279

An assessment of BWR (boiling water reactor) Mark-II containment challenges, failure modes, and potential improvements in performance  

SciTech Connect

This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs.

Kelly, D.L.; Jones, K.R.; Dallman, R.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA)); Wagner, K.C. (Science Applications International Corp., Albuquerque, NM (USA))

1990-07-01T23:59:59.000Z

280

FUEL CYCLE PROGRAM. A BOILING WATER REACTOR RESEARCH AND DEVELOPMENT PROGRAM. Eleventh Quarterly Progress Report, January-March 1963  

SciTech Connect

Even though VBWR shutdowns were required for location and removal of five failed fuel assemblies (HPD Program), the increase in fuel exposure was good. Fuel exposures wili pass the values at which cold worked stainless steel cladding was failing under the HPD Program. Failure of the 0.005-inch cold worked stainless steel clad fuel rods in assembly 8L was traced to strain cycling fatigue. A study of tapered fuel rods indicates a potential advantage for us of a variable water/fuel ratio along the flow channel. Natural circulation tests in the hydraulic stability loop were conducted over a range of conditions from stable, to oscillatory with exponential decay, to self-sustaining oscillation of constant amplitude, to unstable oscillations with divergent amplitude. The response to impulses in power input shows the effect of the time delay for transporting steam voids up through the riser. The data permit calculation of oscillation frequency, damping coefficient, time lags, and show the magnitude and character of pressure and velocity changes. The data, which have an experimental scatter of plus or minus 10% maximum, show that burnout heat fiux: decreases with increasing flow up to 2 x 10/sup 8/ lb/hr-ft/sup 2/; has a maximum for hydraulic diameter between 0.25 and 0.5 inch; and decreases for pressure increases between 600 to 1400 psi. A correlating equation for the data is given. The data are compared to results of others. Tests of special geometries show that the burnout heat flux: decreases 22 to 50% when the heated rod is within 0.033 inch of the channel wall; is unchanged upstream of a plate-type spacer; decreases 35 to 50% when the rod surface is roughened by sandblasting; is increased 20 to 40% by use of a rough liner. The four-rod test section is operating satisfactorily and 17 critical heat fiux data points are obtained at 1000 psia and flows of 0.5, 1.0, and 1.5 x 10/sup 6/ lb/hr-ft/sup 2/. In each case the critical heat flux occurred at the exit end and on the side of the rod facing the corner of the channel. The evaluation of film trippers (rough liner) on the unheated channel walls indicates considerable promise for increasing the burnout heat flux limit. The theory of operation is that the liquid film on the unheated wail is sheared off and dispersed, thus adding to the liquid film on the heated rod. Measurements with a heater rod bowed so that it is in contact with the channel wall show that the critical heat flux is decreased by a factor of two or more from values with normal clearance. Temperature measurements on the rod, when operating past the critical heat flux, were in the order of magnitude of 1000 deg F for heat fluxes of about 500,000--600,000 Btu/hr-ft/sup 2/. Chemical analyses for radial variations in isotopic composition within a fuel pellet are nearly completed and are compiled for interpretation. (N.W.R.)

Howard, C.L. comp.

1963-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "naph tha boiling" 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

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

282

Optimization of Boiling Water Reactor Fuel Crud Characteristics for Reducing Radiation Fields: Evaluation of BWR Fuel Crud Properties  

Science Conference Proceedings (OSTI)

Fuel crud formation and its properties are the combined result of many factors, including corrosion product input, zinc addition rates, reactor coolant chemistry, and fuel and core design. Crud deposition may impact fuel performance as well as radiation field generation. Many projects have evaluated changes in fuel crud properties resulting from changing reactor coolant chemistry. However, the desired crud properties for both good fuel performance and mitigation of radiation field source term are ...

2013-11-26T23:59:59.000Z

283

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

284

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

E-Print Network (OSTI)

across a tube bundle and determined the void fraction, the frictional pressure drop, and the local heat tubes, Trans. ASME 80 (1949) 276. [16] R. Dowlati, M. Kawaji, A.M.C. Chan, Void fraction and friction frictional multiplier and the void fraction. Diabatic studies mainly focus on investigating the two

Peles, Yoav

285

A model for calculation of RCS pressure during reflux boiling under reduced inventory conditions and its assessment against PKL data  

SciTech Connect

There has been recent interest in the United States concerning the loss of residual heat removal system (RHRS) under reduced coolant inventory conditions for pressurized water reactors. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-HI experimental facility at Siemens-KWU to supply applicable data. Recently, an NRC-sponsored effort has been undertaken at the Idaho-National Engineering Laboratory to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following the long term loss-of-RHRS during reduced inventory operation. The thermal-hydraulic response of a closed reactor coolant system during such a transient is investigated in this report. Some of the specific processes investigated include: reflux condensation in the steam generators, the corresponding pressure increase in the reactor coolant system, and void fraction distributions on the primary side of the system. Mathematical models of these and other physical processes Experiment B4.5.

Palmrose, D.E. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Mandl, R.M. [Siemens AG, Berlin (Germany)

1991-12-31T23:59:59.000Z

286

Development of a model to predict flow oscillations in low-flow sodium boiling. [Loss-of-Piping Integrity accidents  

Science Conference Proceedings (OSTI)

Tests performed in a small scale water loop showed that voiding oscillations, similar to those observed in sodium, were present in water, as well. An analytical model, appropriate for either sodium or water, was developed and used to describe the water flow behavior. The experimental results indicate that water can be successfully employed as a sodium simulant, and further, that the condensation heat transfer coefficient varies significantly during the growth and collapse of vapor slugs during oscillations. It is this variation, combined with the temperature profile of the unheated zone above the heat source, which determines the oscillatory behavior of the system. The analytical program has produced a model which qualitatively does a good job in predicting the flow behavior in the wake experiment. The amplitude discrepancies are attributable to experimental uncertainties and model inadequacies. Several parameters (heat transfer coefficient, unheated zone temperature profile, mixing between hot and cold fluids during oscillations) are set by the user. Criteria for the comparison of water and sodium experiments have been developed.

Levin, A.E.; Griffith, P.

1980-04-01T23:59:59.000Z

287

Energy Efficiency Improvement and Cost Saving Opportunities for Breweries: An ENERGY STAR(R) Guide for Energy and Plant Managers  

E-Print Network (OSTI)

compression filter (mashing) Wort boiling and cooling Vapora Use of compression filter 2 Wort boiling and cooling Vapor

Galitsky, Christina; Martin, Nathan; Worrell, Ernst; Lehman, Bryan

2003-01-01T23:59:59.000Z

288

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

289

The Path of Carbon in Photosynthesis VI.  

E-Print Network (OSTI)

factors which determined the rate at which carbon dioxidefactor in the experiments designed to discover tha complex pro- cess by which carbon dioxide

Calvin, M.

1949-01-01T23:59:59.000Z

290

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

291

Development of computer code models for analysis of subassembly voiding in the LMFBR : interim report of the MIT Sodium Boiling Project, covering work through September 30, 1979  

E-Print Network (OSTI)

The research program discussed in this report was started in FY1979 under the combined sponsorship of the U.S. Department of Energy (DOE), General Electric (GE) and Hanford Engineering Development Laboratory (HEDL). The ...

Finkle, William Dean

1979-01-01T23:59:59.000Z

292

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

293

A comparison of factors impacting on radiation buildup at the Vermont Yankee and Monticello BWRs (boiling-water reactors): Interim report  

SciTech Connect

Design and operating features of the Monticello and Vermont Yankee BWRs were compared in an attempt to explain why shutdown radiation levels at Vermont Yankee were significantly higher than at Monticello. The plants were shown to be similar in many respects, for example, condenser and feedwater system design and materials, condensate treatment system design, feedwater iron and copper concentrations, reactor water piping materials and fabrication techniques, reactor water cleanup system flowrates and equipment type, fuel cycle lengths, and fuel failure history. Differences were noted in core power density, jet pump design, reactor water conductivity, volume of radwaste recycle, and the amount of Stellite bearing materials in the feedwater system. Corrosion films on reactor system decontamination flanges from the two plants also were very different. At Monticello, the film was typical of that observed at other BWRs. The Vermont Yankee film contained significantly higher levels of zinc, chromium, and cobalt. Since reactor water Co-60 concentrations at Monticello were about twice those at Vermont Yankee, the Vermont Yankee corrosion film must exhibit a greater tendency to incorporate Co-60.

Palino, G.F.; Hobart, R.L.; Sawochka, S.G.

1987-03-01T23:59:59.000Z

294

Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports  

Science Conference Proceedings (OSTI)

This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

Not Available

1994-01-15T23:59:59.000Z

295

A model for calculation of RCS pressure during reflux boiling under reduced inventory conditions and its assessment against PKL data. [Reactor Cooling Systems (RCS)  

Science Conference Proceedings (OSTI)

There has been recent interest in the United States concerning the loss of residual heat removal system (RHRS) under reduced coolant inventory conditions for pressurized water reactors. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-HI experimental facility at Siemens-KWU to supply applicable data. Recently, an NRC-sponsored effort has been undertaken at the Idaho-National Engineering Laboratory to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following the long term loss-of-RHRS during reduced inventory operation. The thermal-hydraulic response of a closed reactor coolant system during such a transient is investigated in this report. Some of the specific processes investigated include: reflux condensation in the steam generators, the corresponding pressure increase in the reactor coolant system, and void fraction distributions on the primary side of the system. Mathematical models of these and other physical processes Experiment B4.5.

Palmrose, D.E. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Mandl, R.M. (Siemens AG, Berlin (Germany))

1991-01-01T23:59:59.000Z

296

A Water Hypsometer Utilizing High-Precision Thermocouples  

Science Conference Proceedings (OSTI)

A boiling-point barometer—commonly called hypsometer—has been developed for use on meteorological radiosondes. In this hypsometer, water is heated electrically, and its boiling temperature is measured with a thermocouple. Once the boiling ...

Hans Richner; Jürg Joss; Paul Ruppert

1996-02-01T23:59:59.000Z

297

Documentation of soil conditions at liquefaction and non-liquefaction sites from 1999 Chi-Chi (Taiwan) earthquake  

E-Print Network (OSTI)

with sediment boils and ground subsidence. Lateral spreadinglocations where ground subsidence beneath buildings did andsand boils, building subsidence and/or foundation failure,

2004-01-01T23:59:59.000Z

298

Meeting the Need for Safe Drinking Water in Rural Mexico through Point-of-Use Treatment  

E-Print Network (OSTI)

are boiling and adding chlorine in the form of tablets orto water quality problems. Chlorine pills and bleach requireassociated with boiling, chlorine tablets, and bleach. Two

Lang, Micah; Kaser, Forrest; Reygadas, Fermin; Nelson, Kara; Kammen, Daniel M.

2006-01-01T23:59:59.000Z

299

2004 AQUAINT Challenge Problem  

Science Conference Proceedings (OSTI)

... Ben Barnett, Trina Pitcher, John Calhoun, Eileen Boiling (Admin); Antonio Sanfilippo (Project Manager). Roadmap. Feb 23. ...

2004-10-07T23:59:59.000Z

300

RADIATION RESEARCH 153, 220238 (2000) 0033-7587/00 $5.00  

E-Print Network (OSTI)

): the light water reactor [(LWR) in both pressurized and boil- ing versions]; heavy water (CANDU) reactor

Note: This page contains sample records for the topic "naph tha boiling" 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

NIST Image Gallery: Image Details  

Science Conference Proceedings (OSTI)

... Title: Traces of Nanobubbles Determine Nano-boiling. Description: Ultra-highspeed photographs of microbubbles forming ...

302

1197 South Lumpkin Street Athens Georgia 30602-3603 weddings@georgiacenter.uga.edu 706.542.2654 The minimum charge For a dinner recePTion is $600.00.  

E-Print Network (OSTI)

with Assorted Breads and Condiments Smoked Side of Salmon with Classic Garnishes Boiled Shrimp with Spicy

Arnold, Jonathan

303

Mineralini tr poveikio miko dirvoemiui ir pu radialiajam prieaugiui Akmens cemento gamyklos ... ISSN 02357224. E k o l o g i j a (Vilnius). 2001. Nr. 1  

E-Print Network (OSTI)

A0 I 02 7,2 6,8 100,0 30 6,8 6,5 1020 Fosfogipsas (10 t/ha) Phosphogypsum (10 t/ha) A0 II 25 7,8 7 Fosfogipsas (5 t/ha) Phosphogypsum (5 t/ha) A0 II 25 7,8 7,2 25,5 33 2,3 4,5 130 T 1 520 7,4 7,2 25,5 7,0 2/ha veikliosios A0 II 25 7,8 7,2 30,0 32 2,0 4,9 1080 medþiagos) Mixture of phosphogypsum (5 t

304

The Fission of thorium with Alpha Particles  

E-Print Network (OSTI)

cm 2 FIG. 2 YIELD OF SOME FISSION PRODUCTS FROM THORIUM +-in aalou.latlnl~ I tha fission yield ot tlla 2.3;; dnyci~S~~SIO NASS HO. FIG. I FISSION YIELD SPECTRUM OF THORIUM

Newton, Amos S.

2010-01-01T23:59:59.000Z

305

Bulletin of Tibetology: Volume 21 Number 3 : Full issue  

E-Print Network (OSTI)

were divided into grades each with its special insignia consisting of ornaments and diplomas of different precious substances. J n general the highest was turquoise, followed by gold, 'phra men, silver, brass, and copper \\ LINV 1071); but in THA p...

Namgyal Institute of Tibetology

306

SSRL Accelerator Phycics Home Page  

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

(29047 bytes) ICFA2000t.gif (31362 bytes) Home Page LCLS Accelerator Physics at SSRL The field tha t can be covered by the Accelerator Physics activities at SSRL is limited...

307

Variation in joint fluid composition and its effect on the tribology of replacement joint articulation  

E-Print Network (OSTI)

Polyethylene wear is a significant clinical problem limiting the long-term survival of joint replacement prostheses, particularly in total hip arthroplasty (THA) and total knee arthroplasty (TKA). Although the tribology ...

Mazzucco, Daniel Clarke, 1976-

2003-01-01T23:59:59.000Z

308

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

309

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

310

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 - appendices. Final report  

SciTech Connect

The NRC staff is in need of decommissioning 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) 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 presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clear 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

311

Concentrating solar power | Open Energy Information  

Open Energy Info (EERE)

worldwide. Many power plants today use fossil fuels as a heat source to boil water. The steam from the boiling water spins a large turbine, which drives a generator to produce...

312

Minimizing corrosion in coal liquid distillation  

DOE Patents (OSTI)

In an atmospheric distillation tower of a coal liquefaction process, tower materials corrosion is reduced or eliminated by introduction of boiling point differentiated streams to boiling point differentiated tower regions.

Baumert, Kenneth L. (Emmaus, PA); Sagues, Alberto A. (Lexington, KY); Davis, Burtron H. (Georgetown, KY)

1985-01-01T23:59:59.000Z

313

Micro/Nano-Scale Phase Change Systems for Thermal Management and Solar Energy Conversion Applications  

E-Print Network (OSTI)

Enhancement in Pool Boiling of Nano-fluids,” InternationalLiquid Spreading Due to Nano/Microstructures on the CriticalWicking Action of Micro/Nano Structures on Pool Boiling

Coso, Dusan

2013-01-01T23:59:59.000Z

314

NREL: Learning - Concentrating Solar Power Basics  

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

Concentrating Solar Power Basics Many power plants today use fossil fuels as a heat source to boil water. The steam from the boiling water spins a large turbine, which drives a...

315

Study of chemistry and irradiation effects on nanofluids to be used in light water reactor accident cooling  

E-Print Network (OSTI)

Nanofluids, colloidal dispersions of nanoparticles in a base fluid, have shown enhancements in both pool boiling and flow boiling critical heat flux (CHF) in laboratory tests. The applicability of this aspect to nuclear ...

Lucas, Timothy R

2008-01-01T23:59:59.000Z

316

Christmas in Early America  

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

one with the only spoon, fat meat of the "wild cattle" and, as a special treat, boiled dog. You probably won't have boiled dog but have a Merry Christmas anyway. To return to the...

317

TMS News Article  

Science Conference Proceedings (OSTI)

“In boiling water canning, I immerse the jars in boiling water for a time sufficient to kill microorganisms,” Holm said. “The heat also causes the air in the jar to ...

318

Important Contacts: JoeDePage,PorterHousingCoordinator  

E-Print Network (OSTI)

of the Pilgrim Nuclear Power Station which is a 670 MWe boiling water reactor which started commercial operation

California at Santa Cruz, University of

319

Electric Vehicles Could Benefit from 50% Improvements in Packaging...  

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

surface. These results have significant implications for reducing the overall thermal resistance in power electronics packagesmodules. Immersion boiling combined with...

320

JOURNAL OF RESEARCH of the National Bureau of ...  

Science Conference Proceedings (OSTI)

... colored. To remove the color, several grams of activated cocoanut charcoal were added to the boiling so- lution. The hot ...

2010-11-23T23:59:59.000Z

Note: This page contains sample records for the topic "naph tha boiling" 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

Building and Fire Publications  

Science Conference Proceedings (OSTI)

... The correlation poorly predicted the heat transfer performance of cross-grooved, micro ... an existing correlation for flow boiling pressure ...

322

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

E-Print Network (OSTI)

Water Reactors (PWRs) CANDU Boiling Water Reactors (BWRs)Reactors (EPRs) Advanced CANDU Reactors (ACRs) AP1000

Grigoriadis, Theocharis N

2009-01-01T23:59:59.000Z

323

Creating Markets for Green Biofuels: Measuring and improving environmental performance  

E-Print Network (OSTI)

boils water to drive a steam turbine generator and then usesremaining in the steam exiting the turbine for process Dry-

Turner, Brian T.; Plevin, Richard J.; O'Hare, Michael; Farrell, Alexander E.

2007-01-01T23:59:59.000Z

325

A New Absorption Cycle: The Single-Effect Regenerative Absoprtion Refrigeration Cycle  

E-Print Network (OSTI)

heat source) transfers heat to boiling process B. A multi-processes in the adiabatic regenerator; that is , heat transfer

Dao, K.

2011-01-01T23:59:59.000Z

326

Silver  

Science Conference Proceedings (OSTI)

Table 5   Corrosion resistance of silver in organic compounds...2 Benzaldehyde, pure and aqueous Boiling Benzene, pure Boiling <0.05 2 Benzotrifluoride, pure Boiling <0.05 2 Benzyl chloride, pure 180 355 <0.05 2 -bromoisovaleryl bromide, pure 100 212 <0.05 2 -bromoisovaleryl urea, pure Melting point <0.05 2 Butyl acetate, pure Boiling <0.05 2 Butyl alcohol,...

327

SCALE Newsletter (Spring 2012) 1 NewsletterNumber 44 Spring 2012  

E-Print Network (OSTI)

-pile instrumentation · ATR (INL) · HFIR (ORNL) · Halden Boiling Water Reactor (Norway) · Jules Horowitz reactor (France

328

Building and Fire Publications  

Science Conference Proceedings (OSTI)

... Proceeding. 2008, 1-22 pp, 2008. Keywords: refrigerants; nanolubricants; heat transfer; boiling; air conditioning; chillers; nanofluids Abstract: ...

329

PNNL: Available Technologies: Electronics  

Current Control Technology for Quantum Cascade Laser and Other Applications; Enhanced Pool-Boiling Heat Transfer Using Nanostructured Surfaces;

330

Signal reconstruction by a GA-optimized ensemble of PCA models P. Baraldi1  

E-Print Network (OSTI)

signals collected from a Swedish nuclear boiling water reactor. 1. Introduction During plant operation at a Swedish nuclear Boiling Water Reactor located in Oskarshamn. Some conclusions on the advantages concerning n =84 signals collected at a nuclear Boiling Water Reactor (BWR) located in Oskarshamn, Sweden

331

Platinum  

Science Conference Proceedings (OSTI)

Table 13   Corrosion of platinum in acids...2 Benzoic, all concentrations 130 265 Benzene sulfonic, pure Room <0.05 2 Boric, saturated Boiling <0.05 2 Butyric, all concentrations Boiling <0.05 2 Carbonic, pure 1400 2550 <0.05 2 Chloric, all concentrations Room <0.05 2 Chlorosulfonic, all concentrations Boiling <0.05 2...

332

Coal liquefaction process with increased naphtha yields  

DOE Patents (OSTI)

An improved process for liquefying solid carbonaceous materials wherein the solid carbonaceous material is slurried with a suitable solvent and then subjected to liquefaction at elevated temperature and pressure to produce a normally gaseous product, a normally liquid product and a normally solid product. The normally liquid product is further separated into a naphtha boiling range product, a solvent boiling range product and a vacuum gas-oil boiling range product. At least a portion of the solvent boiling-range product and the vacuum gas-oil boiling range product are then combined and passed to a hydrotreater where the mixture is hydrotreated at relatively severe hydrotreating conditions and the liquid product from the hydrotreater then passed to a catalytic cracker. In the catalytic cracker, the hydrotreater effluent is converted partially to a naphtha boiling range product and to a solvent boiling range product. The naphtha boiling range product is added to the naphtha boiling range product from coal liquefaction to thereby significantly increase the production of naphtha boiling range materials. At least a portion of the solvent boiling range product, on the other hand, is separately hydrogenated and used as solvent for the liquefaction. Use of this material as at least a portion of the solvent significantly reduces the amount of saturated materials in said solvent.

Ryan, Daniel F. (Friendswood, TX)

1986-01-01T23:59:59.000Z

333

Soluble Substances and Evaporation  

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

Soluble Substances and Evaporation Soluble Substances and Evaporation Name: JD Status: student Grade: 9-12 Location: FL Country: New Zealand Date: Winter 2011-2012 Question: Do soluble substances evaporate with the water? Replies: JD, As a general rule, no. If the soluble substance is a solid, then its boiling point is well above that of water, so it cannot possibly boil off. If the substance is a liquid, it may have a boiling point that is below that of water and will boil off at a lower temperature than water. If the boiling point is higher than that of water, than it will boil off after the water has evaporated. Some substances, like ethanol for example, form an "azeotrope" with water. The combination of ethanol and water form a tight intermolecular connection that makes the two substances boil off at the same time.

334

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

335

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

336

Energy Currents  

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

water purification system ideal for villages in developing nations. In many rural areas, water is disinfected by boiling, an inefficient process that is hampered by...

337

X-ray Diffuse Scattering Measurements of Nucleation Dynamics...  

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

high pressure conditions which relax into the liquid-vapor coexistence region of the phase diagram, leading to a rapid phase-explosion-driven boiling process within the...

338

Performance of Charcoal Cookstoves for Haiti, Part 1: Results...  

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

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 at: http:...

339

Glossary - U.S. Energy Information Administration (EIA)  

U.S. Energy Information Administration (EIA)

Catalytic hydrocracking: A refining process that uses hydrogen and catalysts with relatively low temperatures and high pressures for converting middle boiling or ...

340

A. Philip Bray, 1976 | U.S. DOE Office of Science (SC)  

Office of Science (SC) Website

Feeds FeedbackShare Page Reactors: For contributions to the fundamental knowledge of thermal and hydraulic phenomenon associated with boiling water reactors and the use of the...

Note: This page contains sample records for the topic "naph tha boiling" 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

Oxygen And Carbon Isotope Ratios Of Hydrothermal Minerals From...  

Open Energy Info (EERE)

are consistent with deposition during transient boiling or rock-water exchange (fracturing) events. Author(s): N. C. Sturchio, T. E. C. Keith, K. Muehlenbachs Published:...

342

New Player in Electron Field Emitter Technology Makes for ...  

Science Conference Proceedings (OSTI)

... Thermionic sources use an electric current to boil electrons off the surface of a wire filament, similar to the way an incandescent light bulb uses an ...

2013-03-05T23:59:59.000Z

343

Designing Advanced Scanning Probe Microscopy Instruments  

Science Conference Proceedings (OSTI)

... The UHV insert is cooled via He exchange gas to avoid vibrations from the boiling of the liquid He in the main bath of the cryostat. ...

2011-09-22T23:59:59.000Z

344

Measurement of Liquid to Gel Phase Transition Temperature ...  

Science Conference Proceedings (OSTI)

... The agar is a hydrophilic colloid extracted from some red marine algae which are soluble in boiling water, and performed a reversible liquid to gel ...

2006-07-20T23:59:59.000Z

345

Publications Portal  

Science Conference Proceedings (OSTI)

... on R134a Pool Boiling Heat Transfer Topic: High ... Heat exchanger performance is strongly influenced by the ... in units of differential cross-section per ...

2012-09-17T23:59:59.000Z

346

NIST Tech Beat for May 28, 2013  

Science Conference Proceedings (OSTI)

... Creating premium fuel requires a refinery to boil the mixture at precise ... The new MOF, however, could allow refineries to sidestep this problem by ...

2013-05-28T23:59:59.000Z

347

New Filtration Material Could Make Petroleum Refining ...  

Science Conference Proceedings (OSTI)

... Creating premium fuel requires a refinery to boil the mixture at precise ... The new MOF, however, could allow refineries to sidestep this problem by ...

2013-05-28T23:59:59.000Z

348

International Journal of Applied Electromagnetics and Mechanics 15 (2001/2002) 291294 291 Feature extraction techniques for ultrasonic  

E-Print Network (OSTI)

acquired from weld inspection regions of boiling water reactor piping of nuclear power plants that produces mechanical oscillations inside the inspected material, and acquires the reflected ultrasonic wave

Polikar, Robi

349

Lesson 6 - Atoms to Electricity | Department of Energy  

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

the heat to boil the water. This lesson covers Inside the Reactor Heat Pressure Water Fission Control Fuel assemblies Control rods Coolant Pressure vessel Electricity Generation...

350

Page not found | Department of Energy  

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

Download EA-1394: Final Environmental Assessment Authorizing the Puerto Rico Electric Power Authority (PREPA) to allow Public Access to the Boiling Nuclear...

351

Environmental Assessments (EA) | Department of Energy  

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

352

Conflict Between Economic  

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

progress. The fundamentality of the conflict ultimately boils down to laws of thermodynamics. Physicists and other scholars from the physical sciences are urgently needed for...

353

A Review of Hazardous Chemical Species Associated with CO2 Capture from Coal-Fired Power Plants and Their Potential Fate in CO2 Geologic Storage  

E-Print Network (OSTI)

with conventional steam turbine powered electric generation.used to boil water for steam turbine generation as a secondturbine) and Rankine (steam turbine) cycles, as illustrated

Apps, J.A.

2006-01-01T23:59:59.000Z

354

BWRVIP-212: BWR Vessel and Internals Project, Evaluation of Crack Growth Rates of Nickel-Base Alloys 52, 152, and 690 in BWR Enviro nments  

Science Conference Proceedings (OSTI)

This study summarizes available information on stress corrosion cracking (SCC) performance of high-chromium, nickel-base (Ni-base) alloys in boiling water reactor (BWR) environments.

2009-04-14T23:59:59.000Z

355

Cool! Nanoparticle Research Points to Energy Savings  

Science Conference Proceedings (OSTI)

... The double-bubble effect enhances boiling heat transfer and, ultimately, could help to boost the energy efficiency of industrial-sized cooling systems ...

2011-05-02T23:59:59.000Z

356

Building and Fire Publications  

Science Conference Proceedings (OSTI)

... of Concentration and Additives on R123/Paraffinic Material Oil Boiling Heat ... Between Saffir-Simpson Hurricane Scale Wind Speeds and Peak 3-s ...

357

Complex Fluid Analysis with the Advanced Distillation Curve ...  

Science Conference Proceedings (OSTI)

... Inset (B) shows a total sulfur chromatographic peak. ... The main peaks from the low boiling region ... acidic and basic crude oil components detected by ...

2012-10-09T23:59:59.000Z

358

Optimization of Phase Change Heat Transfer in Biporous Media  

E-Print Network (OSTI)

Aspectcs of Boiling Heat Transfer”. PhD Thesis dissertation,Celled Foams”. Numerical Heat Transfer, Vol. 54, issue 1,Dimensional Fluid Flow and Heat Transfer”. Numerical Heat

Reilly, Sean

2013-01-01T23:59:59.000Z

359

Heat Transfer and Pressure Drop During Condensation of Refrigerants in Microchannels .  

E-Print Network (OSTI)

??Two-phase flow, boiling, and condensation in microchannels have received considerable attention in the recent past due to the growing interest in the high heat fluxes… (more)

Agarwal, Akhil

2006-01-01T23:59:59.000Z

360

Building and Fire Publications  

Science Conference Proceedings (OSTI)

... Effect of Refrigerant Oil Additive on R134a and R123 Boiling Heat Transfer Performance. Effect of Refrigerant Oil Additive ...

Note: This page contains sample records for the topic "naph tha boiling" 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

A new method to cleaner and more efficient CO2 capture  

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

has developed a screening method that would use ionic liquids - a special type of molten salt that becomes liquid under the boiling point of water (100 degrees Celsius) -...

362

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...

363

Two Line Subject Title One Line Title  

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

poor thermal properties Electronic device Pool-boiling refrigerant Extended surface Condenser Fan Vapor Liquid Heat Transfer Fluid Design PRIMARY LOOP HEAT EXCHANGER...

364

DOE/ID-Number  

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

hole assembly BOPE blow-out prevention equipment BRC Blue Ribbon Commission BWR boiling water reactor CSH calcium-silicate-hydrate DBD deep borehole disposal DOE Department of...

365

ANOMALOUS EFFECTS OF WATER IN FIREFIGHTING ...  

Science Conference Proceedings (OSTI)

... As shown in Figures 1 - 3 for benzene, xylene, and water, the boiling point of any liquid or mixture of liquids is that temperature at which the vapor ...

2011-10-27T23:59:59.000Z

366

Using the OLS algorithm to build interpretable rule bases: an application to a depollution problem  

E-Print Network (OSTI)

and biogas .. 107 Table IV.1.1 Identification of organotin species by the comparison of the boiling points and biogas

Paris-Sud XI, Université de

367

LNG storage: Safety analysis. Annual report, January-December 1980  

SciTech Connect

Progress is summarized on three projects in the areas of LNG safety: Rollover phenomena; Simultaneous boiling and spreading of cryogenic liquids; Modelling of LNG tank dynamics.

Reid, R.C.; Smith, K.A.; Virk, P.S.

1981-02-01T23:59:59.000Z

368

NIST Working Fluids Nanolubricants Research Project | Department...  

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

develop a fundamental understanding of how nanolubricants, a lubricant with dispersed nano-size particles, can enhance refrigerant and lubricant pool boiling. A previous National...

369

Reaction to Fukushima: 1663 Science and Technology Magazine ...  

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

destruction, including major damage to the Fukushima Daiichi nuclear power plant. The aging plant consisted of six boiling-water reactors powering electrical generators. The...

370

It's Elemental - The Element Darmstadtium  

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

Roentgenium The Element Darmstadtium Click for Isotope Data 110 Ds Darmstadtium 281 Atomic Number: 110 Atomic Weight: 281 Melting Point: Unknown Boiling Point: Unknown...

371

It's Elemental - The Element Berkelium  

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

Californium The Element Berkelium Click for Isotope Data 97 Bk Berkelium 247 Atomic Number: 97 Atomic Weight: 247 Melting Point: 1323 K (1050C or 1922F) Boiling...

372

Meager genetic variability of the human malaria agent Plasmodium vivax  

E-Print Network (OSTI)

collected from Azerbaijan, Thailand, Turkey, Venezuela, and Ethiopia. Three blood samples were ob- tained; AZE, Azerbaijan; THA, Thailand; TUR, Turkey; VEN, Venezuela; PNG, Papua New Guinea; MOZ, Mozambique variability, P. simium, which comes from South America, is more closely related to P. vivax from Azerbaijan

373

Assessment of Research Quality Civil Engineering  

E-Print Network (OSTI)

programmes or cooperate within knowledge centres. These include: · DUCON Centre for Durable Concrete dropout rate. The launch of the FGS will also coincide with measures to reduce both the time spent on Ph sho programm longer tha that there dropout ra the FGS w measures time spen programm rates and

Dekker, Cees

374

Efficient Thermal Energy Distribution in Commercial Final Report  

E-Print Network (OSTI)

cconolllic studics havc shown, dccp rcductions in CO2 cmissions can nc achicvcd with cxisting tcchnologics cconomy or today's c.lrs thaI havc thc salnc pcrrormanl:C. nut lhcy would c(}sl 11()morc to own control for fucl-hound nitrogen would hc Icss costly for hiomass than for coal. Typical co.lls havc

375

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, draft report for comment. Volume 1  

Science Conference Proceedings (OSTI)

On June 27, 1988, the U.S. Nuclear Regulatory Commission (NRC) published in the Federal Register (53 FR 24018) the final rule for the General Requirements for Decommissioning Nuclear Facilities. With the issuance of the final rule, owners and operators of licensed nuclear power plants are required to prepare, and submit to the NRC for review, decommissioning plans and cost estimates. 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), including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB alternatives, which 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. 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. Costs for labor, transport, and disposal activities are given in 1993 dollars. Sensitivities 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 are also examined.

Smith, R.I.; Bierschbach, M.C.; Konzek, G.J. [Pacific Northwest Lab., Richland, WA (United States)] [and others

1994-09-01T23:59:59.000Z

376

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: Appendices, draft report for comment. Volume 2  

SciTech Connect

On June 27, 1988, the U.S. Nuclear Regulatory Commission (NRC) published in the Federal Register (53 FR 24018) the final rule for the General Requirements for Decommissioning Nuclear Facilities. With the issuance of the final rule, owners and operators of licensed nuclear power plants are required to prepare, and submit to the NRC for review, decommissioning plans and cost estimates. 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 WNP-2, 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, which 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. 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. Costs for labor, materials, transport, and disposal activities are given in 1993 dollars. Sensitivities 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 are also examined.

Smith, R.I.; Bierschbach, M.C.; Konzek, G.J. [Pacific Northwest Lab., Richland, WA (United States)] [and others

1994-09-01T23:59:59.000Z

377

BWRVIP-244: BWR Vessel and Internals Project, Nondestructive Evaluation Development 2010  

Science Conference Proceedings (OSTI)

This report provides 2010 results of the nondestructive evaluation NDE development task of the Boiling Water Reactor Vessel and Internals Project BWRVIP Inspection Focus Group. The scope of activity includes applications of various NDE techniques to boiling water reactor vessels and vessel internals components.

2010-12-23T23:59:59.000Z

378

Photocleavage Agarose Gel Procedure A.Purify pUC 18 plasmid DNA  

E-Print Network (OSTI)

is placed in front of the lamp sample holder. 2. Prepare a grid laying out the appropriate components on stir plate with stir and heat both set to 6. Allow to heat and stir until it comes to a vigorous boil. Let boil a minute and remove from the heat. Gel mixture at this point should have turned from cloudy

Turro, Claudia

379

Lesson 6- Atoms to Electricity  

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

Most power plants make electricity by boiling water to make steam that turns a turbine. A nuclear power plant works this way, too. At a nuclear power plant, splitting atoms produce the heat to boil the water. This lesson covers inside the reactor, fission control and electricity generation.

380

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

Note: This page contains sample records for the topic "naph tha boiling" 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

THERMAL DESIGN METHODOLOGY FOR LOW FLOW RATE SINGLE-PHASE AND TWO-PHASE MICRO-CHANNEL HEAT SINKS  

E-Print Network (OSTI)

in Engine Cooling Systems,'' Experimental Heat Transfer, Fluid Mechanics, and Thermodynamics 1997, June 1997, ``A Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow,'' Ind. Eng. Chem the engines in automotive applications. Heat is transferred essentially under subcooled flow boiling

Qu, Weilin

382

Effect of Narrow Cut Oil Shale Distillates on HCCI Engine Performance  

Science Conference Proceedings (OSTI)

In this investigation, oil shale crude obtained from the Green River Formation in Colorado using Paraho Direct retorting was mildly hydrotreated and distilled to produce 7 narrow boiling point fuels of equal volumes. The resulting derived cetane numbers ranged between 38.3 and 43.9. Fuel chemistry and bulk properties strongly correlated with boiling point.

Eaton, Scott J [ORNL; Bunting, Bruce G [ORNL; Lewis Sr, Samuel Arthur [ORNL; Fairbridge, Craig [National Centre for Upgrading Technology, Canada

2009-01-01T23:59:59.000Z

383

g Ris* Report No. 289 g Danish Atomic Energy Commission  

E-Print Network (OSTI)

of fundamental theory and data. Fundamental data are in this connection data such as nuclear cross sections of Prediction of the Performance of a Boiling Water Reactor by Torben Petersen August 1973 Saks distributors of Prediction of tbe Performence of a Boiling Water Reactor by Torben Petersen E R R A T A Page 1 6, equation

384

HFIR Plant Maintenance - August  

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

November 2012 November 2012 2 Managed by UT-Battelle for the U.S. Department of Energy ORNL Isotope Infrastructure Description * CFD boiling/multiphase models rely on tunable parameters * We study sensitivities of key outputs of a CFD benchmark problem using two codes: Star-CD and NPhase-CMFD. * We present validation of boiling models in Star-CD and Star- CCM+ for DEBORA and PSBT benchmark problems Sensitivity, verification, and validation studies of CFD boiling models (L3 milestone - THM.CFD.P5.03) Approach Results * Nphase will require wall boiling models in order to faithfully simulate CASL-relevant applications * We observed the largest sensitivities to the bubble diameter, the lift coefficient, and the turbulence dispersion model * For current boiling models, a systematic overestimation of

385

Microsoft Word - power_reactors_briggs.doc  

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

Most common - Boiling Water and Pressurized Most common - Boiling Water and Pressurized Water Reactors About 80% of the world's nuclear reactors used for generating electricity are either boiling water reactors or pressurized water reactors. Of these, about 30% are boiling water reactors and 70% are pressurized water reactors. All power reactors currently in use in the United States are of these two types. Both types of reactors have been very successfully used for reliable, on-demand, emissions-free electricity generation for decades. How does a boiling water reactor work? Water flows from the bottom of the fuel to the top of the fuel, and as it moves past the fuel, it carries away the heat produced by the

386

Leidenfrost temperature increase for impacting droplets on carbon-nanofiber surfaces  

E-Print Network (OSTI)

Droplets impacting on a superheated surface can either exhibit a contact boiling regime, in which they make direct contact with the surface and boil violently, or a film boiling regime, in which they remain separated from the surface by their own vapor. The transition from the contact to the film boiling regime depends not only on the temperature of the surface and kinetic energy of the droplet, but also on the size of the structures fabricated on the surface. Here we experimentally show that surfaces covered with carbon-nanofibers delay the transition to film boiling to much higher temperature compared to smooth surfaces. We present physical arguments showing that, because of the small scale of the carbon fibers, they are cooled by the vapor flow just before the liquid impact, thus permitting contact boiling up to much higher temperatures than on smooth surfaces. We also show that, as long as the impact is in the film boiling regime, the spreading factor of impacting droplets follows the same $\\We^{3/10}$ sc...

Nair, Hrudya; Tran, Tuan; van Houselt, Arie; Prosperetti, Andrea; Lohse, Detlef; Sun, Chao

2013-01-01T23:59:59.000Z

387

WASH-  

Office of Legacy Management (LM)

rcc.p,anc. 01 thts arf~cle. tha rcc.p,anc. 01 thts arf~cle. tha yubl~rhe, "r ~u~~iunl riknouu~adqnS the U.S. C;ov.rnmmnf' s rayhr (0 retam l nOn*aClulive.roy~ltV (r-0 ltconso In ma IO Dny Copvrlqhl WASH- covrrm~ the wtvdo. ISADIOLOGICAL SURVEY OF THE SEAWAY INDUSTRIAL PARK W . D. Cottrell, R. W . Leggett and H. W . Dickson Health Physics Division, Oak Ridge National Laboratory Oak Ridge, Tennessee 37830 December 1976 CONTENTS l&t of Tab1 es - . . . List of Illustrations . . Abstract . . . . . . Introduction . . . . ............ ............ ............ ............ Radiological Survey Techniques . . . . . . 1 . . Measurement of External Gamma and Beta-Gamma Radiation Levels . . . . . . . . . . . . Measurement of Radium in the Soil . . . . . . Measurement of Radioactivity in Surface Water

388

I:  

Office of Legacy Management (LM)

076181 076181 .-. -. ,- _- ,^, - THIS AGmmmiT, entered into this & day of MCUV I: 1991, effective as of the - day of , 1991 betvse; TIE UNITED STATES OF ANERICA, (hereinafter called tha %ovarnmanta) , acting through tha DEPARTKENT OP RNRRC!( (harsinaftsr called VOEn), and LCR-PSW PARTtmEm P (hereinafter callsd the "Licenser") uho is the fee owner of the parcel of land (hereinafter called the Premises) vhich is described in the deed title no. 43% R-01817 filed in the New York County Clerks Office and shoun on Exhibit 1, the exhibit being attached hereto and made part hereof. NITNESSETN TNAT: WHEREAS, the DOE desires to enter upon Licenser's Premisea for the purpose of performing certain remedial sctions as part of said program: and UREREAS, the Licenser is agreeable to the performance of

389

Oak Ridge  

Office of Legacy Management (LM)

~, ~, . ., . .- -. -_ .._ ..-. - .- ..- Oak Ridge Associated post Of%ce Box 117 Uniwx.ities Oak Riie. Tennessee 37631-0117 Apill. 1991 Ms. cethy Hickey Bschtel Nstiod Inc. P. 0. Box 350 Oak Ridge, Tn 378314350 Subject: BLDG. 621-527 - BAKER AND WlLLfAMS WAREHOUSES Deer Ms. Hiikey: 8etween March 1 l-22, 1991, the Envfronmental Suvey and Sine Assessment Program fESSAP1 of Oak Ridge Associated Urtiversities fORALl conducted a radiological charscterization euwey of the East end West besernent bays in Building 521-527 of the Etaker and Williams Warehouses. A review of tha rrurvey resufts indicate that ecthity exceeding criteria is present in four (4) locations in the East bay which will requfre decontamination. Dust samples were coflected from the floor and ledges in tha East bay. Direct measurements

390

bpv60e1.tmp  

Office of Scientific and Technical Information (OSTI)

f . ./l \ NANOFLUID TECHNOLOGY: CURRENT STATUS AND FUTURE RESEARCH Stephen U.-S. Choi Energy Technology Division Argonne National Laboratory Argonne, IL 60439 1- Tha subnsitfad manuacrip4 haa beancreatedby tha Univaraify of Chksgo as OpsraforofArgcrrneNatiial Laboratory (%rgonne") underContraof No. W-31-109-ENG- 3S * the U.S. Oaparimsnt of Energy.The U.S. Governmentrataiia forilaaif,and otharaacfingcmits bshaff, a @d-up, norsasdusiva, irrevocabfs woddwfde I&me in aa'darfkfs to raproduoa, preparederivathre works,distribute @@I& tOthe Pubfk,d P#Orrrrpubkiy and d-y Y bv or m behalfofthe Government. To be presented at the Second Korean-American Scientists and Engineers Association Research Trend Study Project Review and the Korea-U.S. Technical Conference on Strategic Technologies, October 22-24, 1998, Vienna, VA. Work supported by the U.S.

391

Synthesis and thermal decomposition properties of hydrogen-rich phosphorus salts.  

DOE Green Energy (OSTI)

Complex metal hydrides continue to be investigated as solid-materials for hydrogen storage. Traditional interstitial metal hydrides offer favorable thermodynamics and kinetics for hydrogen release but do not meet energy density requires. Anionic metal hydrides, and complex metal hydrides like magnesium borohydride have higher energy densities compared to interstitial metal hydrides, but poor kinetics and/or thermodynamically unfavorable side products limit their deployment as hydrogen storage materials in transportation applications. Main-group anionic materials such as the bis(borane)hypophosphite salt [PH2(BH3)2] have been known for decades, but only recently have we begun to explore their ability to release hydrogen. We have developed a new procedure for synthesizing the lithium and sodium hypophosphite salts. Routes for accessing other metal bis(borane)hypophosphite salts will be discussed. A significant advantage of this class of material is the air and water stability of the anion. Compared to metal borohydrides, which reactive violently with water, these phosphorus-based salts can be dissolved in protic solvents, including water, with little to no decomposition over the course of multiple days. The ability of these salts to release hydrogen upon heating has been assessed. While preliminary results indicate phosphine and boron-containing species are released, hydrogen is also a major component of the volatile species observed during the thermal decomposition. Additives such as NaH or KH mixed with the sodium salt Na[PH2(BH3)2] significantly perturb the decomposition reaction and greatly increase the mass loss as determined by thermal gravimetric analysis (TGA). This symbiotic behavior has the potential to affect the hydrogen storage ability of bis(borane)hypophosphite salts.

Cordaro, Joseph Gabriel

2010-12-01T23:59:59.000Z

392

Impacts of Low-NOX Combustion and Activated Carbon Injection on Particulate Control Device Performance  

Science Conference Proceedings (OSTI)

This report summarizes the results of a computational fluid dynamics (CFD) model study of the re-entrainment of carbon from the hoppers of a typical utility electrostatic precipitator (ESP). During earlier phases of this study, hopper re-entrainment was identified as the principle mechanism responsible for the low collection efficiency of carbon by ESPs. This statement was found to be true for both unburned carbon from the boiler and activated carbon injected for mercury control. The results indicate tha...

2008-03-31T23:59:59.000Z

393

Operation and Maintenance Guidelines for Selective Catalytic Reduction Systems  

Science Conference Proceedings (OSTI)

This document is the 2010 version of Operation and Maintenance Guidelines for Selective Catalytic Reduction Systems, originally published in 2001 and updated annually. New content this year includes: (1) A section on static mixers added in Chapter 3; (2) Substantial expansion of the discussion on inspection of ammonia storage and delivery equipment in Chapter 8; (3) Expanded coverage of unit startup, shutdown, and low-load operation in Chapter 15; (4) a new chapter, Chapter 18, on the means to ensure tha...

2010-11-29T23:59:59.000Z

394

Collateral Risk Analytics for Energy Trading and Portfolio Risk Management  

Science Conference Proceedings (OSTI)

This report describes the need for a strong collateral risk management function as an integral part of an energy company’s risk management program. It reviews the basics of margining and collateral both in over-the-counter markets and on exchanges. In addition, it details the technology available to measure collateral risk properly. Then it reviews the recent efforts to regulate OTC derivatives, the potential impact that it could have on energy companies’ management of cash collateral, and strategies tha...

2010-12-21T23:59:59.000Z

395

Program on Technology Innovation: Information Integration for Equipment Reliability at Nuclear Plants  

Science Conference Proceedings (OSTI)

This report investigates the status of information integration for equipment reliability (ER) at nuclear power plants. ER consists of a comprehensive set of processes that span the organization and require extensive data gathering, retrieval, and information integration. To assist nuclear operators, the Institute of Nuclear Power Operations (INPO) issued AP-913, Equipment Reliability Process Description, as a standard approach to implement effective ER processes among its members. Despite the success tha...

2009-04-30T23:59:59.000Z

396

Diagnosis and Therapy According to the rGyud-bzi  

E-Print Network (OSTI)

If ba ra!!i zo dar mar gsar ri dvags sa 1/ ra sa skom sa gsar pa chag che dan If skyabs dai. khur chad chab tha chu bsil dai. 1/ bskol gram mkhris J'il!!i nod kyi zas su bsati /I lug, dai. g-yag rgod gcan gzan iia yi sa /I sbran rti skom sa!!i f...

Finckh, Elizabeth

1993-01-01T23:59:59.000Z

397

Coal liquefaction process  

DOE Patents (OSTI)

A process is described for the liquefaction of coal wherein raw feed coal is dissolved in recycle solvent with a slurry containing recycle coal minerals in the presence of added hydrogen at elevated temperature and pressure. The highest boiling distillable dissolved liquid fraction is obtained from a vacuum distillation zone and is entirely recycled to extinction. Lower boiling distillable dissolved liquid is removed in vapor phase from the dissolver zone and passed without purification and essentially without reduction in pressure to a catalytic hydrogenation zone where it is converted to an essentially colorless liquid product boiling in the transportation fuel range. 1 fig.

Wright, C.H.

1986-02-11T23:59:59.000Z

398

BWR refill-reflood program Task 4. 7: constitutive correlations for shear and heat transfer for the BWR version of TRAC  

SciTech Connect

TRAC (Transient Reactor Analysis Code) is a computer code for best estimate analysis of the thermal hydraulic conditions in a reactor system. The constitutive correlations for shear and heat transfer in the boiling water reactor (BWR) version of TRAC are described. A new model, that accounts for the effect of phase and velocity profiles, has been developed for the interfacial shear and a new set of constitutive correlations are derived. Improvements have been made to the heat transfer in the area of subcooled boiling, boiling transition, and thermal radiation.

Andersen, J.G.M.; Chu, K.H.

1982-11-01T23:59:59.000Z

399

Coal liquefaction process  

DOE Green Energy (OSTI)

A process for the liquefaction of coal wherein raw feed coal is dissolved in recycle solvent with a slurry containing recycle coal minerals in the presence of added hydrogen at elevated temperature and pressure. The highest boiling distillable dissolved liquid fraction is obtained from a vacuum distillation zone and is entirely recycled to extinction. Lower boiling distillable dissolved liquid is removed in vapor phase from the dissolver zone and passed without purification and essentially without reduction in pressure to a catalytic hydrogenation zone where it is converted to an essentially colorless liquid product boiling in the transportation fuel range.

Wright, Charles H. (Overland Park, KS)

1986-01-01T23:59:59.000Z

400

R. S. Driof, Process Demlopnant Dranch, Production Division  

Office of Legacy Management (LM)

S. Driof, Process Demlopnant Dranch, S. Driof, Process Demlopnant Dranch, Production Division 7 i 7; I; $ " k>JSTI'IC AT TIE Cif~iICAL CCNSTXICTIOS COXi'O+TIO:? P$IX)T PIGIT-JUL'I 31, 19% Chemico ban fouzd tw proossses, b&h involving the initial H2SOl lwc:?, sutisfoctory. On.3 process (rerun) produces a U-Cu precipitate r&ich is ralsachad; the U and Cu can ba s+paratzd by various nothods. The second process (sts~~~ise) ?rucipitn?es co?ps~ and thee uranium. &j- ditioml 1abnretorJ xork is being dona so that thase processas cm b4 coa>c'r3d uooixmically. Discussion Kcrssru. Dvshor, ?icksns, Trincs, Snrkssian, and Atkins of Chonico and %assrs. Rridf and Coddo of the AEC mt at tha Linden Pilot Plaflt m july 31 to reviaw tha most recer,t dvvnloPnant work pcrfomad by tha

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While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
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401

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

402

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

403

BWRVIP-239: BWR Vessel and Internals Project, Updated Evaluation of the Integrated Surveillance Program (ISP) Capsule Withdrawal Sch edule  

Science Conference Proceedings (OSTI)

This report evaluates updated reactor pressure vessel and surveillance capsule fluence data for potential impacts on the Boiling Water Reactor Vessel and Internals Project Integrated Surveillance Program (BWRVIP ISP) capsule withdrawal schedule.

2010-07-16T23:59:59.000Z

404

Bubble dynamics on structured surface in microchannel  

E-Print Network (OSTI)

Surface enhancement is a potential way to improve the performance of flow boiling in microchannels, which is considered to be one of the most promising cooling methods to solve thermal management challenges faced by future ...

Chen, Siyu, S.M. Massachusetts Institute of Technology

2013-01-01T23:59:59.000Z

405

Page not found | Department of Energy  

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

make electricity by boiling water to make steam that turns a turbine. A nuclear power plant works this way, too. At a nuclear power plant, splitting atoms produce the heat to...

406

TWOZONE USERS MANUAL  

E-Print Network (OSTI)

C-OIL) The current cost of electricity. (11-20) BOIL, gal. (current cost of natural gas. The current cost of fuel oil. (CASE ENERGr USE AND COST JOOO.O THERMS. fUEL OIL- FUEL OIL-

Gadgil, Ashok J.

2008-01-01T23:59:59.000Z

407

TWOZONE USERS MANUAL  

E-Print Network (OSTI)

C-OIL) The current cost of electricity. (11-20) BOIL, gal. (current cost of natural gas. The current cost of fuel oil. (CASE ENERGr USE AND COST JOOO.O THERMS. fUEL OIL- FUEL OIL-

Gadgil, Ashok J.

2010-01-01T23:59:59.000Z

408

Cross section generation strategy for high conversion light water reactors  

E-Print Network (OSTI)

High conversion water reactors (HCWR), such as the Resource-renewable Boiling Water Reactor (RBWR), are being designed with axial heterogeneity of alternating fissile and blanket zones to achieve a conversion ratio of ...

Herman, Bryan R. (Bryan Robert)

2011-01-01T23:59:59.000Z

409

BWRVIP-229: BWR Vessel and Internals Project, Alloy X-750 Replacement Roadmap  

Science Conference Proceedings (OSTI)

Utilities require degradation-resistant, high-strength material for use in performing repairs to boiling water reactors (BWRs) as well as for use in new construction. This report reviews a number of alloys for applicability to BWR service conditions.

2009-12-11T23:59:59.000Z

410

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

411

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 ...

412

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...

413

The use of oil shale ash in the production of biodiesel from waste vegetable oil  

Science Conference Proceedings (OSTI)

Oil shale ash obtained from combustion of local oil shale deposits was used in this study as a heterogeneous catalyst to produce biodiesel from waste vegetable oil (WVO). Two alcohols with high and low boiling points

A. Al-Otoom; M. Allawzi; A. Ajlouni; F. Abu-Alrub; M. Kandah

2012-01-01T23:59:59.000Z

414

An Updated Conceptual Model Of The Los Humeros Geothermal Reservoir...  

Open Energy Info (EERE)

sea level (a.s.l.) the pressure profile of which corresponds to a 300-330C boiling water column and a deeper low-liquid-saturation reservoir located between 850 and 100 m...

415

Effects of thermal aging on Stress Corrosion Cracking and mechanical properties of stainless steel weld metals  

E-Print Network (OSTI)

Stress Corrosion Cracking (SCC) in and around primary loop piping welds in Boiling Water Reactors has been observed worldwide as plants continue to operate at temperatures and pressures near 2880C (5500F) and 6.9 MPa (1000 ...

Hixon, Jeff

2006-01-01T23:59:59.000Z

416

Stress corrosion cracking and crack tip characterization of Alloy X-750 in light water reactor environments  

E-Print Network (OSTI)

Stress corrosion cracking (SCC) susceptibility of Inconel Alloy X-750 in the HTH condition has been evaluated in high purity water at 93 and 288°C under Boiling Water Reactor Normal Water Chemistry (NWC) and Hydrogen Water ...

Gibbs, Jonathan Paul

2011-01-01T23:59:59.000Z

417

Use of once-through treat gas to remove the heat of reaction in solvent hydrogenation processes  

DOE Patents (OSTI)

In a coal liquefaction process wherein feed coal is contacted with molecular hydrogen and a hydrogen-donor solvent in a liquefaction zone to form coal liquids and vapors and coal liquids in the solvent boiling range are thereafter hydrogenated to produce recycle solvent and liquid products, the improvement which comprises separating the effluent from the liquefaction zone into a hot vapor stream and a liquid stream; cooling the entire hot vapor stream sufficiently to condense vaporized liquid hydrocarbons; separating condensed liquid hydrocarbons from the cooled vapor; fractionating the liquid stream to produce coal liquids in the solvent boiling range; dividing the cooled vapor into at least two streams; passing the cooling vapors from one of the streams, the coal liquids in the solvent boiling range, and makeup hydrogen to a solvent hydrogenation zone, catalytically hydrogenating the coal liquids in the solvent boiling range and quenching the hydrogenation zone with cooled vapors from the other cooled vapor stream.

Nizamoff, Alan J. (Convent Station, NJ)

1980-01-01T23:59:59.000Z

418

Proceedings of IMECE04 2004 ASME International Mechanical Engineering Congress  

E-Print Network (OSTI)

that in most cases there is no boiling in the coolant. The nuclear fuel is constituted of ceramic pellets is still lacking. Over the last twenty years the U.S. nuclear industry has significantly increased capacity

Motta, Arthur T.

419

Brie and Mushroom Soup A vegetarian soup of assorted mushrooms simmered in a creamy vegetable broth  

E-Print Network (OSTI)

-Almond Chicken Salad Diced chicken breast blended with honey, mayonnaise, onion, and celery and garnished. Garnished with cucumber, tomato, and boiled egg. $6.95 Cheese and Fresh Fruit Plate Smoked Gouda, Brie

Arnold, Jonathan

420

Numerical modeling of pool spreading, heat transfer and evaporation in liquefied natural gas (LNG).  

E-Print Network (OSTI)

?? This master's thesis is a continuation of previous theses written at ComputIT AS. It treats heat transfer to LNG pools boiling on water through… (more)

Myrmo, Øystein

2011-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "naph tha boiling" 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

Nanofluid heat transfer enhancement for nuclear reactor applications  

E-Print Network (OSTI)

Colloidal dispersions of nanoparticles are known as `nanofluids'. Such engineered fluids offer the potential for enhancing heat transfer, particularly boiling heat transfer, while avoiding the drawbacks (i.e., erosion, ...

Buongiorno, Jacopo

422

Heat transfer to impacting drops and post critical heat flux dispersed flow  

E-Print Network (OSTI)

Heat transfer to drops impacting on a hot surface is examined in context of dispersions of flowing, boiling fluids. The liquid contribution to heat transfer from a hot tube to a two-phase dispersion is formulated in terms ...

Kendall, Gail E.

1978-01-01T23:59:59.000Z

423

It's Elemental - The Element Promethium  

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

(Samarium) Samarium The Element Promethium Click for Isotope Data 61 Pm Promethium 145 Atomic Number: 61 Atomic Weight: 145 Melting Point: 1315 K (1042C or 1908F) Boiling...

424

Development of Heat Transfer Enhancement Techniques for ExternalCooling of an Advanced Reactor Vessel.  

E-Print Network (OSTI)

??Nucleate boiling is a well-recognized means for passively removinghigh heat loads (up to 106 W/m2) generated by amolten reactor core under severe accident conditions whilemaintaining… (more)

Yang, Jun

2005-01-01T23:59:59.000Z

425

NASA Advisory Council Space Operations Committee July 28, 2010  

E-Print Network (OSTI)

panels, Hubble spectrometer · Corrosion Lab, e.g., coating with microcapsules, self detection · Cryogenics Test Lab, e.g., tank insulation to minimize boil-off; aerogels; wire insulation

426

NASA Advisory Council Space Operations Committee July 2010  

E-Print Network (OSTI)

to solar panels Hubble spectrometer · · · Corrosion Lab · Coatings with microcapsules ­ self healing to minimize boil-off Aerogels Wire insulation ­ detection and healing layer · · Collaboration between NASA

427

Conceptual Design of Molten Salt Loop Experiment for MIT Research Reactor  

E-Print Network (OSTI)

Molten salt is a promising coolant candidate for Advanced High Temperature Reactor (AHTR) Gen-IV designs. The low neutron absorption, high thermal capacity, chemical inertness, and high boiling point at low pressure of ...

Bean, Malcolm K.

2011-08-01T23:59:59.000Z

428

Mechanics and Mechanisms of Environmentally Assisted Cracking of Alloys 132/182 in BWR and PWR Environments  

Science Conference Proceedings (OSTI)

This report documents research on the mechanics and mechanisms of environmentally assisted cracking of Alloys 132/182 in boiling water reactor (BWR) and pressurized water reactor (PWR) environments.

2004-10-18T23:59:59.000Z

429

System design and dynamic signature identification for intelligent energy management in residential buildings.  

E-Print Network (OSTI)

hvac (t) (1 for on Then the system identication process boils down to optimizationHVAC switch are briey explained. The rest of this chapter is fo- cused on optimization

Jang, Jaehwi

2008-01-01T23:59:59.000Z

430

Stability analysis of natural circulation in BWRs at high pressure conditions  

E-Print Network (OSTI)

At rated conditions, a natural circulation boiling water reactor (NCBWR) depends completely on buoyancy to remove heat from the reactor core. This raises the issue of potential unstable flow. oscillations. The objective ...

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

2007-01-01T23:59:59.000Z

431

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 ...

432

An Examination of Avoided Costs in Utah  

E-Print Network (OSTI)

ultimately accepted a natural gas price projection that wasfrom the NWPPC’ s natural gas price forecast (basis East-about future natural gas prices, this issue really boils dow

Bolinger, Mark; Wiser, Ryan

2005-01-01T23:59:59.000Z

433

Stress Corrosion Cracking and Crack Tip Characterization of Alloy X-750 in Light Water Reactor Environments  

E-Print Network (OSTI)

Stress corrosion cracking (SCC) susceptibility of Inconel Alloy X-750 in the HTH condition has been evaluated in high purity water at 93 and 288°C under Boiling Water Reactor Normal Water Chemistry (NWC) and Hydrogen Water ...

Gibbs, Jonathan Paul

434

Micro/Nano-Scale Phase Change Systems for Thermal Management and Solar Energy Conversion Applications  

E-Print Network (OSTI)

Investigation of a High Flux Heat Pipe Heat Sink,” JournalBoiling in a Flat Grooved Heat Pipe,” International JournalReay D. A. , 1976, Heat Pipes, Pergamon Press, Elmsford. [

Coso, Dusan

2013-01-01T23:59:59.000Z

435

Design of annular fuel for high power density BWRs  

E-Print Network (OSTI)

Enabling high power density in the core of Boiling Water Reactors (BWRs) is economically profitable for existing or new reactors. In this work, we examine the potential for increasing the power density in BWR plants by ...

Morra, Paolo

2005-01-01T23:59:59.000Z

436

Environmental Energy Technologies Division News  

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

involves bringing to a boil a thin mixture of water and flour. The flour is a mix of sorghum, millet, wheat, and corn, supplied by the United Nations World Food Program. More...

437

A PROCESS FOR SEPARATING AZEOTROPIC MIXTURES BY EXTRACTIVE AND CONVECTIVE DISTILLATION  

DOE Patents (OSTI)

A method is described for separating an azeotrope of carbon tetrachloride and 1,1,2,2-tetrafluorodinitroethane boiling at 60 deg C. The ndethod comnprises, specifically, feeding azeotrope vapors admixed with a non- reactive gas into an extractive distillation column heated to a temperature preferably somewhat above the boiling point of the constant boiling mixture. A solvent, di-n-butylphthalate, is metered into the column above the gas inlet and permitted to flow downward, earrying with it the higher bomling fraction, while the constituent having the lower boiling point passes out of the top of the column with the non-reactive gas and is collected in a nitrogen cold trap. Other solvents which alter the vapor pressure relationship may be substituted. The method is generally applicable to azeotropic mixtures. A number of specific mixtures whicb may be separated are disclosed. (AEC)

Frazer, J.W.

1961-12-19T23:59:59.000Z

438

EPRI Perspective on TUV Evaluation of KKM Core Shroud Repair  

Science Conference Proceedings (OSTI)

This report provides the Electric Power Research Institute (EPRI) perspective for why reliance on tie rods alone is an acceptable approach to managing stress corrosion cracking (SCC) of boiling water reactor (BWR) core shroud horizontal welds.

2009-04-29T23:59:59.000Z

439

The 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12) Sheraton Station Square, Pittsburgh, Pennsylvania, U.S.A. September 30-October 4, 2007  

E-Print Network (OSTI)

to the insolated top Pyrex surface, which suggests that the void fraction was significantly higher than it appears, P., 1999, "Experimental Investigations on Boiling of N-Pentane Across a Horizontal Tube Bundle: Two

Paris-Sud XI, Université de

440

Fictitious domain methods for two-phase flow energy balance computations in nuclear  

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

Paris-Sud XI, Université de

Note: This page contains sample records for the topic "naph tha boiling" 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

Process for the hydroisomerization and hydrocracking of Fisher-Tropsch waxes to produce a syncrude and upgraded hydrocarbon products  

SciTech Connect

A process is described for producing a pumpable syncrude from a Fischer-Tropsch wax containing oxygenate compounds, which comprises: (1) separating the Fischer-Tropsch wax into (a) a low-boiling fraction which contains most of the oxygenate compounds and (b) a high-boiling fraction which is substantially free of water and oxygenate compounds, (2) reacting the high-boiling fraction from step (1) with hydrogen at hydroisomerization and mild hydrocracking conditions in the presence of a fluorided Group VIII metal-on-alumina catalyst to produce a C/sub 5/ + hydrocarbon product, and (3) combining the C/sub 5/ + hydrocarbon product from step (2) with the low-boiling fraction from step (1) to produce a pumpable, refinery processable syncrude that can be transported at atmospheric conditions.

Hamner, G.P.

1989-05-23T23:59:59.000Z

442

Measurements of the Electrical Conductivities of Air over Hot Water  

Science Conference Proceedings (OSTI)

Measurements of the conduction current between two electrodes in air over recently boiled water have been interpreted by Carlon as indicating that the humidified air became highly conductive and that large numbers of ions were produced in the air ...

C. B. Moore; B. Vonnegut

1988-03-01T23:59:59.000Z

443

Effects of Micro/Nano-Scale Surface Characteristics on the Leidenfrost Point Temperature of Water  

E-Print Network (OSTI)

In recent film boiling heat transfer studies with nanofluids, it was reported that deposition of nanoparticles on a surface significantly increases the nominal minimum heat flux (MHF) or Leidenfrost Point (LFP) temperature, ...

Hu, Lin-Wen

444

Fermilab | Science at Fermilab | Experiments & Projects | Cosmic...  

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

will not boil unless disturbed. The first COUPP detector was a thin-walled quartz bell jar containing a liter of iodotrifluoromethane, a fire-extinguishing liquid also called...

445

Fermilab Today | Experiment Profiles Archive | COUPP  

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

number of years. PHYSICS FRONTIER EXPLORING: Cosmic Frontier HOW DOES IT WORK? A quartz jar serves as a bubble chamber and holds a liquid kept just above its normal boiling point,...

446

Glossary Term - Liquid Nitrogen  

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

Lepton Previous Term (Lepton) Glossary Main Index Next Term (Mercury) Mercury Liquid Nitrogen Liquid nitrogen boils in a frying pan on a desk. The liquid state of the element...

447

Team Bug Bag Biogas For Nicaragua  

E-Print Network (OSTI)

Team Bug Bag Biogas For Nicaragua Project Recap The task for Team Bug Bag was to create for under $100 (USD), and be able to produce biogas that could boil water for a thirty minute time period

Demirel, Melik C.

448

MSU Extension Publication Archive Archive copy of publication, do not use for current recommendations. Up-to-date  

E-Print Network (OSTI)

. NUTRITION INFORMATION One medium raw tomato has: · 26 Calories. · 0 grams fat. · 1 gram fiber. Tomatoes with chicken or tuna salad, cottage cheese or potato salad. · To peel fresh tomatoes, place them in boiling

449

The hardening of Type 316L stainless steel welds with thermal aging  

E-Print Network (OSTI)

Welded stainless steel piping is a component of boiling water reactors (BWRs). Reirculation and other large diameter piping are fabricated from Type 304 or 316 stainless steels. Delta ferrite is present in welds, because ...

Ayers, Lauren Juliet

2012-01-01T23:59:59.000Z

450

BWR Assembly Optimization for Minor Actinide Recycling  

Science Conference Proceedings (OSTI)

The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

2010-03-22T23:59:59.000Z

451

Program on Technology Innovation: Effects of High Temperature on ESBWR Primary Containment  

Science Conference Proceedings (OSTI)

The civil/structural design of the Economic and Simplified Boiling Water Reactor (ESBWR), General Electric's most advanced boiling water reactor, is an important aspect of the design that will be reviewed by the NRC as part of obtaining Final Design Approval. This report describes the advanced engineering modeling and nonlinear analysis techniques currently being used in support of the structural design of the ESBWR primary containment and reactor building for operations at elevated temperatures. The rep...

2005-12-19T23:59:59.000Z

452

BWRVIP-60-A: BWR Vessel and Internals Project, Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in th e BWR Environment  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals materials issues. This report provides a methodology for assessing crack growth in BWR low alloy steel pressure vessels and nozzles. A previous version of this report was published as BWRVIP-60 (TR-108709). This report (BWRVIP-60-A) incorporates the U.S. Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) and ot...

2003-06-09T23:59:59.000Z

453

5.8. Treatment of Extreme Operating Conditions As discussed in Chapter 4, the model may encounter conditions not intended by the  

E-Print Network (OSTI)

of vaporization, kJ/kg hsp Single-phase convective heat transfer coefficient, W/m2 o C htp Flow boiling heat wall heat transfer Nu4 Nusselt number for laminar fully-developed flow for four wall heat transfer Pout be evaluated from Eqs. (4)-(7) with Tc,out replaced by Tsat and hsp by htp, the flow boiling heat transfer

454

Coal Liquefaction desulfurization process  

DOE Patents (OSTI)

In a solvent refined coal liquefaction process, more effective desulfurization of the high boiling point components is effected by first stripping the solvent-coal reacted slurry of lower boiling point components, particularly including hydrogen sulfide and low molecular weight sulfur compounds, and then reacting the slurry with a solid sulfur getter material, such as iron. The sulfur getter compound, with reacted sulfur included, is then removed with other solids in the slurry.

Givens, Edwin N. (Bethlehem, PA)

1983-01-01T23:59:59.000Z

455

SUBCOOLING DETECTOR  

DOE Patents (OSTI)

A system for detecting and measuring directly the subcooling margin in a liquid bulk coolant is described. A thermocouple sensor is electrically heated, and a small amount of nearly stagnant bulk coolant is heated to the boiling point by this heated thermocouple. The sequential measurement of the original ambient temperature, zeroing out this ambient temperature, and then measuring the boiling temperature of the coolant permits direct determination of the subcooling margin of the ambient liquid. (AEC)

McCann, J.A.

1963-12-17T23:59:59.000Z

456

Assessment of Experimental Data to Support Computational Fluid Dynamics Analysis of PWR Rod Bundle Heat Transfer Studies  

Science Conference Proceedings (OSTI)

Crud-induced cladding corrosion (CILC) is a localized phenomenon, which is directly related to subcooled nucleate boiling (SNB) in rod bundles of pressurized water reactors (PWRs). Local boiling on fuel rod surfaces leads to preferential deposition of corrosion products circulating in the reactor coolant. Typical thermal hydraulic codes/methods used in core design do not have sufficient resolution to predict susceptible "hot spots" on fuel rod surfaces when SNB is elevated. Hence, detailed local computat...

2010-08-30T23:59:59.000Z

457

BWRVIP-34-A: BWR Vessel and Internals Project, Technical Basis for Part Circumference Weld Overlay Repair of Vessel Internal Core Sp ray Piping  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals issues. This report summarizes the results of the design and analysis activities and the testing programs conducted to provide BWR utilities with a contingency repair option for internal core spray piping for BWR2/6 plants. A previous version of this report was published as BWRVIP-34 (TR-108198). This report (BWRVIP-34...

2008-03-11T23:59:59.000Z

458

BWRVIP-18, Revision 1-A: BWR Vessel and Internals Project, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June 1994, is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals issues. This BWRVIP report contains generic guidelines that describe locations on the core spray piping and spargers for which inspection is needed, categories of plants for which inspection needs would differ, extent of inspection and reinspection for each location, and flaw evaluation procedures to determine ...

2012-04-09T23:59:59.000Z

459

BWRVIP-27-A: BWR Vessel and Internals Project, BWR Standby Liquid Control System / Core Plate Delta-P Inspection and Flaw Evaluation Guidelines  

Science Conference Proceedings (OSTI)

The Boiling Water Reactor Vessel and Internals Project (BWRVIP), formed in June, 1994, is an association of utilities focused exclusively on boiling water reactor (BWR) vessel and internals issues. This BWRVIP report defines inspection requirements for BWR standby liquid control (SLC) system piping from the vessel nozzle safe-end inward. A previous version of this report was published as BWRVIP-27 (TR-107286). This report (BWRVIP-27-A) incorporates changes proposed by the BWRVIP in response to "U.S. Nucl...

2003-08-04T23:59:59.000Z

460

Highly Selective Membranes For The Separation Of Organic Vapors Using Super-Glassy Polymers  

DOE Patents (OSTI)

A process for separating hydrocarbon gases of low boiling point, particularly methane, ethane and ethylene, from nitrogen. The process is performed using a membrane made from a super-glassy material. The gases to be separated are mixed with a condensable gas, such as a C.sub.3+ hydrocarbon. In the presence of the condensable gas, improved selectivity for the low-boiling-point hydrocarbon gas over nitrogen is achieved.

Pinnau, Ingo (Palo Alto, CA); Lokhandwala, Kaaeid (Menlo Park, CA); Nguyen, Phuong (Fremont, CA); Segelke, Scott (Mountain View, CA)

1997-11-18T23:59:59.000Z

Note: This page contains sample records for the topic "naph tha boiling" 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

BWRVIP-118: BWR Vessel and Internals Project: NMCA Experience Report and Applications Guidelines, 2003 Revision  

Science Conference Proceedings (OSTI)

The boiling water reactor (BWR) fleet has widely embraced noble metal chemical addition (NMCA) to provide protection against intergranular stress corrosion cracking (IGSCC). This report, prepared by a Boiling Water Reactor Vessel and Internals Project (BWRVIP) focus group, updates a report issued in 2001 that compiled data from plants operating on NMCA. It provides guidance for BWRs planning to implement NMCA, and information about expected plant response to operation with NMCA. It also identifies steps ...

2003-11-24T23:59:59.000Z

462

Towards CFD Modelling of Critical Heat Flux in Fuel Rod Bundles  

SciTech Connect

The paper describes actual CFD approaches to subcooled boiling and investigates their capability to contribute to fuel assembly design. In a prototype version of the CFD code CFX a wall boiling model is implemented based on a wall heat flux partition algorithm. It can be shown, that the wall boiling model is able, to calculate the cross sectional averaged vapour volume fraction with good agreement to published measurements. The most sensitive parameters of the model are identified. Needs for more detailed experiments are established which are necessary to support further model development. Nevertheless in the paper the model is applied for the investigation of the phenomena inside a hot channel in a fuel assembly. Here the essential parameter is the critical heat flux. Although subcooled boiling represents only a preliminary state toward critical heat flux essential parameters like the swirl, the cross flow between adjacent channels and concentration regions of bubbles can be determined. By calculating the temperature at the rod surface the critical regions can be identified which might later on lead to departure from nucleate boiling and possible damage of the fuel pin. The application of up-to-date CFD with a subcooled boiling model for the simulation of a hot channel enables the comparison and the evaluation of different geometrical designs of the spacer grids of a fuel rod bundle. (authors)

Krepper, Eckhard [Forschungszentrum Rossendorf e.V., Institute of Safety Research, D-01314 Dresden, POB 510119 (Germany); Egorov, Yury [ANSYS Germany GmbH Staudenfeldweg 12, D-83624 Otterfing (Germany); Koncar, Bostjan ['Jozef Stefan' Institute Jamova 39, 1000 Ljubljana (Slovenia)

2006-07-01T23:59:59.000Z

463

IDENTIFICATION AND EXPERIMENTAL DATABASE FOR BINARY AND MULTICOMPONENT MIXTURES WITH POTENTIAL FOR INCREASING OVERALL CYCLE EFFICIENCY  

SciTech Connect

This report describes an experimental investigation designed to identify binary and multicomponent mixture systems that may be for increasing the overall efficiency of a coal fired unit by extracting heat from flue gases. While ammonia-water mixtures have shown promise for increasing cycle efficiencies in a Kalina cycle, the costs and associated range of thermal conditions involved in a heat recovery system may prohibit its use in a relatively low temperature heat recovery system. This investigation considered commercially available non-azeotropic binary mixtures with a boiling range applicable to a flue gas initially at 477.6 K (400 F) and developed an experimental database of boiling heat transfer coefficients for those mixtures. In addition to their potential as working fluids for increasing cycle efficiency, cost, ease of handling, toxicity, and environmental concerns were considered in selection of the mixture systems to be examined experimentally. Based on this review, water-glycol systems were identified as good candidates. However, previous investigations of mixture boiling have focused on aqueous hydrocarbon mixtures, where water is the heaviest component. There have been few studies of water-glycol systems, and those that do exist have investigated boiling on plain surfaces only. In water-glycol systems, water is the light component, which makes these systems unique compared to those that have been previously examined. This report examines several water-glycol systems, and documents a database of experimental heat transfer coefficients for these systems. In addition, this investigation also examines the effect of an enhanced surface on pool boiling in water-glycol mixtures, by comparing boiling on a smooth surface to boiling on a Turbo IIIB. The experimental apparatus, test sections, and the experimental procedures are described. The mixture systems tested included water-propylene glycol, water-ethylene glycol, and water-diethylene glycol. All experimental data were obtained at atmospheric pressure with the test section oriented horizontally. The effect of subcooling in pool boiling of mixtures is another area that has received limited attention. Therefore, experimental data were obtained for the water-propylene glycol and water-ethylene glycol systems for subcoolings ranging from 0 to 30 C. The experimental data showed that boiling heat transfer coefficients were found to have significant degradation due to the mixture effect for each of the water-glycol systems examined. This result is consistent with previous studies which examined water-hydrocarbon mixtures with large boiling ranges. The Turbo BIII surface was found to significantly increase heat transfer in each mixture and pure component in comparison to that for the smooth surface.

Stephen M Bajorek; J. Schnelle

2002-05-01T23:59:59.000Z

464

On-sun test results from second-generation and advanced-concepts alkali-metal pool-boiler receivers  

DOE Green Energy (OSTI)

Two 75-kW{sub t} alkali-metal pool-boiler solar receivers have been successfully tested at Sandia National Laboratories` National Solar Thermal Test Facility. The first one, Sandia`s `` second-generation pool-boiler receiver,`` was designed to address commercialization issues identified during post-test assessment of Sandia`s first-generation pool-boiler receiver. It was constructed from Haynes alloy 230 and contained the alkali-metal alloy NaK-78. The absorber`s wetted side had a brazed-on powder-metal coating to stabilize boiling. This receiver was evaluated for boiling stability, hot- and warm-restart behavior, and thermal efficiency. Boiling was stable under all conditions. All of the hot restarts were successful. Mild transient hot spots observed during some hot restarts were eliminated by the addition of 1/3 torr of xenon to the vapor space. All of the warm restarts were also successful. The heat-transfer crisis that damaged the first receiver did not recur. Thermal efficiency was 92.3% at 750{degrees}C with 69.6 kW{sub t} solar input. The second receiver tested, Sandia`s ``advanced-concepts receiver,`` was a replica of the first-generation receiver except that the cavities, which were electric-discharge-machined in the absorber for boiling stability, were eliminated. This step was motivated by bench-scale test results that showed that boiling stability improved with increased heated-surface area, tilt of the heated surface from vertical, and added xenon. The bench-scale results suggested that stable boiling might be possible without heated-surface modification in a 75-kW{sub t} receiver. Boiling in the advanced-concepts receiver with 1/3 torr of xenon added has been stable under all conditions, confirming the bench-scale tests.

Moreno, J.B.; Andraka, C.E.; Moss, T.A.; Cordeiro, P.G.; Dudley, V.E.; Rawlinson, K.S.

1994-05-01T23:59:59.000Z

465

CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling  

SciTech Connect

In-vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding of the reactor cavity during a severe accident. As part of a joint Korean – United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boiling experiments were performed in the SBLB (Subscale Boundary Layer Boiling) facility at Penn State using test vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady state boiling, the vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the vessel outer surface but with very little effect of aging.

Fan-Bill Cheung; Joy L. Rempe

2004-06-01T23:59:59.000Z

466

Method and apparatus for processing a test sample to concentrate an analyte in the sample from a solvent in the sample  

DOE Patents (OSTI)

A method of processing a test sample to concentrate an analyte in the sample from a solvent in the sample includes: (a) boiling the test sample containing the analyte and solvent in a boiling chamber to a temperature greater than or equal to the solvent boiling temperature and less than the analyte boiling temperature to form a rising sample vapor mixture; (b) passing the sample vapor mixture from the boiling chamber to an elongated primary separation tube, the separation tube having internal sidewalls and a longitudinal axis, the longitudinal axis being angled between vertical and horizontal and thus having an upper region and a lower region; (c) collecting the physically transported liquid analyte on the internal sidewalls of the separation tube; and (d) flowing the collected analyte along the angled internal sidewalls of the separation tube to and pass the separation tube lower region. The invention also includes passing a turbulence inducing wave through a vapor mixture to separate physically transported liquid second material from vaporized first material. Apparatus is also disclosed for effecting separations. Further disclosed is a fluidically powered liquid test sample withdrawal apparatus for withdrawing a liquid test sample from a test sample container and for cleaning the test sample container. 8 figs.

Turner, T.D.; Beller, L.S.; Clark, M.L.; Klingler, K.M.

1997-10-14T23:59:59.000Z

467

Method and apparatus for processing a test sample to concentrate an analyte in the sample from a solvent in the sample  

DOE Patents (OSTI)

A method of processing a test sample to concentrate an analyte in the sample from a solvent in the sample includes: a) boiling the test sample containing the analyte and solvent in a boiling chamber to a temperature greater than or equal to the solvent boiling temperature and less than the analyte boiling temperature to form a rising sample vapor mixture; b) passing the sample vapor mixture from the boiling chamber to an elongated primary separation tube, the separation tube having internal sidewalls and a longitudinal axis, the longitudinal axis being angled between vertical and horizontal and thus having an upper region and a lower region; c) collecting the physically transported liquid analyte on the internal sidewalls of the separation tube; and d) flowing the collected analyte along the angled internal sidewalls of the separation tube to and pass the separation tube lower region. The invention also includes passing a turbulence inducing wave through a vapor mixture to separate physically transported liquid second material from vaporized first material. Apparatus are also disclosed for effecting separations. Further disclosed is a fluidically powered liquid test sample withdrawal apparatus for withdrawing a liquid test sample from a test sample container and for cleaning the test sample container.

Turner, Terry D. (Idaho Falls, ID); Beller, Laurence S. (Idaho Falls, ID); Clark, Michael L. (Menan, ID); Klingler, Kerry M. (Idaho Falls, ID)

1997-01-01T23:59:59.000Z

468

Catalytic two-stage coal hydrogenation process using extinction recycle of heavy liquid fraction  

DOE Patents (OSTI)

A process is described for catalytic two-stage hydrogenation and liquefaction of coal with selective extinction recycle of all heavy liquid fractions boiling above a distillation cut point of about 600--750 F to produce increased yields of low-boiling hydrocarbon liquid and gas products. In the process, the particulate coal feed is slurried with a process-derived liquid solvent normally boiling above about 650 F and fed into a first stage catalytic reaction zone operated at conditions which promote controlled rate liquefaction of the coal, while simultaneously hydrogenating the hydrocarbon recycle oils. The first stage reactor is maintained at 710--800 F temperature, 1,000--4,000 psig hydrogen partial pressure, and 10-90 lb/hr per ft[sup 3] catalyst space velocity. Partially hydrogenated material withdrawn from the first stage reaction zone is passed directly to the second stage catalytic reaction zone maintained at 760--860 F temperature for further hydrogenation and hydroconversion reactions. A 600--750 F[sup +] fraction containing 0--20 W % unreacted coal and ash solids is recycled to the coal slurrying step. If desired, the cut point lower boiling fraction can be further catalytically hydrotreated. By this process, the coal feed is successively catalytically hydrogenated and hydroconverted at selected conditions, to provide significantly increased yields of desirable low-boiling hydrocarbon liquid products and minimal production of hydrocarbon gases, and no net production of undesirable heavy oils and residuum materials. 2 figs.

MacArthur, J.B.; Comolli, A.G.; McLean, J.B.

1989-10-17T23:59:59.000Z

469

Modeling Fluid Flow and Electrical Resistivity in Fractured Geothermal Reservoir Rocks  

DOE Green Energy (OSTI)

Phase change of pore fluid (boiling/condensing) in rock cores under conditions representative of geothermal reservoirs results in alterations of the electrical resistivity of the samples. In fractured samples, phase change can result in resistivity changes that are more than an order of magnitude greater than those measured in intact samples. These results suggest that electrical resistivity monitoring may provide a useful tool for monitoring the movement of water and steam within fractured geothermal reservoirs. We measured the electrical resistivity of cores of welded tuff containing fractures of various geometries to investigate the resistivity contrast caused by active boiling and to determine the effects of variable fracture dimensions and surface area on water extraction. We then used the Nonisothermal Unsaturated Flow and Transport model (NUFT) (Nitao, 1998) to simulate the propagation of boiling fronts through the samples. The simulated saturation profiles combined with previously reported measurements of resistivity-saturation curves allow us to estimate the evolution of the sample resistivity as the boiling front propagates into the rock matrix. These simulations provide qualitative agreement with experimental measurements suggesting that our modeling approach may be used to estimate resistivity changes induced by boiling in more complex systems.

Detwiler, R L; Roberts, J J; Ralph, W; Bonner, B P

2003-01-14T23:59:59.000Z

470

Distillation of liquid fuels by thermogravimetry  

Science Conference Proceedings (OSTI)

In this paper, design and operation of a custom-built thermogravimetric apparatus for the distillation of liquid fuels are reported. Using a sensitive balance with scale of 0.001 g and ASTM distillation glassware, several petroleum and petroleum-derived samples have been analyzed by the thermogravimetric distillation method. When the ASTM distillation glassware is replaced by a micro-scale unit, sample size could be reduced from 100 g to 5-10 g. A computer program has been developed to transfer the data into a distillation plot, e.g. Weight Percent Distilled vs. Boiling Point. It also generates a report on the characteristic distillation parameters, such as, IBP (Initial Boiling Point), FBP (Final Boiling Point), and boiling point at 50 wt% distilled. Comparison of the boiling point distributions determined by TG (thermogravimetry) with those by SimDis GC (Simulated-Distillation Gas Chromatography) on two liquid fuel samples (i.e. a decanted oil and a filtered crude oil) are also discussed in this paper.

Huang, He; Wang, Keyu; Wang, Shaojie; Klein, M.T.; Calkins, W.H.

1996-12-31T23:59:59.000Z

471

Biomass and nutrient accumulation in young Prosopis Juliflora at Mombasa, Kenya  

SciTech Connect

Data are presented for 6-yr old P. juliflora, grown for quarry reclamation on: biomass of stems, large branches, small branches and leaves; height and volume of stems and large branches. All were calculated from regressions on based diameter. Volume was 209 cubic m/ha (stems), 75 cubic m/ha (large branches). Total biomass was 216 t/ha (77% in stems and large branches). Leaves plus small branches (22.6% of biomass) contained over 50% of the pool of nutrients N, P, K and Mg. Implications are discussed for site depletion as a result of total tree use for fuelwood and fodder. 25 references.

Maghembe, J.A.; Kariuki, E.M.; Haller, R.D.

1983-01-01T23:59:59.000Z

472

High speed, low power 100 MS/s front end track-and-hold amplifier for ten-bit pipelined ADC  

Science Conference Proceedings (OSTI)

The work focuses on the design of a high speed, low power track-and-hold amplifier (THA) for ten-bit 100 MS/s pipelined analogue-to-digital converter (ADC). A wide bandwidth and high gain two-stage ... Keywords: #, 47, CMFB, HPSA, MDAC, MHz, MS&, amplifier design, common-mode feedback, digital to analogue converters, high-performance systems architecture, hold amplifiers, mega samples per second, megahertz, multiplying DAC, nanometres, nm, operational transconductance amplifiers, peak, peak-to-, s, switched capacitors, track-and-

D. Meganathan; Raja Paul Perinbam; R. Deepalakshmi

2009-12-01T23:59:59.000Z

473

P?~P; Cambridge journal of undergraduate philosophy  

E-Print Network (OSTI)

method which enables us to test the alternative theories with controlled experimen ts and to choose those tha t work better, prOViding us with a reliable model of the order of the world. In order to do the same, however, human sciences would need to solve... I' .1••••.............. ·· Vi.·.·.··· .. fi Pv-f' EDITORIAL. , , , , , . P 1 WHAT IS WRONG WITH THE WAY PHILOSOPHY IS TAUGHT AT CAt1BRIDGE? , . P 2 THE POVERTY OF PHILOSOPHy , , , , , , . , .. , .. P 4 A FIRST ESSAY , , , , , . , , , P 6 OLD...

Anderson, Janet; Susijn, Laura

474

Midweek: Beyond the Headlines Volume 1, Number 3, 20-26 September 2006  

E-Print Network (OSTI)

t proposes to distribute as wel l as ensure tha t the government distributes the benef i t s to genuine beneficiaries and not on the basis of political allegiance. The par ty has a l so threatened to take legal action against the Indian Oil... cameraman with a leading national entertainment TV channel is doubtful. One of the many frauds Debashish played was of promising to get Scorpios which were supposed to be auctioned at New Delhi at a cheap rate, for people here. Nearly ten persons bought...

Zulca, Mita

475

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

476

U.S. DOE Office of Energy Efficiency and Renewable Energy Categorical Exclusion Determination Form  

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

Analysis and Measurement Services Corp (2) Analysis and Measurement Services Corp (2) Location: Knoxville, TN Project Title Online Monitoring Implementation in Boiling Water Reactors Proposed Action or Project Description American Recovery and Reinvestment Act: Analysis and Measurement Services Corporation proposes to develop and commercialize an on-line condition monitoring system for Boiling Water Reactors (BWRs). An online monitoring (OLM) system was developed in Phase I and Phase II projects for pressurized water reactors (PWRs). It helps measure the performance of PWR equipment and monitors the health of the plant. In Phase III, the system would be adapted to boiling water reactors (BWRs) through a research and development (R&D) effort involving analytical work, laboratory measurements, and in-plant demonstrations. From an OLM point of view,

477

CX-004148: Categorical Exclusion Determination | Department of Energy  

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

48: Categorical Exclusion Determination 48: Categorical Exclusion Determination CX-004148: Categorical Exclusion Determination Online Monitoring Implementation in Boiling Water Reactors CX(s) Applied: B3.6, B5.1 Date: 09/17/2010 Location(s): Knoxville, Tennessee Office(s): Energy Efficiency and Renewable Energy Analysis and Measurement Services Corporation proposes to develop and commercialize an on-line condition monitoring system for Boiling Water Reactors (BWRs). An online monitoring (OLM) system was developed in Phase I and Phase II projects for pressurized water reactors (PWRs). It helps measure the performance of PWR equipment and monitors the health of the plant. In Phase III, the system would be adapted to boiling water reactors (BWRs) through a research and development (R&D) effort involving analytical

478

Liquid-Gas Phase Transition of Supernova Matter and Its Relation to Nucleosynthesis  

E-Print Network (OSTI)

We investigate the liquid-gas phase transition of dense matter in supernova explosion by the relativistic mean field approach and fragment based statistical model. The boiling temperature is found to be high (T_{boil} >= 0.7 MeV for rho_B >= 10^{-7} fm^{-3}), and adiabatic paths are shown to go across the boundary of coexisting region even with high entropy. This suggests that materials experienced phase transition can be ejected to outside. We calculated fragment mass and isotope distribution around the boiling point. We found that heavy elements at the iron, the first, second, and third peaks of r-process are abundantly formed at rho_B = 10^{-7}, 10^{-5}, 10^{-3} and 10^{-2} fm^{-3}, respectively.

C. Ishizuka; A. Ohnishi; K. Sumiyoshi

2002-08-12T23:59:59.000Z

479

Methods of cracking a crude product to produce additional crude products  

DOE Patents (OSTI)

A method for producing a crude product is disclosed. Formation fluid is produced from a subsurface in situ heat treatment process. The formation fluid is separated to produce a liquid stream and a first gas stream. The first gas stream includes olefins. The liquid stream is fractionated to produce one or more crude products. At least one of the crude products has a boiling range distribution from 38.degree. C. and 343.degree. C. as determined by ASTM Method D5307. The crude product having the boiling range distribution from 38.degree. C. and 343.degree. C. is catalytically cracked to produce one or more additional crude products. At least one of the additional crude products is a second gas stream. The second gas stream has a boiling point of at most 38.degree. C. at 0.101 MPa.

Mo, Weijian (Sugar Land, TX); Roes, Augustinus Wilhelmus Maria (Houston, TX); Nair, Vijay (Katy, TX)

2009-09-08T23:59:59.000Z

480

Oligomerization process  

DOE Patents (OSTI)

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled. 2 figures.

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1991-03-26T23:59:59.000Z

Note: This page contains sample records for the topic "naph tha boiling" 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

Transetherification method  

DOE Patents (OSTI)

Transetherification is carried out in a catalytic distillation reactor, wherein the catalytic structure also serves as a distillation structure, by feeding a first ether to the catalyst bed to at least partially dissociate it into a first olefin and a first alcohol while concurrently therewith feeding either a second olefin (preferably a tertiary olefin) having a higher boiling point than said first olefin or a second alcohol having a higher boiling point than said first alcohol to the catalyst whereby either the second olefin and the first alcohol or the first olefin and the second alcohol react to form a second ether which has a higher boiling point than the first ether, which second ether is concurrently removed as a bottoms in the concurrent reaction-distillation to force that reaction to completion, while the unreacted first olefin or first alcohol is removed in the overhead. 1 fig.

Hearn, D.

1985-04-09T23:59:59.000Z

482

Transetherification method  

DOE Patents (OSTI)

Transetherification is carried out in a catalytic distillation reactor, wherein the catalytic structure also serves as a distillation structure, by feeding a first ether to the catalyst bed to at least partially dissociate it into a first olefin and a first alcohol while concurrently therewith feeding either a second olefin (preferably a tertiary olefin) having a higher boiling point than said first olefin or a second alcohol having a higher boiling point than said first alcohol to the catalyst whereby either the second olefin and the first alcohol or the first olefin and the second alcohol react to form a second ether which has a higher boiling point than the first ether, which second ether is concurrently removed as a bottoms in the concurrent reaction-distillation to force that reaction to completion, while the unreacted first olefin or first alcohol is removed in the overhead.

Hearn, Dennis (Houston, TX)

1985-01-01T23:59:59.000Z

483

Method for conducting exothermic reactions  

DOE Patents (OSTI)

A liquid phase process for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Smith, L. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-01-05T23:59:59.000Z

484

Reactor for exothermic reactions  

DOE Patents (OSTI)

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02T23:59:59.000Z

485

8. Innovative Technologies: Two-Phase Heat Transfer in Water-Based Nanofluids for Nuclear Applications Final Report  

SciTech Connect

Abstract Nanofluids are colloidal dispersions of nanoparticles in water. Many studies have reported very significant enhancement (up to 200%) of the Critical Heat Flux (CHF) in pool boiling of nanofluids (You et al. 2003, Vassallo et al. 2004, Bang and Chang 2005, Kim et al. 2006, Kim et al. 2007). These observations have generated considerable interest in nanofluids as potential coolants for more compact and efficient thermal management systems. Potential Light Water Reactor applications include the primary coolant, safety systems and severe accident management strategies, as reported in other papers (Buongiorno et al. 2008 and 2009). However, the situation of interest in reactor applications is often flow boiling, for which no nanofluid data have been reported so far. In this project we investigated the potential of nanofluids to enhance CHF in flow boiling. Subcooled flow boiling heat transfer and CHF experiments were performed with low concentrations of alumina, zinc oxide, and diamond nanoparticles in water (? 0.1 % by volume) at atmospheric pressure. It was found that for comparable test conditions the values of the nanofluid and water heat transfer coefficient (HTC) are similar (within ?20%). The HTC increased with mass flux and heat flux for water and nanofluids alike, as expected in flow boiling. The CHF tests were conducted at 0.1 MPa and at three different mass fluxes (1500, 2000, 2500 kg/m2s) under subcooled conditions. The maximum CHF enhancement was 53%, 53% and 38% for alumina, zinc oxide and diamond, respectively, always obtained at the highest mass flux. A post-mortem analysis of the boiling surface reveals that its morphology is altered by deposition of the particles during nanofluids boiling. A confocal-microscopy-based examination of the test section revealed that nanoparticles deposition not only changes the number of micro-cavities on the surface, but also the surface wettability. A simple model was used to estimate the ensuing nucleation site density changes, but no definitive correlation between the nucleation site density and the heat transfer coefficient data could be found. Wettability of the surface was substantially increased for heater coupons boiled in alumina and zinc oxide nanofluids, and such wettability increase seems to correlate reasonably well with the observed marked CHF enhancement for the respective nanofluids. Interpretation of the experimental data was conducted in light of the governing surface parameters (surface area, contact angle, roughness, thermal conductivity) and existing models. It was found that no single parameter could explain the observed HTC or CHF phenomena.

Buongiorno, Jacopo; Hu, Lin-wen

2009-07-31T23:59:59.000Z

486

Superconducting magnet cooling system  

DOE Patents (OSTI)

A device is provided for cooling a conductor to the superconducting state. The conductor is positioned within an inner conduit through which is flowing a supercooled liquid coolant in physical contact with the conductor. The inner conduit is positioned within an outer conduit so that an annular open space is formed therebetween. Through the annular space is flowing coolant in the boiling liquid state. Heat generated by the conductor is transferred by convection within the supercooled liquid coolant to the inner wall of the inner conduit and then is removed by the boiling liquid coolant, making the heat removal from the conductor relatively independent of conductor length.

Vander Arend, Peter C. (Center Valley, PA); Fowler, William B. (St. Charles, IL)

1977-01-01T23:59:59.000Z

487

Thermal-hydraulic aspects of flow inversion in a research reactor  

SciTech Connect

PARET, a neutronics and thermal-hydraulics computer code, has been modified to account for natural convection in a reactor core. The code was then used to analyze the flow inversion that occurs in a reactor with heat removal by forced convection in the downward direction after a pump failure. Typical results are shown for a number of parameters. Research reactors normally operating much above ten MW are predicted to experience nucleate boiling in the event of a flow inversion. Comparison with experimental results from the Belgian BR2 reactor indicated general agreement although nucleate boiling that was analytically predicted was not noted in the BR2 data.

Smith, R.S.; Woodruff, W.L.

1986-11-01T23:59:59.000Z

488

Photographic study of the mechanism of heat transfer enhancement by electrolytic hydrogen gas  

SciTech Connect

A mechanism of promoting heat transfer, by which a remarkably high heat flux is obtained with a heat source having a small temperature difference is elucidated. The method consists of generating a small amount of electrolytic hydrogen gas from a heating surface undergoing nucleate boiling and natural convection. Photographs of a boiling process in the presence of electrolytic hydrogen gas evolution from the heating surface were taken. By analyzing high-speed motion pictures it is shown that the electrolytic hydrogen gas permits vapor bubble production with a small degree of superheat and increases the number of vapor bubble nuclei.

Nakayama, A.; Kano, M.

1983-06-01T23:59:59.000Z

489

GE PowerPoint Template  

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

Steels for Steels for Accident Tolera nt Fuel Cla ddings Ferritic Ma rtensitic Alloys a s Accident Tolera nt Fuel (ATF) Cla dding Ma teria l for Light Wa ter Rea ctors Ra ul B. Reba k, GE Globa l Resea rch DOE Integra tion Meeting, Sa lt La ke City 27-August-2013 DE NE 568 2 / GE Reba k - DOE Integra tion Meeting, Sa lt La ke City, 27-August-2013/ GE Project Tea m 3 / GE Reba k - DOE Integra tion Meeting, Sa lt La ke City, 27-August-2013/ Approa ch of GE Resea rch Proposa l * Demonstra te tha t sta inless iron ba sed bulk a lloys or Adva nced Steels ca n be used a s fuel cla dding ma teria ls in commercia l nuclea r rea ctors * The proposed ma teria l should be a s good a s Zr a lloys (or better tha n Zr a lloys) under norma l opera tion conditions 1. Resista nt to genera l corrosion a nd environmenta l cra

490

Measurement and interpretation of threshold stress intensity factors for steels in high-pressure hydrogen gas.  

DOE Green Energy (OSTI)

Threshold stress intensity factors were measured in high-pressure hydrogen gas for a variety of low alloy ferritic steels using both constant crack opening displacement and rising crack opening displacement procedures. The sustained load cracking procedures are generally consistent with those in ASME Article KD-10 of Section VIII Division 3 of the Boiler and Pressure Vessel Code, which was recently published to guide design of high-pressure hydrogen vessels. Three definitions of threshold were established for the two test methods: K{sub THi}* is the maximum applied stress intensity factor for which no crack extension was observed under constant displacement; K{sub THa} is the stress intensity factor at the arrest position for a crack that extended under constant displacement; and K{sub JH} is the stress intensity factor at the onset of crack extension under rising displacement. The apparent crack initiation threshold under constant displacement, K{sub THi}*, and the crack arrest threshold, K{sub THa}, were both found to be non-conservative due to the hydrogen exposure and crack-tip deformation histories associated with typical procedures for sustained-load cracking tests under constant displacement. In contrast, K{sub JH}, which is measured under concurrent rising displacement and hydrogen gas exposure, provides a more conservative hydrogen-assisted fracture threshold that is relevant to structural components in which sub-critical crack extension is driven by internal hydrogen gas pressure.

Nibur, Kevin A.

2010-11-01T23:59:59.000Z

491

HEAVY-WATER-MODERATED POWER REACTORS ENGINEERING AND ECONOMIC EVALUATIONS. VOLUME I. SUMMARY REPORT  

SciTech Connect

Capital investments and the cost of power were estimated for 21 heavy- water-moderated, natural-uraniumfueled power-reactor plants, ranging in capacity from 100 to 460 Mw(e). Comparisons were made of hot- and coldmoderator reactors and of the relative merits of pressuretube and pressure-vessel designs. Reactors cooled with liquid D/sub 2/O, boiling D/sub 2/O, /sub 2/O steam, and helium were evalunted. A cold-moderator pressure-tube reactor cooled with boiling D/sub 2/O shows the most economic promise of the D/sub 2/Omoderated reactor systems studied to date. Reactors of this type have sufficient reactivity to permit satisfactory fuel exposures, but the development of additional technology is a prerequisite for optimum designs. At capacities of 300 and 400 Mw(e), the estimated power costs from the current designs of boiling-D/sub 2/O pressure-tabe reactor plants are 11.3 and 9.8 mills/kwh, respectively. From liquid-D/sub 2/-cooled concepts of comparable capacities the indicated power costs are 7 to 20% higher. With an active development program, a power cost of 8.0 to 8.5 mills/kwh may be attained in a 300 Mw(e) boiling-D/sub 2/O reactor plant within the next decade. (auth)

1960-06-01T23:59:59.000Z

492

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 ...

493

REMOVAL OF EBWR FUEL ELEMENT SCALE BY SLURRY HONING  

SciTech Connect

The scale deposit on the Experimental Boiling Water Reactor fuel plates can be removed by slurry honing the plates with an abrasive-water mixture. Problems inherent in any production operation of this type are discussed. Areas of continued investigation of the method are suggested. (auth)

Charak, I.

1960-09-01T23:59:59.000Z

494

Nuclear Maintenance Applications Center: Reactor Recirculation Pump Seal Maintenance Guide  

Science Conference Proceedings (OSTI)

In 1993, the Electric Power Research Institute (EPRI) published Main Coolant Pump Seal Maintenance Guide (report TR-100855). That report provides guidance on maintenance and troubleshooting of main coolant pump seals but was generic in naturethe information was relevant to pressurized water reactors (PWRs) and boiling water reactors (BWRs).

2008-12-18T23:59:59.000Z

495

Managed by UT-Battelle for the Department of Energy  

E-Print Network (OSTI)

the recovered Pu from spent uranium fuel with thorium instead of depleted uranium (DU), i.e. using Th þ RGPu-grade plutonium and depleted uranium (MOX). Hence, it can be expected that the introduction of Th þ RGPu fuel of life BWR Boiling-water reactor CVC Coolant void coefficient CZP Cold zero power DU Depleted uranium EOL

Pennycook, Steve

496

Appendix I. List of Acronyms and Abbreviations, and Glossary of Terms  

E-Print Network (OSTI)

the recovered Pu from spent uranium fuel with thorium instead of depleted uranium (DU), i.e. using Th þ RGPu-grade plutonium and depleted uranium (MOX). Hence, it can be expected that the introduction of Th þ RGPu fuel of life BWR Boiling-water reactor CVC Coolant void coefficient CZP Cold zero power DU Depleted uranium EOL

497

Program on Technology Innovation: Prediction and Evaluation of Environmentally Assisted Cracking in LWR Structural Materials  

Science Conference Proceedings (OSTI)

This report describes the final results of Phase III of a three-year joint research program sponsored by EPRI in collaboration with Fracture & Reliability Research Institute (FRRI) at Tohoku University, Japanese utilities, vendors, and international organizations. The program addressed environmentally assisted cracking (EAC) of light water reactor (LWR) structural materials in pressurized water reactor (PWR) and boiling water reactor (BWR) environments.

2007-11-20T23:59:59.000Z

498

NERC GUIDANCE ON SAFE USE OF CRYOGENICS Version 1.7 Date: December 2012  

E-Print Network (OSTI)

be necessary when using, storing, and transporting low temperature liquefied or solidified gases (commonly produced from a gas that can be liquefied, and in some cases solidified, by the application of pressure within a research environment, with associated boiling points at atmospheric pressure, are Nitrogen (-196

Edinburgh, University of

499

Cryogenic Safety This course will provide basic information concerning cryogens and  

E-Print Network (OSTI)

handles two cryogenic liquids, helium and nitrogen. Liquefied nitrogen has a boiling point, at atmospheric, thermal stress, air condensation, and cold embrittlement. n Identify liquid nitrogen and helium and incidents. #12;What is a cryogen? A cryogen is an extremely cold element or compound. Cryogens are liquefied

Weston, Ken

500

Methyl Bromide o Bromomethane, monobromomethane, isobrome, Brom-o-Gas, Bromomethane, Celume,  

E-Print Network (OSTI)

bromide produced in the U.S. goes into pesticidal formulations (as of 1996) Total use of 711,175 lb in 2009, 78% on imported and 22% on exported material under Plant Protection and Quarantine oversight of exports requiring MB fumigation in 2005-2009 $2.2 billion/year o Methylating solvent, low-boiling solvent

Toohey, Darin W.