National Library of Energy BETA

Sample records for naph tha boiling

  1. BOILING REACTORS

    DOE Patents [OSTI]

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  2. Geysering in boiling channels

    SciTech Connect (OSTI)

    Aritomi, Masanori; Takemoto, Takatoshi; Chiang, Jing-Hsien

    1995-09-01

    A concept of natural circulation BWRs such as the SBWR has been proposed and seems to be promising in that the primary cooling system can be simplified. The authors have been investigating thermo-hydraulic instabilities which may appear during the start-up in natural circulation BWRs. In our previous works, geysering was investigated in parallel boiling channels for both natural and forced circulations, and its driving mechanism and the effect of system pressure on geysering occurrence were made clear. In this paper, geysering is investigated in a vertical column and a U-shaped vertical column heated in the lower parts. It is clarified from the results that the occurrence mechanism of geysering and the dependence of system pressure on geysering occurrence coincide between parallel boiling channels in circulation systems and vertical columns in non-circulation systems.

  3. CHIMNEY FOR BOILING WATER REACTOR

    DOE Patents [OSTI]

    Petrick, M.

    1961-08-01

    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)

  4. Validation Data Plan Implementation: Subcooled Flow Boiling

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Validation Data Plan Implementation: Subcooled Flow Boiling Case Study Anh Bui and Nam Dinh Idaho National Laboratory September 28, 2012 CASL-U-2012-0162-000 CASL-U-2012-0162-000 Subcooled Flow Boiling Case Study INL-MIS-12-27303 1 INL/MIS-12-27303 September 2012 Validation Data Plan Implementation: Subcooled Flow Boiling Case Study Anh Bui Nam Dinh CASL-U-2012-0162-000 Subcooled Flow Boiling Case Study INL-MIS-12-27303 2 ABSTRACT This report presents a step forward in the development and

  5. Lattice Boltzmann modeling of boiling heat transfer: The boiling curve and the effects of wettability

    DOE PAGES-Beta [OSTI]

    Li, Q.; Kang, Q. J.; Francois, M. M.; He, Y. L.; Luo, K. H.

    2015-03-03

    A hybrid thermal lattice Boltzmann (LB) model is presented to simulate thermal multiphase flows with phase change based on an improved pseudopotential LB approach (Li et al., 2013). The present model does not suffer from the spurious term caused by the forcing-term effect, which was encountered in some previous thermal LB models for liquid–vapor phase change. Using the model, the liquid–vapor boiling process is simulated. The boiling curve together with the three boiling stages (nucleate boiling, transition boiling, and film boiling) is numerically reproduced in the LB community for the first time. The numerical results show that the basic featuresmore » and the fundamental characteristics of boiling heat transfer are well captured, such as the severe fluctuation of transient heat flux in the transition boiling and the feature that the maximum heat transfer coefficient lies at a lower wall superheat than that of the maximum heat flux. Moreover, the effects of the heating surface wettability on boiling heat transfer are investigated. It is found that an increase in contact angle promotes the onset of boiling but reduces the critical heat flux, and makes the boiling process enter into the film boiling regime at a lower wall superheat, which is consistent with the findings from experimental studies.« less

  6. Lattice Boltzmann modeling of boiling heat transfer: The boiling curve and the effects of wettability

    SciTech Connect (OSTI)

    Li, Q.; Kang, Q. J.; Francois, M. M.; He, Y. L.; Luo, K. H.

    2015-03-03

    A hybrid thermal lattice Boltzmann (LB) model is presented to simulate thermal multiphase flows with phase change based on an improved pseudopotential LB approach (Li et al., 2013). The present model does not suffer from the spurious term caused by the forcing-term effect, which was encountered in some previous thermal LB models for liquid–vapor phase change. Using the model, the liquid–vapor boiling process is simulated. The boiling curve together with the three boiling stages (nucleate boiling, transition boiling, and film boiling) is numerically reproduced in the LB community for the first time. The numerical results show that the basic features and the fundamental characteristics of boiling heat transfer are well captured, such as the severe fluctuation of transient heat flux in the transition boiling and the feature that the maximum heat transfer coefficient lies at a lower wall superheat than that of the maximum heat flux. Moreover, the effects of the heating surface wettability on boiling heat transfer are investigated. It is found that an increase in contact angle promotes the onset of boiling but reduces the critical heat flux, and makes the boiling process enter into the film boiling regime at a lower wall superheat, which is consistent with the findings from experimental studies.

  7. Experimental Investigation of Subcooled Flow Boiling

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    8 Experimental Investigation of Subcooled Flow Boiling Yassin A. Hassan TAMU September 30, 2013 CASL-8-2013-0214-000 TEXAS A&M UNIVERSITY Experimental Investigation of Subcooled Flow Boiling Milestone Report PI: Yassin A. Hassan 9/30/2013 CASL-U-2013-0214-000 Contents Introduction ....................................................................................................................................................... 5 Experimental Setup

  8. Boiling water reactor-full length emergency core cooling heat...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Boiling water reactor-full length emergency core cooling heat transfer ... Citation Details In-Document Search Title: Boiling water reactor-full length emergency ...

  9. 2010 Inspection and Status Report for the Boiling Nuclear Superheater...

    Office of Legacy Management (LM)

    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...

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

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    of Power Electronics of Electric Vehicles with Small Channel Coolant Boiling Cooling Boiling in Head Region - PACCAR Integrated Underhood Thermal and External Aerodynamics- Cummins

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

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Circular Tube Efficient Cooling in Engines with Nucleated Boiling Cooling Boiling in Head Region - PACCAR Integrated Underhood Thermal and External Aerodynamics- Cummins

  12. Preliminary design study of small long life boiling water reactor...

    Office of Scientific and Technical Information (OSTI)

    boiling water reactor (BWR) with tight lattice thorium nitride fuel Citation Details In-Document Search Title: Preliminary design study of small long life boiling water reactor ...

  13. Great Boiling Springs Geothermal Area | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Great Boiling Springs Geothermal Area Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Great Boiling Springs Geothermal Area Contents 1 Area Overview 2 History and...

  14. Modeling acid-gas generation from boiling chloride brines (Journal...

    Office of Scientific and Technical Information (OSTI)

    Modeling acid-gas generation from boiling chloride brines Citation Details In-Document Search Title: Modeling acid-gas generation from boiling chloride brines This study ...

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

    SciTech Connect (OSTI)

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

    2006-12-01

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

  16. SUPERHEATING IN A BOILING WATER REACTOR

    DOE Patents [OSTI]

    Treshow, M.

    1960-05-31

    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.

  17. CONTINUOUS ANALYZER UTILIZING BOILING POINT DETERMINATION

    DOE Patents [OSTI]

    Pappas, W.S.

    1963-03-19

    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)

  18. Subcooled Flow Boiling Heat Transfer to Water and Ethylene Glycol...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Subcooled Flow Boiling Heat Transfer to Water and Ethylene GlycolWater Mixtures in a Bottom-Heated Tube Title Subcooled Flow Boiling Heat Transfer to Water and Ethylene Glycol...

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

    ScienceCinema (OSTI)

    None

    2012-12-31

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

  20. PACCAR CRADA: Experimental Investigation in Coolant Boiling in a

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Half-Heated Circular Tube | Department of Energy PACCAR CRADA: Experimental Investigation in Coolant Boiling in a Half-Heated Circular Tube PACCAR CRADA: Experimental Investigation in Coolant Boiling in a Half-Heated Circular Tube 2012 DOE Hydrogen and Fuel Cells Program and Vehicle Technologies Program Annual Merit Review and Peer Evaluation Meeting vss079_yu_2012_o.pdf (585.51 KB) More Documents & Publications CRADA with PACCAR Experimental Investigation in Coolant Boiling in a

  1. Enhanced Pool-Boiling Heat Transfer Using Nanostructured Surfaces...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    creates optimal surface wettability characteristics that allow better capillary flow of water on the liquid boiling surfaces often used to cool electronics. the dense...

  2. Geothermal Technology Breakthrough in Alaska: Harvesting Heat below Boiling Temperatures

    Energy.gov [DOE]

    The Energy Department is supporting geothermal exploration at lower temperatures, thanks to a technology breakthrough that allows geothermal energy to be produced at temperatures below the boiling...

  3. (Boiling water reactor (BWR) CORA experiments)

    SciTech Connect (OSTI)

    Ott, L.J.

    1990-10-16

    To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of the BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.

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

    DOE Patents [OSTI]

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

    2000-01-01

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

  5. L3:THM.CLS.P7.09 Advancements on Wall Boiling Modeling in CFD...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    ... Nanofluid Pool Boiling Phenomena," Nanoscale Research Letters, vol. 6, no. 232, 2011. 2 J. Garnier, E. Manon and G. Cubizolles, "Local Measurements on Flow Boiling of Refrigerant ...

  6. Flow Boiling Carolyn Coyle, Jacopo Buongiorno, Thomas McKrell

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    ... Systems I: Thermal Hydraulic Fundamentals, Taylor & Francis 1993. 13. ... Nuclear Engineering and Design 241.3 (2011): 792-98. 17. Pan, C., Jones, B.G., "Wick Boiling ...

  7. Correlations estimate volume distilled using gravity, boiling point

    SciTech Connect (OSTI)

    Moreno, A.; Consuelo Perez de Alba, M. del; Manriquez, L.; Guardia Mendoz, P. de la

    1995-10-23

    Mathematical nd graphic correlations have been developed for estimating cumulative volume distilled as a function of crude API gravity and true boiling point (TBP). The correlations can be used for crudes with gravities of 21--34{degree} API and boiling points of 150--540 C. In distillation predictions for several mexican and Iraqi crude oils, the correlations have exhibited accuracy comparable to that of laboratory measurements. The paper discusses the need for such a correlation and the testing of the correlation.

  8. CASL-U-2015-0040-000 Initial Boiling Water

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    40-000 Initial Boiling Water Reactor (BWR) Input Specifications Scott Palmtag Core Physics February 28, 2015 Initial Boiling Water Reactor (BWR) Input Specification Consortium for Advanced Simulation of LWRs ii CASL-U-2015-0040-000 REVISION LOG Revision Date Affected Pages Revision Description 0 02/28/2015 All Original Report for L3:PHI.VCS.P10.02 Document pages that are: Export Controlled NO IP/Proprietary/NDA Controlled NO Sensitive Controlled NO Requested Distribution: To: Copy: Initial

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

    DOE Patents [OSTI]

    White, Terry L.

    1991-01-01

    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.

  10. Modeling acid-gas generation from boiling chloride brines

    SciTech Connect (OSTI)

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

    2009-11-16

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

  11. Efficiency of a solar collector with internal boiling

    SciTech Connect (OSTI)

    Neeper, D.A.

    1986-01-01

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

  12. SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

    DOE Patents [OSTI]

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

    1960-08-16

    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.

  13. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    DOE Patents [OSTI]

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

    1962-08-14

    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)

  14. Union job fight boiling at DOE cleanup sites

    SciTech Connect (OSTI)

    Setzer, S.W.

    1993-11-15

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

  15. EERE Success Story—Geothermal Technology Breakthrough in Alaska: Harvesting Heat below Boiling Temperatures

    Energy.gov [DOE]

    The Energy Department is supporting geothermal exploration at lower temperatures, thanks to a technology breakthrough that allows geothermal energy to be produced at temperatures below the boiling...

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

    DOE Patents [OSTI]

    Gluntz, Douglas M.

    1996-01-01

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

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

    DOE Patents [OSTI]

    Gluntz, D.M.

    1996-03-12

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

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

    DOE Patents [OSTI]

    Forsberg, C.W.

    1985-02-19

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

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

    DOE Patents [OSTI]

    Forsberg, Charles W.

    1987-01-01

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

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

    SciTech Connect (OSTI)

    PetrusTakaki, N.

    2012-07-01

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

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

    SciTech Connect (OSTI)

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

    1980-07-01

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

  2. BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

    DOE Patents [OSTI]

    Treshow, M.

    1963-04-30

    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)

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

    SciTech Connect (OSTI)

    Not Available

    1992-11-01

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

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

    SciTech Connect (OSTI)

    Not Available

    1992-11-01

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

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

    SciTech Connect (OSTI)

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

    1986-09-01

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

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

    SciTech Connect (OSTI)

    Chen, G. F.; Gong, M. Q.; Wu, J. F.; Zou, X.; Wang, S.

    2014-01-29

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

  7. Multi-cycle boiling water reactor fuel cycle optimization

    SciTech Connect (OSTI)

    Ottinger, K.; Maldonado, G.I.

    2013-07-01

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

  8. Camera Inspection Arm for Boiling Water Reactors - 13330

    SciTech Connect (OSTI)

    Martin, Scott; Rood, Marc

    2013-07-01

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

  9. Pebble Bed Boiling Water Reactor Concept With Superheated Steam

    SciTech Connect (OSTI)

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

    2002-07-01

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

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

    SciTech Connect (OSTI)

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    1995-03-01

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

  11. Boiling-Water Reactor internals aging degradation study. Phase 1

    SciTech Connect (OSTI)

    Luk, K.H.

    1993-09-01

    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.

  12. Development of 1000 MWe Advanced Boiling Water Reactor

    SciTech Connect (OSTI)

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

    2006-07-01

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

  13. An experimental study of pool boiling heat transfer in reduced gravity

    SciTech Connect (OSTI)

    Shatto, D.P.; Renzi, K.I.; Peterson, G.P.; Morris, T.K.; Aaron, J.W.

    1996-12-31

    Experiments were performed in which pool boiling of pure water at reduced pressures was observed for behavior of the critical heatflux (CHF) and nucleate boiling heat transfer coefficients in a reduced gravitational environment. The experiments took place while alternating between microgravity and g/g{sub o} = 1.8 during parabolic flights aboard the NASA 930 (KC-135A). Heat transfer data were also obtained at Martian gravity levels (g/g{sub o} = 1/3). Parts of the test chamber were constructed of transparent materials to allow viewing and recording of the various boiling regimes encountered during the experiments. Results indicate that the onset of nucleate boiling occurred at lower heat fluxes in reduced gravity, resulting in higher two-phase heat transfer coefficients for g/g{sub o} < 1 than for g/g{sub o} = 1.8. In addition, the results indicate a significant reduction in the critical heat flux under reduced gravity conditions.

  14. Enhancement of pool boiling from a vertical rod using guide disks

    SciTech Connect (OSTI)

    Whitehouse, J.C.

    1992-11-01

    This report provides experimental and theoretical investigation of the boiling process which used a system of evenly spaced disks to constrain the path of bubbles from point origin to point of collapse. The experiments identified five distinct heat-transfer regimes, two of which (flange and strobe) are unique to this geometry and cannot be explained by conventional heat-transfer correlations. Bubble and wave models developed for flange and strobe boiling, respectively, predict these phenomena with reasonable success.

  15. Enhancement of pool boiling from a vertical rod using guide disks

    SciTech Connect (OSTI)

    Whitehouse, J.C.

    1992-01-01

    This report provides experimental and theoretical investigation of the boiling process which used a system of evenly spaced disks to constrain the path of bubbles from point origin to point of collapse. The experiments identified five distinct heat-transfer regimes, two of which (flange and strobe) are unique to this geometry and cannot be explained by conventional heat-transfer correlations. Bubble and wave models developed for flange and strobe boiling, respectively, predict these phenomena with reasonable success.

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

    SciTech Connect (OSTI)

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

    1992-04-01

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

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

    SciTech Connect (OSTI)

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

    1992-04-01

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

  18. CASL - Initial Modeling and Analysis of the Departure from Nucleate Boiling

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Challenge Problem Initial Modeling and Analysis of the Departure from Nucleate Boiling Challenge Problem Yixing Sung, Jin Yan, Zeses E. Karoutas of Westinghouse Electric Company LLC Anh V. Bui, Hongbin Zhang of Idaho National Laboratories Nam Dinh of North Carolina State University Departure from Nucleate Boiling (DNB) is one of the safety-related Challenge Problems (CP) that CASL is addressing in support of Pressurized Water Reactor (PWR) power uprate, high fuel burnup and plant lifetime

  19. Passive gamma analysis of the boiling-water-reactor assemblies

    DOE PAGES-Beta [OSTI]

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; et al

    2016-06-04

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in themore » past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.« less

  20. Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report

    SciTech Connect (OSTI)

    Aksan, S.N.; Stierli, F.; Analytis, G.T.

    1992-03-01

    The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 seconds earlier than rods with embedded thermocouples. Additionally, the external surface-thermocouples give readings up to 20 K lower than those obtained with internal surface thermocouples (in the absence of external thermocouples) in the peak cladding temperature zone. Some of the boil-off data obtained from the NEPTUN test facility are used for the assessment of the thermal-hydraulic transient computer codes. These calculations were performed extensively using the frozen version of TRAC-BD1/MOD1 (version 22). A limited number of assessment calculations were done with RELAP5/MOD2 (version 36.02). In this report the main results and conclusions of these calculations are presented with the identification of problem areas in relation to models relevant to boil-off phenomena. On the basis of further analysis and calculations done, changing some of the models such as the bubbly/slug flow interfacial friction correlation which eliminate some of the problems are recommended.

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

    SciTech Connect (OSTI)

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

    2002-07-01

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

  2. Natural Convection and Boiling for Cooling SRP Reactors During Loss of Circulation Conditions

    SciTech Connect (OSTI)

    Buckner, M.R.

    2001-06-26

    This study investigated natural convection and boiling as a means of cooling SRP reactors in the event of a loss of circulation accident. These studies show that single phase natural convection cooling of SRP reactors in shutdown conditions with the present piping geometry is probably not feasible.

  3. Explosive boiling of metals upon irradiation by a nanosecond laser pulse

    SciTech Connect (OSTI)

    Mazhukin, V I; Demin, M M; Shapranov, A V; Samokhin, A A

    2014-04-28

    A repeated effect of explosive boiling has been found in metals exposed to a nanosecond laser pulse in the framework of molecular dynamic simulations combined with a continuum description of a conduction band electrons system. This effect can be used, in particular, as a marker of approaching critical parameters of the region in the irradiated matter. (letters)

  4. Modeling the onset of flow instability for subcooled boiling in downflow

    SciTech Connect (OSTI)

    Qureshi, Z. ); Barry, J.J.; Crowley, C.J. )

    1990-01-01

    A postulated loss-of-coolant accident (LOCA) scenario for the Savannah River Plant (SRP) production reactors involves a double-ended break of a reactor primary coolant pipe. The flow of coolant (D{sub 2}O) in the reactor may decrease in such an event. As the flow into the reactor decreases, boiling may occur, followed by dryout and failure of the fuel due to overheating. A typical SRP fuel assembly consists of multiple concentric tubes containing the fuel and target materials. Coolant passes through the annular passages in the assembly in downflow. Under normal operating conditions, the flow rate is maintained high enough to suppress or minimize subcooled boiling, i.e. the flow remains essentially single phase throughout. At high coolant flow rates, the flow is single phase or partially developed subcooled boiling, and the pressure drop decreases with decreasing flow rate. Here friction dominates the pressure gradient, and the flow is stable. Below a certain flow rate, however, pressure drop may increase with decreasing flow rate. This occurs when significant voids are produced by boiling, resulting in a large acceleration component to the pressure drop. The negative slope of the curve leads to an instability because the pressure drop cannot adjust to compensate -- the flow is driven to a lower value. Overheating of the channel may result. 15 refs., 14 figs.

  5. Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors

    SciTech Connect (OSTI)

    Rosa, M.P.; Podowski, M.Z.

    1995-09-01

    This paper is concerned with the analysis of dynamics and stability of boiling channels and systems. The specific objectives are two-fold. One of them is to present the results of a study aimed at analyzing the effects of various modeling concepts and numerical approaches on the transient response and stability of parallel boiling channels. The other objective is to investigate the effect of closed-loop feedback on stability of a boiling water reactor (BWR). Various modeling and computational issues for parallel boiling channels are discussed, such as: the impact of the numerical discretization scheme for the node containing the moving boiling boundary on the convergence and accuracy of computations, and the effects of subcooled boiling and other two-phase flow phenomena on the predictions of marginal stability conditions. Furthermore, the effects are analyzed of local loss coefficients around the recirculation loop of a boiling water reactor on stability of the reactor system. An apparent paradox is explained concerning the impact of changing single-phase losses on loop stability. The calculations have been performed using the DYNOBOSS computer code. The results of DYNOBOSS validation against other computer codes and experimental data are shown.

  6. ASTRID: A 3D Eulerian software for subcooled boiling modelling - comparison with experimental results in tubes and annuli

    SciTech Connect (OSTI)

    Briere, E.; Larrauri, D.; Olive, J.

    1995-09-01

    For about four years, Electricite de France has been developing a 3-D computer code for the Eulerian simulation of two-phase flows. This code, named ASTRID, is based on the six-equation two-fluid model. Boiling water flows, such as those encountered in nuclear reactors, are among the main applications of ASTRID. In order to provide ASTRID with closure laws and boundary conditions suitable for boiling flows, a boiling model has been developed by EDF and the Institut de Mecanique des Fluides de Toulouse. In the fluid, the heat and mass transfer between a bubble and the liquid is being modelled. At the heating wall, the incipient boiling point is determined according to Hsu`s criterion and the boiling heat flux is split into three additive terms: a convective term, a quenching term and a vaporisation term. This model uses several correlations. EDF`s program in boiling two-phase flows also includes experimental studies, some of which are performed in collaboration with other laboratories. Refrigerant subcooled boiling both in tubular (DEBORA experiment, CEN Grenoble) and in annular geometry (Arizona State University Experiment) have been computed with ASTRID. The simulations show the satisfactory results already obtained on void fraction and liquid temperature. Ways of improvement of the model are drawn especially on the dynamical part.

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

    SciTech Connect (OSTI)

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

    1981-04-01

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

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

    SciTech Connect (OSTI)

    Yu, W.; France, D. M.; Routbort, J. L.

    2011-01-19

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

  9. Explosive boiling of a metallic glass superheated by nanosecond pulse laser ablation

    SciTech Connect (OSTI)

    Jiang, M. Q. E-mail: lhdai@lnm.imech.ac.cn; Wei, Y. P.; Wilde, G.; Dai, L. H. E-mail: lhdai@lnm.imech.ac.cn

    2015-01-12

    We report an explosive boiling in a Zr-based (Vitreloy 1) bulk metallic glass irradiated by a nanosecond pulse laser with a single shot. This critical phenomenon is accompanied by the ejection of high-temperature matter from the target and the formation of a liquid-gas spinodal pattern on the irradiated area. An analytical model reveals that the glassy target experiences the normal heating (melting) and significant superheating, eventually culminating in explosive boiling near the spinodal limit. Furthermore, the time lag of nucleation and the critical radius of vapor bubbles are theoretically predicted, which are in agreement with the experimental observations. This study provides the investigation on the instability of a metallic glass liquid near the thermodynamic critical temperature.

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

    SciTech Connect (OSTI)

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

    2011-09-02

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

  11. Boiling Water Reactor Fuel Cycle Optimization for Prevention of Channel-Blade Interference

    SciTech Connect (OSTI)

    Kropaczek, David J.; Karve, Atul A.; Oyarzun, Christian C.; Asgari, Mehdi

    2006-07-01

    A formal optimization method for eliminating the potential of Boiling Water Reactor channel-blade interference is presented within the context of fuel cycle design. The method is based on the use of threshold constraints on blade force as penalty terms within an objective function that are employed as part of a search algorithm. Results demonstrate the effectiveness of the constraint formulation in eliminating channel-blade interference as part of the design of the core loading and operational strategy. (authors)

  12. Experiment and RELAP5 Analysis for the Downcomer Boiling of APR1400 under LBLOCA

    SciTech Connect (OSTI)

    Dong Won Lee; Hee Cheon No; Eu Hwak Lee; Seung Jong Oh; Chul-Hwa Song

    2004-07-01

    The direct vessel injection (DVI) mode of a safety injection system is adopted instead of a conventional cold leg injection (CLI) mode as one of the advanced design features of the APR1400 (Advanced Power Reactor 1400 MW). From the calculation results of RELAP5 with full plant, it is found out that the sudden boiling happens in the downcomer due to heat transfer from the reactor vessel wall and it can affect the reactor safety. In the present study, experimental tests are carried out to observe the actual boiling phenomena in the downcomer and to validate RELAP5. The heated wall of test section has its thickness of 8.2 cm and the same material as the prototype (APR1400) with chrome coating against rusting. From the experiment, we visually observe the vapor jetting near the heated wall with small bubble migration to the bulk region and liquid circulation. The data shows a rapid wall temperature drop generating a large amount of vapor initially. The calculation results of RELAP5 using the three nodal schemes are compared with experimental ones in aspects of water level, void fraction, wall temperatures and phase velocities. It turns out that the double nodal scheme with circulation produces better results than the nodal scheme without circulation to simulate the boiling phenomena in the downcomer. (authors)

  13. Effect of subcooling and wall thickness on pool boiling from downward-facing curved surfaces in water

    SciTech Connect (OSTI)

    El-Genk, M.S.; Glebov, A.G.

    1995-09-01

    Quenching experiments were performed to investigate the effects of water subcooling and wall thickness on pool boiling from a downward-facing curved surface. Experiments used three copper sections of the same diameter (50.8 mm) and surface radius (148 mm), but different thickness (12.8, 20 and 30 mm). Local and average pool boiling curves were obtained at saturation and 5 K, 10 K, and 14 K subcooling. Water subcooling increased the maximum heat flux, but decreased the corresponding wall superheat. The minimum film boiling heat flux and the corresponding wall superheat, however, increased with increased subcooling. The maximum and minimum film boiling heat fluxes were independent of wall thickness above 20 mm and Biot Number > 0.8, indicating that boiling curves for the 20 and 30 thick sections were representative of quasi steady-state, but not those for the 12.8 mm thick section. When compared with that for a flat surface section of the same thickness, the data for the 12.8 mm thick section showed significant increases in both the maximum heat flux (from 0.21 to 0.41 MW/m{sup 2}) and the minimum film boiling heat flux (from 2 to 13 kW/m{sup 2}) and about 11.5 K and 60 K increase in the corresponding wall superheats, respectively.

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

    SciTech Connect (OSTI)

    Wheeler, Timothy A.; Liao, Huafei

    2014-12-01

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

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

    SciTech Connect (OSTI)

    Orlov, Andrey; Degueldre, Claude; Kaufmann, Wilfried

    2013-01-15

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

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

    SciTech Connect (OSTI)

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

    2005-12-01

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

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

    SciTech Connect (OSTI)

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

    1988-06-20

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

  18. Local pressure gradients due to incipience of boiling in subcooled flows

    SciTech Connect (OSTI)

    Ruggles, A.E.; McDuffee, J.L.

    1995-09-01

    Models for vapor bubble behavior and nucleation site density during subcooled boiling are integrated with boundary layer theory in order to predict the local pressure gradient and heat transfer coefficient. Models for bubble growth rate and bubble departure diameter are used to scale the movement of displaced liquid in the laminar sublayer. An added shear stress, analogous to a turbulent shear stress, is derived by considering the liquid movement normal to the heated surface. The resulting mechanistic model has plausible functional dependence on wall superheat, mass flow, and heat flux and agrees well with data available in the literature.

  19. CASL-U-2015-0248-000 Modeling Boiling Water Reactor

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    8-000 Modeling Boiling Water Reactor Designs using MPACT Andrew P. Fitzgerald Brendan Kochunas Daniel Jabaay Thomas Downar University of Michigan July 7, 2015 CASL-U-2015-0248-000 ATRIUM TM 10: K-inf vs burn-up for the ATRIUM TM 10 lattice from various transport codes. MPACT is shown to have the ability to model some BWR features such as (square) channel boxes, water rods, and water channels with reasonable accuracy. The ATRIUM TM 10 comparison has shown MPACT can predict k-inf with similar

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

    SciTech Connect (OSTI)

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

    1997-01-01

    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.

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

    SciTech Connect (OSTI)

    Murphy, T.J.

    1987-09-01

    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.

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

    SciTech Connect (OSTI)

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

    2011-06-01

    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.

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

    SciTech Connect (OSTI)

    Chaubet, C.; Raymond, A.; Bisotto, I.; Harmand, J. C.; Kubisa, M.; Zawadzki, W.

    2013-12-04

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

  4. Waves on the surface of a boiling liquid at various medium stratifications

    SciTech Connect (OSTI)

    Sinkevich, O. A.

    2015-08-15

    The stability of relatively small perturbations of the stationary state consisting of a plane liquid layer and a vapor film is studied when no liquid evaporation or vapor condensation occurs in the stationary state. In this case, heat from a hot to cold wall is removed through a vapor–liquid layer via heat conduction. The boundary conditions that take into account liquid evaporation (appearance of a mass flux) at the vapor–liquid phase surface and the temperature dependence of the saturation pressure are derived. Dispersion equations are obtained. The wave processes for the stable (light vapor under a liquid layer) and unstable stratifications of the phases at rest and during their relative motion are studied. The deformation of the phase boundary results in liquid evaporation, changes in the boiling temperature and the saturation pressure, and generation of weakly damped low-amplitude waves of a new type. These waves ensure the stability of a vapor film under a liquid layer at rest or a liquid layer moving at a constant velocity in the gravity field. The velocities of these waves are much higher than the gravity wave velocities. The critical heat flows and wavelengths at which wave boiling regimes at normal pressure can exist are determined, and the calculated and experimental data are compared.

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

    SciTech Connect (OSTI)

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

    2004-05-07

    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.

  6. A Study of the Role of Adjoint-Equipped CFD in VUQ Analysis of Channel Boiling Simulations -Slides

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Study of the Role of Adjoint- Equipped CFD in VUQ Analysis of Channel Boiling Simulations Krzysztof Fidkowski University of Michigan Milestone L3:THM.CFD.P7.08 November 21, 2013 CASL-U-2013-0192-000-b L3-THM-CFD-P7-08 A Study of the Role of Adjoint-Equipped CFD in VUQ Analysis of Channel Boiling Simulations Milestone owner: Krzysztof Fidkowski, U. Michigan Additional personnel: Isaac Asher, U. Michigan 2 CASL-U-2013-0192-000-b Milestone Execution Responsibility & Personnel * Contact:

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

    SciTech Connect (OSTI)

    Schlereth, J.R.; Pennington, D.

    1996-12-01

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

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

    SciTech Connect (OSTI)

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

    1991-01-01

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

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

    SciTech Connect (OSTI)

    PACE, M.E.

    2004-01-13

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

  10. Statistical modeling support for calibration of a multiphysics model of subcooled boiling flows

    SciTech Connect (OSTI)

    Bui, A. V.; Dinh, N. T.; Nourgaliev, R. R.; Williams, B. J.

    2013-07-01

    Nuclear reactor system analyses rely on multiple complex models which describe the physics of reactor neutronics, thermal hydraulics, structural mechanics, coolant physico-chemistry, etc. Such coupled multiphysics models require extensive calibration and validation before they can be used in practical system safety study and/or design/technology optimization. This paper presents an application of statistical modeling and Bayesian inference in calibrating an example multiphysics model of subcooled boiling flows which is widely used in reactor thermal hydraulic analysis. The presence of complex coupling of physics in such a model together with the large number of model inputs, parameters and multidimensional outputs poses significant challenge to the model calibration method. However, the method proposed in this work is shown to be able to overcome these difficulties while allowing data (observation) uncertainty and model inadequacy to be taken into consideration. (authors)

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

    DOE Patents [OSTI]

    Hill, Paul R.

    1994-01-01

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

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

    DOE Patents [OSTI]

    Hill, P.R.

    1994-12-27

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

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

    SciTech Connect (OSTI)

    Uddin, Rizwan

    2012-01-01

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

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

    SciTech Connect (OSTI)

    Not Available

    1986-09-01

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

  15. Transient pool boiling heat transfer due to increasing heat inputs in subcooled water at high pressures

    SciTech Connect (OSTI)

    Fukuda, K.; Shiotsu, M.; Sakurai, A.

    1995-09-01

    Understanding of transient boiling phenomenon caused by increasing heat inputs in subcooled water at high pressures is necessary to predict correctly a severe accident due to a power burst in a water-cooled nuclear reactor. Transient maximum heat fluxes, q{sub max}, on a 1.2 mm diameter horizontal cylinder in a pool of saturated and subcooled water for exponential heat inputs, q{sub o}e{sup t/T}, with periods, {tau}, ranging from about 2 ms to 20 s at pressures from atmospheric up to 2063 kPa for water subcoolings from 0 to about 80 K were measured to obtain the extended data base to investigate the effect of high subcoolings on steady-state and transient maximum heat fluxes, q{sub max}. Two main mechanisms of q{sub max} exist depending on the exponential periods at low subcoolings. One is due to the time lag of the hydrodynamic instability which starts at steady-state maximum heat flux on fully developed nucleate boiling (FDNB), and the other is due to the heterogenous spontaneous nucleations (HSN) in flooded cavities which coexist with vapor bubbles growing up from active cavities. The shortest period corresponding to the maximum q{sub max} for long period range belonging to the former mechanism becomes longer and the q{sub max}mechanism for long period range shifts to that due the HSN on FDNB with the increase of subcooling and pressure. The longest period corresponding to the minimum q{sub max} for the short period range belonging to the latter mechanism becomes shorter with the increase in saturated pressure. On the contrary, the longest period becomes longer with the increase in subcooling at high pressures. Correlations for steady-state and transient maximum heat fluxes were presented for a wide range of pressure and subcooling.

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

    SciTech Connect (OSTI)

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

    2004-10-06

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

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

    SciTech Connect (OSTI)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

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

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

    SciTech Connect (OSTI)

    Jessee, Matthew Anderson

    2011-01-01

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

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

    SciTech Connect (OSTI)

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

    2012-07-01

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

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

    SciTech Connect (OSTI)

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

    2012-07-01

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

  1. Thermalhydraulic calculation for boiling water reactor and its natural circulation component

    SciTech Connect (OSTI)

    Trianti, Nuri Nurjanah,; Su’ud, Zaki; Arif, Idam; Permana, Sidik

    2015-09-30

    Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density and inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.

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

    SciTech Connect (OSTI)

    1995-08-01

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

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

    SciTech Connect (OSTI)

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

    2010-09-01

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

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

    SciTech Connect (OSTI)

    Fellhauer, C.R.; Boing, L.E.; Aldana, J.

    1997-03-01

    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).

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

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

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

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

    SciTech Connect (OSTI)

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

    2005-01-01

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

  7. Hydrogen water chemistry for BWRs (boiling water reactors): Materials behavior: Interim report

    SciTech Connect (OSTI)

    Gordon, B.M.; Jewett, C.W.; Pickett, A.E.; Walker, W.L.; Indig, M.E.; Andresen, P.L.; Niedrach, L.W.; Davis, R.B.

    1987-03-01

    The objective of this research program is to provide test data to guide future actions by boiling water reactor (BWR) owners regarding the use of hydrogen additions to the feedwater to mitigate pipe cracking during power operation. Numerous laboratory testing methods and approaches are being utilized in this program to evaluate and quantify the effects of this hydrogen water chemistry (HWC) on the corrosion performance of reactor materials, including full-scale pipe testing, fatigue crack initiation and growth studies, constant load tests, electrochemical potential (ECP) measurements, constant extension rate technique (CERT) testing, straining electrode tests (SET), oxide film analysis, fracture mechanics studies, general corrosion investigations and bent beam tests. The results to date are summarized in this report and indicate that HWC (which implies an ECP of Type-304 stainless steel below -230 mV/sub SHE/ coupled with a low water conductivity) generally has a beneficial effect on the corrosion performance of BWR structural materials. Specifically, HWC mitigates intergranular stress corrosion cracking (IGSCC) initiation and propagation in piping, provides an improved margin against environmental cracking in carbon steel and low alloy steel, and does not promote environmental cracking in other materials. A measurable, but acceptable, increase in the initial general corrosion kinetics of carbon and low alloy steel also accompanies the use of HWC. The laboratory data, together with the in-reactor test results, clearly indicate that HWC is an effective method of reducing the likelihood and rate of BWR pipe cracking.

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

    SciTech Connect (OSTI)

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

    2010-09-15

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

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

    SciTech Connect (OSTI)

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

    1980-06-01

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

  10. Effect of surface oxidation on the onset of nucleate boiling in a materials test reactor coolant channel

    DOE PAGES-Beta [OSTI]

    Forrest, Eric C.; Don, Sarah M.; Hu, Lin -Wen; Buongiorno, Jacopo; McKrell, Thomas J.

    2016-02-29

    The onset of nucleate boiling (ONB) serves as the thermal-hydraulic operating limit for many research and test reactors. However, boiling incipience under forced convection has not been well-characterized in narrow channel geometries or for oxidized surface conditions. This study presents experimental data for the ONB in vertical upflow of deionized (DI) water in a simulated materials test reactor (MTR) coolant channel. The channel gap thickness and aspect ratio were 1.96 mm and 29:1, respectively. Boiling surface conditions were carefully controlled and characterized, with both heavily oxidized and native oxide surfaces tested. Measurements were performed for mass fluxes ranging from 750more » to 3000 kg/m2s and for subcoolings ranging from 10 to 45°C. ONB was identified using a combination of high-speed visual observation, surface temperature measurements, and channel pressure drop measurements. Surface temperature measurements were found to be most reliable in identifying the ONB. For the nominal (native oxide) surface, results indicate that the correlation of Bergles and Rohsenow, when paired with the appropriate single-phase heat transfer correlation, adequately predicts the ONB heat flux. Furthermore, incipience on the oxidized surface occurred at a higher heat flux and superheat than on the plain surface.« less

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

    SciTech Connect (OSTI)

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

    2011-01-01

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

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

    SciTech Connect (OSTI)

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

    2012-07-01

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

  13. Nondestructive assay of spent boiling water reactor fuel by active neutron interrogation

    SciTech Connect (OSTI)

    Blakeman, E.D.; Ricker, C.W.; Ragan, G.L.; Difilippo, F.C.; Slaughter, G.G.

    1981-01-01

    Spent boiling water reactor (BWR) fuel from Dresden I was assayed for total fissile mass, using the active neutron interrogation method. The nondestructive assay (NDA) system used has four Sb-Be sources for interrogation of the fuels; the induced fission neutrons from the fuel are counted by four lead-shielded methane-filled proportional counters biased above the energy of the source neutrons. Spent fuel rods containing 9 kg of heavy metal were chopped into 5-cm segments and loaded into three 1-liter cans. The three cans were assayed in seven combinations of one, two, or three cans, enabling an evaluation of the precision and accuracy of the NDA system for different amounts of fissile material. The fissile mass in each combination was determined by comparing the induced-fission-neutron counts with the counts obtained from a known standard comprising chopped segments of unirradiated Dresden fuel. These masses were compared to the masses determined by chemical analyses of the spent fuel. The results from the nondestructive assays agreed with results from the chemical analyses to within 2 to 3%. Similar agreement was obtained when two combinations of canned spent fuel were used as standards for the nondesctuctive assays. The assay of BWR spent fuel served as a test of the NDA system which was developed at the Oak Ridge National Laboratory for the assay of spent liquid metal fast breeder reactor (LMFBR) fuel subassemblies at the heat-end of a reprocessing plant. Results of previous experiments and calculations reported earlier using simulated LMFBR fuel subassemblies indicated that the NDA system can measure the fissile masses of spent fuel subassemblies to within an accuracy of 3%. Results of the assays of spent BWR fuel reported herein support this conclusion.

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

    SciTech Connect (OSTI)

    1994-04-30

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

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

    SciTech Connect (OSTI)

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

    2003-06-16

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

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

    SciTech Connect (OSTI)

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

    2010-08-11

    production in a boiling water reactor fuel bundle based on measurements taken from the corresponding fuel assembly channel. Our preliminary results are in good agreement with the actual operating history of the reactor during the cycle that the fuel bundle was resident in the core.

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

    SciTech Connect (OSTI)

    Zhegang Ma; Diego Mandelli; Curtis Smith

    2011-07-01

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

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

    SciTech Connect (OSTI)

    Anh Bui; Nam Dinh; Brian Williams

    2013-09-01

    In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Such sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this works 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

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

    SciTech Connect (OSTI)

    Boing, L.E.; Henley, D.R. ); Manion, W.J.; Gordon, J.W. )

    1989-12-01

    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.

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

    SciTech Connect (OSTI)

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

    2006-10-20

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

  1. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    SciTech Connect (OSTI)

    Korteniemi, V.; Haapalehto, T.; Puustinen, M.

    1995-09-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

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

    SciTech Connect (OSTI)

    Not Available

    1994-07-01

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

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

    SciTech Connect (OSTI)

    Not Available

    1994-07-01

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

  4. SWR 1000: An Advanced, Medium-Sized Boiling Water Reactor, Ready for Deployment

    SciTech Connect (OSTI)

    Brettschuh, Werner

    2006-07-01

    The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 is particularly suitable for countries whose power systems are not designed for large-capacity generating facilities. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (I and C) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10 x 10 rod array to a 12 x 12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70

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

    SciTech Connect (OSTI)

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

    1982-03-01

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

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

    SciTech Connect (OSTI)

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

    2007-10-15

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

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

    SciTech Connect (OSTI)

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

    1982-03-01

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

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

    SciTech Connect (OSTI)

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

    2008-07-08

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

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

    SciTech Connect (OSTI)

    1993-09-15

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

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

    SciTech Connect (OSTI)

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

    1983-03-01

    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.

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

    SciTech Connect (OSTI)

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

    2004-01-15

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

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

    SciTech Connect (OSTI)

    Knotek, P.; Tichy, L.

    2013-09-01

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

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

    SciTech Connect (OSTI)

    Not Available

    1993-05-13

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

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

    SciTech Connect (OSTI)

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

    2014-09-30

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

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

    SciTech Connect (OSTI)

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

    2007-10-15

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

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

    SciTech Connect (OSTI)

    Dykin, V.; Pazsit, I.

    2012-07-01

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

  17. Development, implementation and assessment of specific, two-fluid closure laws for inverted-annular film-boiling

    SciTech Connect (OSTI)

    Cachard, F. de

    1995-09-01

    Inverted-Annular Film-Boiling (IAFB) is one of the post-burnout heat transfer modes taking place during the reflooding phase of the loss-of-coolant accident, when the liquid at the quench front is subcooled. Under IAFB conditions, a continuous, liquid core is separated from the wall by a superheated vapour film. the heat transfer rate in IAFB is influenced by the flooding rate, liquid subcooling, pressure, and the wall geometry and temperature. These influences can be accounted by a two-fluid model with physically sound closure laws for mass, momentum and heat transfers between the wall, the vapour film, the vapour-liquid interface, and the liquid core. Such closure laws have been developed and adjusted using IAFB-relevant experimental results, including heat flux, wall temperature and void fraction data. The model is extensively assessed against data from three independent sources. A total of 46 experiments have been analyzed. The overall predictions are good. The IAFB-specific closure laws proposed have also intrinsic value, and may be used in other two-fluid models. They should allow to improve the description of post-dryout, low quality heat transfer by the safety codes.

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

    SciTech Connect (OSTI)

    1997-05-01

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

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

    SciTech Connect (OSTI)

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

    1994-10-01

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

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

    SciTech Connect (OSTI)

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

    1980-06-01

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

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

    SciTech Connect (OSTI)

    Douglas M. Gerstner

    2009-05-01

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

  2. Estimating boiling water reactor decommissioning costs: A user`s manual for the BWR Cost Estimating Computer Program (CECP) software. Final report

    SciTech Connect (OSTI)

    Bierschbach, M.C.

    1996-06-01

    Nuclear power plant licensees are required to submit to the US Nuclear Regulatory Commission (NRC) for review their decommissioning cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning boiling water reactor (BWR) power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

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

    SciTech Connect (OSTI)

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

    2000-02-15

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

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

    SciTech Connect (OSTI)

    Sterbentz, J.W.

    1994-10-01

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

  5. An experimental study on sub-cooled flow boiling CHF of R134a at low pressure condition with atmospheric pressure (AP) plasma assisted surface modification

    SciTech Connect (OSTI)

    Kim, Seung Jun; Zou, Ling; Jones, Barclay G.

    2015-02-01

    In this study, sub-cooled flow boiling critical heat flux tests at low pressure were conducted in a rectangular flow channel with one uniformly heated surface, using simulant fluid R-134a as coolant. The experiments were conducted under the following conditions: (1) inlet pressure (P) of 400-800 kPa, (2) mass flux (G) of 124-248 kg/m2s, (3) inlet sub-cooling enthalpy (ΔHi) of 12~ 26 kJ/kg. Parametric trends of macroscopic system parameters (G, P, Hi) were examined by changing inlet conditions. Those trends were found to be generally consistent with previous understandings of CHF behavior at low pressure condition (i.e. reduced pressure less than 0.2). A fluid-to-fluid scaling model was utilized to convert the test data obtained with the simulant fluid (R-134a) into the prototypical fluid (water). The comparison between the converted CHF of equivalent water and CHF look-up table with same operation conditions were conducted, which showed good agreement. Furthermore, the effect of surface wettability on CHF was also investigated by applying atmospheric pressure plasma (AP-Plasma) treatment to modify the surface characteristic. With AP-Plasma treatment, the change of microscopic surface characteristic was measured in terms of static contact angle. The static contact angle was reduced from 80° on original non-treated surface to 15° on treated surface. An enhancement of 18% on CHF values under flow boiling conditions were observed on AP-Plasma treated surfaces compared to those on non-treated heating surfaces.

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

    SciTech Connect (OSTI)

    Slater, C.O.

    1990-07-01

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

  7. Technology, safety and costs of decommissioning a reference boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

    SciTech Connect (OSTI)

    Konzek, G.J.; Smith, R.I.

    1988-07-01

    Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.

  8. 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 (OSTI)

    Konzek, G.J.; Smith, R.I. )

    1990-12-01

    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.

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

    SciTech Connect (OSTI)

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

    2013-09-01

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

  10. TH-A-9A-06: Inverse Planning of Gamma Knife Radiosurgery Using...

    Office of Scientific and Technical Information (OSTI)

    obtained by solving a constrained integer-linear problem. (4) The shots are placed into ... Subject: 60 APPLIED LIFE SCIENCES; ALGORITHMS; GEOMETRY; KERNELS; NEOPLASMS; OPTIMIZATION; ...

  11. TH-A-9A-04: Incorporating Liver Functionality in Radiation Therapy Treatment Planning

    SciTech Connect (OSTI)

    Wu, V; Epelman, M; Feng, M; Cao, Y; Wang, H; Romeijn, E; Matuszak, M

    2014-06-15

    Purpose: Liver SBRT patients have both variable pretreatment liver function (e.g., due to degree of cirrhosis and/or prior treatments) and sensitivity to radiation, leading to high variability in potential liver toxicity with similar doses. This work aims to explicitly incorporate liver perfusion into treatment planning to redistribute dose to preserve well-functioning areas without compromising target coverage. Methods: Voxel-based liver perfusion, a measure of functionality, was computed from dynamic contrast-enhanced MRI. Two optimization models with different cost functions subject to the same dose constraints (e.g., minimum target EUD and maximum critical structure EUDs) were compared. The cost functions minimized were EUD (standard model) and functionality-weighted EUD (functional model) to the liver. The resulting treatment plans delivering the same target EUD were compared with respect to their DVHs, their dose wash difference, the average dose delivered to voxels of a particular perfusion level, and change in number of high-/low-functioning voxels receiving a particular dose. Two-dimensional synthetic and three-dimensional clinical examples were studied. Results: The DVHs of all structures of plans from each model were comparable. In contrast, in plans obtained with the functional model, the average dose delivered to high-/low-functioning voxels was lower/higher than in plans obtained with its standard counterpart. The number of high-/low-functioning voxels receiving high/low dose was lower in the plans that considered perfusion in the cost function than in the plans that did not. Redistribution of dose can be observed in the dose wash differences. Conclusion: Liver perfusion can be used during treatment planning potentially to minimize the risk of toxicity during liver SBRT, resulting in better global liver function. The functional model redistributes dose in the standard model from higher to lower functioning voxels, while achieving the same target EUD and satisfying dose limits to critical structures. This project is funded by MCubed and grant R01-CA132834.

  12. TH-A-17A-01: Innovation in PET Instrumentation and Applications

    SciTech Connect (OSTI)

    Casey, M; Miyaoka, R; Shao, Y

    2014-06-15

    Innovation in PET instrumentation has led to the new millennium revolutionary imaging applications for diagnosis, therapeutic guidance, and development of new molecular imaging probes, etc. However, after several decades innovations, will the advances of PET technology and applications continue with the same trend and pace? What will be the next big thing beyond the PET/CT, PET/MRI, and Time-of-flight PET? How will the PET instrumentation and imaging performance be further improved by novel detector research and advanced imaging system development? Or will the development of new algorithms and methodologies extend the limit of current instrumentation and leapfrog the imaging quality and quantification for practical applications? The objective of this session is to present an overview of current status and advances in the PET instrumentation and applications with speakers from leading academic institutes and a major medical imaging company. Presenting with both academic research projects and commercial technology developments, this session will provide a glimpse of some latest advances and challenges in the field, such as using semiconductor photon-sensor based PET detectors to improve performance and enable new applications, as well as the technology trend that may lead to the next breakthrough in PET imaging for clinical and preclinical applications. Both imaging and image-guided therapy subjects will be discussed. Learning Objectives: Describe the latest innovations in PET instrumentation and applications Understand the driven force behind the PET instrumentation innovation and development Learn the trend of PET technology development for applications.

  13. TH-A-18A-01: Innovation in Clinical Breast Imaging

    SciTech Connect (OSTI)

    Liu, B; Yang, K; Yaffe, M; Chen, J

    2014-06-15

    Several novel modalities have been or are on the verge of being introduced into the breast imaging clinic. These include tomosynthesis imaging, dedicated breast CT, contrast-enhanced digital mammography, and automated breast ultrasound, all of which are covered in this course. Tomosynthesis and dedicated breast CT address the problem of tissue superimposition that limits mammography screening performance, by improved or full resolution of the 3D breast morphology. Contrast-enhanced digital mammography provides functional information that allows for visualization of tumor angiogenesis. 3D breast ultrasound has high sensitivity for tumor detection in dense breasts, but the imaging exam was traditionally performed by radiologists. In automated breast ultrasound, the scan is performed in an automated fashion, making for a more practical imaging tool, that is now used as an adjunct to digital mammography in breast cancer screening. This course will provide medical physicists with an in-depth understanding of the imaging physics of each of these four novel imaging techniques, as well as the rationale and implementation of QC procedures. Further, basic clinical applications and work flow issues will be discussed. Learning Objectives: To be able to describe the underlying physical and physiological principles of each imaging technique, and to understand the corresponding imaging acquisition process. To be able to describe the critical system components and their performance requirements. To understand the rationale and implementation of quality control procedures, as well as regulatory requirements for systems with FDA approval. To learn about clinical applications and understand risks and benefits/strength and weakness of each modality in terms of clinical breast imaging.

  14. Efficient Cooling in Engines with Nucleated Boiling

    Office of Energy Efficiency and Renewable Energy (EERE)

    2009 DOE Hydrogen Program and Vehicle Technologies Program Annual Merit Review and Peer Evaluation Meeting, May 18-22, 2009 -- Washington D.C.

  15. Self-Sustaining Thorium Boiling Water Reactors

    SciTech Connect (OSTI)

    Greenspan, Ehud; Gorman, Phillip M.; Bogetic, Sandra; Seifried, Jeffrey E.; Zhang, Guanheng; Varela, Christopher R.; Fratoni, Massimiliano; Vijic, Jasmina J.; Downar, Thomas; Hall, Andrew; Ward, Andrew; Jarrett, Michael; Wysocki, Aaron; Xu, Yunlin; Kazimi, Mujid; Shirvan, Koroush; Mieloszyk, Alexander; Todosow, Michael; Brown, Nicolas; Cheng, Lap

    2015-03-15

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  16. Boiling Springs Geothermal Area | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    100C373.15 K 212 F 671.67 R 1 USGS Estimated Reservoir Volume: 1 km 1 USGS Mean Capacity: 6 MW 1 Click "Edit With Form" above to add content History and...

  17. TH-A-9A-08: Knowledge-Based Quality Control of Clinical Stereotactic Radiosurgery Treatment Plans

    SciTech Connect (OSTI)

    Shiraishi, S; Moore, K L; Tan, J; Olsen, L

    2014-06-15

    Purpose: To develop a quality control tool to reduce stereotactic radiosurgery (SRS) planning variability using models that predict achievable plan quality metrics (QMs) based on individual patient anatomy. Methods: Using a knowledge-based methodology that quantitatively correlates anatomical geometric features to resultant organ-at-risk (OAR) dosimetry, we developed models for predicting achievable OAR dose-volume histograms (DVHs) by training with a cohort of previously treated SRS patients. The DVH-based QMs used in this work are the gradient measure, GM=(3/4pi)^1/3*[V50%^1/3−V100%^1/3], and V10Gy of normal brain. As GM quantifies the total rate of dose fall-off around the planning target volume (PTV), all voxels inside the patient's body contour were treated as OAR for DVH prediction. 35 previously treated SRS plans from our institution were collected; all were planned with non-coplanar volumetric-modulated arc therapy to prescription doses of 12–25 Gy. Of the 35-patient cohort, 15 were used for model training and 20 for model validation. Accuracies of the predictions were quantified by the mean and the standard deviation of the difference between clinical and predicted QMs, δQM=QM-clin−QM-pred. Results: Best agreement between predicted and clinical QMs was obtained when models were built separately for V-PTV<2.5cc and V-PTV>2.5cc. Eight patients trained the V-PTV<2.5cc model and seven patients trained the V-PTV>2.5cc models, respectively. The mean and the standard deviation of δGM were 0.3±0.4mm for the training sets and −0.1±0.6mm for the validation sets, demonstrating highly accurate GM predictions. V10Gy predictions were also highly accurate, with δV10Gy=0.8±0.7cc for the training sets and δV10Gy=0.7±1.4cc for the validation sets. Conclusion: The accuracy of the models in predicting two key SRS quality metrics highlights the potential of this technique for quality control for SRS treatments. Future investigations will seek to determine whether QM variations are due to residual model inaccuracies or true plan quality variations in the clinical sample. Support from Varian Medical Systems.

  18. TH-A-18C-04: Ultrafast Cone-Beam CT Scatter Correction with GPU-Based Monte Carlo Simulation

    SciTech Connect (OSTI)

    Xu, Y; Bai, T; Yan, H; Ouyang, L; Wang, J; Pompos, A; Jiang, S; Jia, X; Zhou, L

    2014-06-15

    Purpose: Scatter artifacts severely degrade image quality of cone-beam CT (CBCT). We present an ultrafast scatter correction framework by using GPU-based Monte Carlo (MC) simulation and prior patient CT image, aiming at automatically finish the whole process including both scatter correction and reconstructions within 30 seconds. Methods: The method consists of six steps: 1) FDK reconstruction using raw projection data; 2) Rigid Registration of planning CT to the FDK results; 3) MC scatter calculation at sparse view angles using the planning CT; 4) Interpolation of the calculated scatter signals to other angles; 5) Removal of scatter from the raw projections; 6) FDK reconstruction using the scatter-corrected projections. In addition to using GPU to accelerate MC photon simulations, we also use a small number of photons and a down-sampled CT image in simulation to further reduce computation time. A novel denoising algorithm is used to eliminate MC scatter noise caused by low photon numbers. The method is validated on head-and-neck cases with simulated and clinical data. Results: We have studied impacts of photo histories, volume down sampling factors on the accuracy of scatter estimation. The Fourier analysis was conducted to show that scatter images calculated at 31 angles are sufficient to restore those at all angles with <0.1% error. For the simulated case with a resolution of 512×512×100, we simulated 10M photons per angle. The total computation time is 23.77 seconds on a Nvidia GTX Titan GPU. The scatter-induced shading/cupping artifacts are substantially reduced, and the average HU error of a region-of-interest is reduced from 75.9 to 19.0 HU. Similar results were found for a real patient case. Conclusion: A practical ultrafast MC-based CBCT scatter correction scheme is developed. The whole process of scatter correction and reconstruction is accomplished within 30 seconds. This study is supported in part by NIH (1R01CA154747-01), The Core Technology Research in Strategic Emerging Industry, Guangdong, China (2011A081402003)

  19. TH-A-18C-06: A Scatter Elimination Scheme for Cone Beam CT Using An Oscillating Narrow Beam

    SciTech Connect (OSTI)

    Yan, H; Folkerts, M; Jia, X; Jiang, S; Xu, Y

    2014-06-15

    Purpose: While cone beam CT (CBCT) has been widely used in image guided radiation therapy, its low image quality, primarily caused by scattered x-rays, hinders advanced clinical applications, e.g., CBCT based on-line adaptive re-planning. We propose in this abstract a new scheme called oscillating narrow beam CBCT (ONB-CBCT) to eliminate scatter signals. Methods: ONB-CBCT consists of two major components. 1) Oscillating narrow beam (ONB) scan and 2) partitioned flat panel containing multiple individual detector strips and their own readouts. Both the beam oscillation and detector partition are along the superior-inferior (SI) direction. During data acquisition, at a given projection, the narrow beam sweep through the detector region, and different portions of the detector acquires projection data in synchrony with the narrow beam. ONB can be generated by a rotating slit collimator design with conventional tube with single focal spot, or by directly using a new source with multiple focal spots. A proof-of-principle study via Monte Carlo simulation is conducted to demonstrate the feasibility of ONB-CBCT. Results: As the beam becomes narrower, more and more scatter signals are eliminated. For the case with a bowtie filter and using 15 ONBs, the maximum and the average intensity error due to scatter are below 20 and 10 HU, respectively. Conclusion: ONB yields a narrowed exposure field at each snapshot and hence an inherently negligible scatter effect. Meanwhile, the individualized detector units guarantee high frame rate detection and hence a same large volume coverage as that in conventional CBCT. In summary, ONB-CBCT is a promising design to achieve high-quality CBCT imaging. This study is supported in part by NIH (1R01CA154747-01)

  20. TH-A-9A-01: Active Optical Flow Model: Predicting Voxel-Level Dose Prediction in Spine SBRT

    SciTech Connect (OSTI)

    Liu, J; Wu, Q.J.; Yin, F; Kirkpatrick, J; Cabrera, A; Ge, Y

    2014-06-15

    Purpose: To predict voxel-level dose distribution and enable effective evaluation of cord dose sparing in spine SBRT. Methods: We present an active optical flow model (AOFM) to statistically describe cord dose variations and train a predictive model to represent correlations between AOFM and PTV contours. Thirty clinically accepted spine SBRT plans are evenly divided into training and testing datasets. The development of predictive model consists of 1) collecting a sequence of dose maps including PTV and OAR (spinal cord) as well as a set of associated PTV contours adjacent to OAR from the training dataset, 2) classifying data into five groups based on PTV's locations relative to OAR, two “Top”s, “Left”, “Right”, and “Bottom”, 3) randomly selecting a dose map as the reference in each group and applying rigid registration and optical flow deformation to match all other maps to the reference, 4) building AOFM by importing optical flow vectors and dose values into the principal component analysis (PCA), 5) applying another PCA to features of PTV and OAR contours to generate an active shape model (ASM), and 6) computing a linear regression model of correlations between AOFM and ASM.When predicting dose distribution of a new case in the testing dataset, the PTV is first assigned to a group based on its contour characteristics. Contour features are then transformed into ASM's principal coordinates of the selected group. Finally, voxel-level dose distribution is determined by mapping from the ASM space to the AOFM space using the predictive model. Results: The DVHs predicted by the AOFM-based model and those in clinical plans are comparable in training and testing datasets. At 2% volume the dose difference between predicted and clinical plans is 4.2±4.4% and 3.3±3.5% in the training and testing datasets, respectively. Conclusion: The AOFM is effective in predicting voxel-level dose distribution for spine SBRT. Partially supported by NIH/NCI under grant #R21CA161389 and a master research grant by Varian Medical System.

  1. TH-A-18C-09: Ultra-Fast Monte Carlo Simulation for Cone Beam CT Imaging of Brain Trauma

    SciTech Connect (OSTI)

    Sisniega, A; Zbijewski, W; Stayman, J; Yorkston, J; Aygun, N; Koliatsos, V; Siewerdsen, J

    2014-06-15

    Purpose: Application of cone-beam CT (CBCT) to low-contrast soft tissue imaging, such as in detection of traumatic brain injury, is challenged by high levels of scatter. A fast, accurate scatter correction method based on Monte Carlo (MC) estimation is developed for application in high-quality CBCT imaging of acute brain injury. Methods: The correction involves MC scatter estimation executed on an NVIDIA GTX 780 GPU (MC-GPU), with baseline simulation speed of ~1e7 photons/sec. MC-GPU is accelerated by a novel, GPU-optimized implementation of variance reduction (VR) techniques (forced detection and photon splitting). The number of simulated tracks and projections is reduced for additional speed-up. Residual noise is removed and the missing scatter projections are estimated via kernel smoothing (KS) in projection plane and across gantry angles. The method is assessed using CBCT images of a head phantom presenting a realistic simulation of fresh intracranial hemorrhage (100 kVp, 180 mAs, 720 projections, source-detector distance 700 mm, source-axis distance 480 mm). Results: For a fixed run-time of ~1 sec/projection, GPU-optimized VR reduces the noise in MC-GPU scatter estimates by a factor of 4. For scatter correction, MC-GPU with VR is executed with 4-fold angular downsampling and 1e5 photons/projection, yielding 3.5 minute run-time per scan, and de-noised with optimized KS. Corrected CBCT images demonstrate uniformity improvement of 18 HU and contrast improvement of 26 HU compared to no correction, and a 52% increase in contrast-tonoise ratio in simulated hemorrhage compared to “oracle” constant fraction correction. Conclusion: Acceleration of MC-GPU achieved through GPU-optimized variance reduction and kernel smoothing yields an efficient (<5 min/scan) and accurate scatter correction that does not rely on additional hardware or simplifying assumptions about the scatter distribution. The method is undergoing implementation in a novel CBCT dedicated to brain trauma imaging at the point of care in sports and military applications. Research grant from Carestream Health. JY is an employee of Carestream Health.

  2. TH-A-16A-01: Image Quality for the Radiation Oncology Physicist: Review of the Fundamentals and Implementation

    SciTech Connect (OSTI)

    Seibert, J; Imbergamo, P

    2014-06-15

    The expansion and integration of diagnostic imaging technologies such as On Board Imaging (OBI) and Cone Beam Computed Tomography (CBCT) into radiation oncology has required radiation oncology physicists to be responsible for and become familiar with assessing image quality. Unfortunately many radiation oncology physicists have had little or no training or experience in measuring and assessing image quality. Many physicists have turned to automated QA analysis software without having a fundamental understanding of image quality measures. This session will review the basic image quality measures of imaging technologies used in the radiation oncology clinic, such as low contrast resolution, high contrast resolution, uniformity, noise, and contrast scale, and how to measure and assess them in a meaningful way. Additionally a discussion of the implementation of an image quality assurance program in compliance with Task Group recommendations will be presented along with the advantages and disadvantages of automated analysis methods. Learning Objectives: Review and understanding of the fundamentals of image quality. Review and understanding of the basic image quality measures of imaging modalities used in the radiation oncology clinic. Understand how to implement an image quality assurance program and to assess basic image quality measures in a meaningful way.

  3. TH-A-19A-10: Fast Four Dimensional Monte Carlo Dose Computations for Proton Therapy of Lung Cancer

    SciTech Connect (OSTI)

    Mirkovic, D; Titt, U; Mohan, R; Yepes, P

    2014-06-15

    Purpose: To develop and validate a fast and accurate four dimensional (4D) Monte Carlo (MC) dose computation system for proton therapy of lung cancer and other thoracic and abdominal malignancies in which the delivered dose distributions can be affected by respiratory motion of the patient. Methods: A 4D computer tomography (CT) scan for a lung cancer patient treated with protons in our clinic was used to create a time dependent patient model using our in-house, MCNPX-based Monte Carlo system (“MC{sup 2}”). The beam line configurations for two passively scattered proton beams used in the actual treatment were extracted from the clinical treatment plan and a set of input files was created automatically using MC{sup 2}. A full MC simulation of the beam line was computed using MCNPX and a set of phase space files for each beam was collected at the distal surface of the range compensator. The particles from these phase space files were transported through the 10 voxelized patient models corresponding to the 10 phases of the breathing cycle in the 4DCT, using MCNPX and an accelerated (fast) MC code called “FDC”, developed by us and which is based on the track repeating algorithm. The accuracy of the fast algorithm was assessed by comparing the two time dependent dose distributions. Results: The error of less than 1% in 100% of the voxels in all phases of the breathing cycle was achieved using this method with a speedup of more than 1000 times. Conclusion: The proposed method, which uses full MC to simulate the beam line and the accelerated MC code FDC for the time consuming particle transport inside the complex, time dependent, geometry of the patient shows excellent accuracy together with an extraordinary speed.

  4. TH-A-18C-03: Noise Correlation in CBCT Projection Data and Its Application for Noise Reduction in Low-Dose CBCT

    SciTech Connect (OSTI)

    ZHANG, H; Huang, J; Ma, J; Chen, W; Ouyang, L; Wang, J

    2014-06-15

    Purpose: To study the noise correlation properties of cone-beam CT (CBCT) projection data and to incorporate the noise correlation information to a statistics-based projection restoration algorithm for noise reduction in low-dose CBCT. Methods: In this study, we systematically investigated the noise correlation properties among detector bins of CBCT projection data by analyzing repeated projection measurements. The measurements were performed on a TrueBeam on-board CBCT imaging system with a 4030CB flat panel detector. An anthropomorphic male pelvis phantom was used to acquire 500 repeated projection data at six different dose levels from 0.1 mAs to 1.6 mAs per projection at three fixed angles. To minimize the influence of the lag effect, lag correction was performed on the consecutively acquired projection data. The noise correlation coefficient between detector bin pairs was calculated from the corrected projection data. The noise correlation among CBCT projection data was then incorporated into the covariance matrix of the penalized weighted least-squares (PWLS) criterion for noise reduction of low-dose CBCT. Results: The analyses of the repeated measurements show that noise correlation coefficients are non-zero between the nearest neighboring bins of CBCT projection data. The average noise correlation coefficients for the first- and second- order neighbors are about 0.20 and 0.06, respectively. The noise correlation coefficients are independent of the dose level. Reconstruction of the pelvis phantom shows that the PWLS criterion with consideration of noise correlation (PWLS-Cor) results in a lower noise level as compared to the PWLS criterion without considering the noise correlation (PWLS-Dia) at the matched resolution. Conclusion: Noise is correlated among nearest neighboring detector bins of CBCT projection data. An accurate noise model of CBCT projection data can improve the performance of the statistics-based projection restoration algorithm for low-dose CBCT.

  5. TH-A-19A-08: Intel Xeon Phi Implementation of a Fast Multi-Purpose Monte Carlo Simulation for Proton Therapy

    SciTech Connect (OSTI)

    Souris, K; Lee, J; Sterpin, E

    2014-06-15

    Purpose: Recent studies have demonstrated the capability of graphics processing units (GPUs) to compute dose distributions using Monte Carlo (MC) methods within clinical time constraints. However, GPUs have a rigid vectorial architecture that favors the implementation of simplified particle transport algorithms, adapted to specific tasks. Our new, fast, and multipurpose MC code, named MCsquare, runs on Intel Xeon Phi coprocessors. This technology offers 60 independent cores, and therefore more flexibility to implement fast and yet generic MC functionalities, such as prompt gamma simulations. Methods: MCsquare implements several models and hence allows users to make their own tradeoff between speed and accuracy. A 200 MeV proton beam is simulated in a heterogeneous phantom using Geant4 and two configurations of MCsquare. The first one is the most conservative and accurate. The method of fictitious interactions handles the interfaces and secondary charged particles emitted in nuclear interactions are fully simulated. The second, faster configuration simplifies interface crossings and simulates only secondary protons after nuclear interaction events. Integral depth-dose and transversal profiles are compared to those of Geant4. Moreover, the production profile of prompt gammas is compared to PENH results. Results: Integral depth dose and transversal profiles computed by MCsquare and Geant4 are within 3%. The production of secondaries from nuclear interactions is slightly inaccurate at interfaces for the fastest configuration of MCsquare but this is unlikely to have any clinical impact. The computation time varies between 90 seconds for the most conservative settings to merely 59 seconds in the fastest configuration. Finally prompt gamma profiles are also in very good agreement with PENH results. Conclusion: Our new, fast, and multi-purpose Monte Carlo code simulates prompt gammas and calculates dose distributions in less than a minute, which complies with clinical time constraints. It has been successfully validated with Geant4. This work has been financialy supported by InVivoIGT, a public/private partnership between UCL and IBA.

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

    Office of Scientific and Technical Information (OSTI)

    1, Ohio K TABLE OF CONTENTS Page ABSTRACT 1 INTRODUCTION 1 PART I. COMPOSITION OF ... p r e s s u r e of pure component Subscript 1 H2O Subscript 2 HNO3. The p a r t i a l ...

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

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Its major applications as designed originally were nuclear power plant loss of coolant accident and other anticipated operational transients. * In its current form, it solves the ...

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

    Office of Scientific and Technical Information (OSTI)

    The storage tank emptied when the liquid-phase excess flow valve tore out of the tank. The ensuing fire engulfed several propane delivery trucks, causing one of them to explode. A ...

  9. Pressure suppression containment system for boiling water reactor

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Nesbitt, Loyd B.

    1997-01-01

    A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

  10. Pressure suppression containment system for boiling water reactor

    DOE Patents [OSTI]

    Gluntz, D.M.; Nesbitt, L.B.

    1997-01-21

    A system is disclosed for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs. 3 figs.

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

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    More Documents & Publications Integrated External Aerodynamic and Underhood Thermal Analysis for Heavy Vehicles CRADA with PACCAR Experimental Investigation in...

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

    SciTech Connect (OSTI)

    Durbin, Samuel; Lindgren, Eric R.

    2015-11-01

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

  13. SOLAR PROMINENCES: “DOUBLE, DOUBLE… BOIL AND BUBBLE”

    SciTech Connect (OSTI)

    Keppens, R.; Xia, C.; Porth, O.

    2015-06-10

    Observations revealed rich dynamics within prominences, the cool (10{sup 4} K), macroscopic (sizes of order 100 Mm) “clouds” in the million degree solar corona. Even quiescent prominences are continuously perturbed by hot, rising bubbles. Since prominence matter is hundredfold denser than coronal plasma, this bubbling is related to Rayleigh–Taylor instabilities. Here we report on true macroscopic simulations well into this bubbling phase, adopting an MHD description from chromospheric layers up to 30 Mm height. Our virtual prominences rapidly establish fully nonlinear (magneto)convective motions where hot bubbles interplay with falling pillars, with dynamical details including upwelling pillars forming within bubbles. Our simulations show impacting Rayleigh–Taylor fingers reflecting on transition region plasma, ensuring that cool, dense chromospheric material gets mixed with prominence matter up to very large heights. This offers an explanation for the return mass cycle mystery for prominence material. Synthetic views at extreme ultraviolet wavelengths show remarkable agreement with observations, with clear indications of shear-flow induced fragmentations.

  14. Critical heat flux for free convection boiling in thin rectangular...

    Office of Scientific and Technical Information (OSTI)

    DOE Contract Number: AC02-76CH00016 Resource Type: Conference Resource Relation: Conference: ASMEAIChEANS national heat transfer conference, Minneapolis, MN (United States), ...

  15. Visual observation of boiling in high power liquid target

    SciTech Connect (OSTI)

    Peeples, J. L.; Stokely, M. H.; Poorman, M. C.; Magerl, M.; Wieland, B. W.

    2012-12-19

    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.

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

    Office of Scientific and Technical Information (OSTI)

    ... S O N O F EQUATION 11-17 WITH ZMOLA'S (42) 51 11-4. M a - t & n m Bubble Diameter and the ... u i d i s given by: AXs Consider a surface area A, and a A T d e p e e s the i n t e r n a ...

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

    Office of Scientific and Technical Information (OSTI)

    ... Concentrated nitric acid may react with ethanol to form ethyl nitrate which is highly ... By adding ethanol to the sample as suggested by the work of L a m b , Carlton, and M e l d ...

  18. Subcooled Boiling Heat Transfer for Cooling of Power Electronics...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    radiator and associated pumping system are still required in HEVs. This additional cooling system adds weight and cost while decreasing the efficiency of HEVs. With the...

  19. Self-Sustaining Thorium Boiling Water Reactors (Technical Report...

    Office of Scientific and Technical Information (OSTI)

    of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; ... Language: English Subject: 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; ...

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

    Office of Scientific and Technical Information (OSTI)

    ... 3 ethanol 4 pen- 5 pentam 6 heptane 7 Propans 8 propane 9 water 10 bornens FIGURE 111-3. QRELQTION OF DATA FOR VARIOUS LIQUXW AT THE CFUTICAL HEAT FLUX DEIWTY IN POOL B O - . ...

  1. Apparatus for pumping liquids at or below the boiling point

    DOE Patents [OSTI]

    Bingham, Dennis N.

    2002-01-01

    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.

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

    Energy.gov [DOE]

    2010 DOE Vehicle Technologies and Hydrogen Programs Annual Merit Review and Peer Evaluation Meeting, June 7-11, 2010 -- Washington D.C.

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

    SciTech Connect (OSTI)

    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)

    2009-09-15

    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)

  4. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOE Patents [OSTI]

    Oosterkamp, Willem Jan; Marquino, Wayne

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereat access to the fuel assemblies is not obstructed.

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

    DOE Patents [OSTI]

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

    1998-02-24

    A bottom head to shell junction assembly which, in one embodiment, includes an annular forging having an integrally formed pump deck and shroud support is described. In the one embodiment, the annular forging also includes a top, cylindrical shaped end configured to be welded to one end of the pressure vessel cylindrical shell and a bottom, conical shaped end configured to be welded to the disk shaped bottom head. Reactor internal pump nozzles also are integrally formed in the annular forging. The nozzles do not include any internal or external projections. Stubs are formed in each nozzle opening to facilitate welding a pump housing to the forging. Also, an upper portion of each nozzle opening is configured to receive a portion of a diffuser coupled to a pump shaft which extends through the nozzle opening. Diffuser openings are formed in the integral pump deck to provide additional support for the pump impellers. The diffuser opening is sized so that a pump impeller can extend at least partially therethrough. The pump impeller is connected to the pump shaft which extends through the nozzle opening. 5 figs.

  6. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOE Patents [OSTI]

    Oosterkamp, W.J.; Marquino, W.

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies is disclosed. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereas access to the fuel assemblies is not obstructed. 11 figs.

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

    SciTech Connect (OSTI)

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

    1980-06-01

    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.

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

    DOE Patents [OSTI]

    Fife, Alex Blair; Ballas, Gary J.

    1998-01-01

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

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

    Energy.gov [DOE]

    2011 DOE Hydrogen and Fuel Cells Program, and Vehicle Technologies Program Annual Merit Review and Peer Evaluation

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

    DOE Patents [OSTI]

    Gluntz, Douglas M.

    1996-01-01

    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.

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

    SciTech Connect (OSTI)

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

    2007-11-01

    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.

  12. Comparison of Calculated and Experimental Results for a Boiling/Condensing Experimental Facility

    SciTech Connect (OSTI)

    Carbajo, Juan J; McDuffee, Joel Lee; Felde, David K

    2016-01-01

    A new experimental facility for materials irradiation and testing at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is being developed. Details of this facility have been presented before [1, 2]. A prototype of this facility, the Thermo-Syphon Test Loop (TSTL) has been built and experimental data have been obtained and analyzed [3, 4]. Pre-test calculations for this facility with the RELAP5-3D code [5] have been presented previously [6] as well as other calculations [7, 8] with the TRACE code [9]. The results of both codes were very different [7]. RELAP5-3D predicted much higher pressures and temperatures than TRACE. This paper compares calculated results with the TSTL experimental data.

  13. Method for controlling boiling point distribution of coal liquefaction oil product

    DOE Patents [OSTI]

    Anderson, Raymond P.; Schmalzer, David K.; Wright, Charles H.

    1982-12-21

    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.

  14. Method for controlling boiling point distribution of coal liquefaction oil product

    DOE Patents [OSTI]

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

    1982-12-21

    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.

  15. Blanket Module Boil-Off Times during a Loss-of-Coolant Accident - Case 0: with Beam Shutdown only

    SciTech Connect (OSTI)

    Hamm, L.L.

    1998-10-07

    This report is one of a series of reports that document LBLOCA analyses for the Accelerator Production of Tritium primary blanket Heat Removal system. This report documents the analysis results of a LBLOCA where the accelerator beam is shut off without primary pump trips and neither the RHR nor the cavity flood systems operation.

  16. Vehicle Technologies Office Merit Review 2015: Thermal Control of Power Electronics of Electric Vehicles with Small Channel Coolant Boiling

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation given by Argonne National Laboratory at 2015 DOE Hydrogen and Fuel Cells Program and Vehicle Technologies Office Annual Merit Review and Peer Evaluation Meeting about thermal control...

  17. On the Modeling of Local Neutronically-Coupled Flow-Induced Oscillations in Advanced Boiling Water Reactors

    SciTech Connect (OSTI)

    Aniel-Buchheit, Sylvie; Podowski, Michael Z.

    2006-07-01

    The purpose of this paper is to discuss the development in progress of a complete space- and time-dependent model of the coupled neutron kinetic and reactor thermal-hydraulics. The neutron kinetics model is based on two-group diffusion equations with Doppler and void reactivity feedback effects. This model is coupled with the model of two-phase flow and heat transfer in parallel coolant channels. The modeling concepts considered for this purpose include one-dimensional drift flux and two-fluid models, as well a CFD model implemented in the NPHASE advanced computational multiphase fluid dynamics (CMFD) computer code. Two methods of solution for the overall model are proposed. One is based on direct numerical integration of the spatially-discretized governing equations. The other approach is based on a quasi-analytical modal approach to the neutronics model, in which a complete set of eigenvectors is found for step-wise temporal changes of the cross-sections of core materials (fuel and coolant/moderator). The issues investigated in the paper include details of model formulation, as well as the results of calculations for neutronically-coupled density-wave oscillations. (authors)

  18. Corrosion and stress corrosion cracking of type 304 stainless steel and carbon steel in simulated boiling water reactor

    SciTech Connect (OSTI)

    Choi, H.J.

    1981-01-01

    It was found that A508 C1.2 steel undergoes corrosion cracking in pure water containing 1 or 8 ppM of oxygen at tempertures ranging from 100 to 288 C. At temperatures of 100 and 150 C, cracks nucleate at corrosion pits. At higher temperatures, however, cracks nucleate beneath hematite crystals which grow via dissolution-precipitation upon a base oxide film at sites of high dissolution. The susceptibility to SCC increase with increasing oxygen concentration but passes through a maximum as a function of temperature at 250 C. The susceptibility of sensitized Type 304 stainless steel and ASTM A508 C1.2 steel to SCC in high temperature oxygenated water, as determined using constant extension rate tests, is found to depend upon the velocity of the fluid flow past the gauge section. This work has demonstrated that in both cases, the failure time increases with increasing flow rate and that both the crack initiation time and the apparent crack propagation rate depend upon the flow velocity. The longer failure time is mainly attributed to an increase in the crack initiation time. The smaller crack propagation rate during the initial crack propagation period is controlled by the physical and electrochemical factors but the higher crack propagation rate results from the fact that as initiation is delayed a higher stress intensity exists at the crack tip.

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

    SciTech Connect (OSTI)

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. ); Wagner, K.C. )

    1990-07-01

    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.

  20. Analysis of high pressure boil-off situation during MSIV closure ATWS in a typical BWR/4

    SciTech Connect (OSTI)

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

    1986-01-01

    The objective of this paper is to provide a best-estimate analysis of the MSIV Closure ATWS in the Browns Ferry Unit 1 BWR with Mark 1 containment. The calculations have been performed using the RAMONA-3B code which has a three-dimensional neutron kinetics model coupled with one-dimensional (multi-channel core representation), four-equation, nonhomogeneous, nonequilibrium thermal hydraulics. The code also allows for one-dimensional neutronic core representation. The 1-D capability of the code has been employed in this calculation since a thorough sensitivity study showed that for a full ATWS, a one-dimensional (axial) neutron kinetics adequately describes the core behavior. (Note that the core steady-state symmetry in this case was preserved throughout the transient so that radial effects could be neglected.) The calculation described in the paper was started from a steady-state fuel condition corresponding to the end of Cycle 5 of the Browns Ferry reactor.

  1. L3: VUQ.V&V.P9.04 Cobra-TF Parameter Exposure Work Vincent Mousseau...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    ... partitioning of the boiling and condensation mass transfer between the liquid ... Nucleate Boiling CTF calculates the heat transfer coefficient in the subcooled nucleate boiling ...

  2. Oak Ridge

    Office of Legacy Management (LM)

    Dust samples were coflected from the floor and ledges in tha East bay. Direct measurements ... surfaces (floors, ledges, and overhead pipes) in tha 5ast bay be cleaned of residual dust. ...

  3. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    SciTech Connect (OSTI)

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  4. Study on critical heat flux enhancement in flow boiling of SiC nano-fluids under low pressure and low flow conditions

    SciTech Connect (OSTI)

    Lee, S. W.; Park, S. D.; Kang, S.; Kim, S. M.; Seo, H.; Lee, D. W.; Bang, I. C.

    2012-07-01

    Critical heat flux (CHF) is the thermal limit of a phenomenon in which a phase change occurs during heating (such as bubbles forming on a metal surface used to heat water), which suddenly decreases the heat transfer efficiency, thus causing localized overheating of the heating surface. The enhancement of CHF can increase the safety margins and allow operation at higher heat fluxes; thus, it can increase the economy. A very interesting characteristics of nano-fluids is their ability to significantly enhance the CHF. nano-fluids are nano-technology-based colloidal dispersions engineered through stable suspending of nanoparticles. All experiments were performed in round tubes with an inner diameter of 0.01041 m and a length of 0.5 m under low pressure and low flow (LPLF) conditions at a fixed inlet temperature using water, 0.01 vol. % Al{sub 2}O{sub 3}/water and SiC/water nano-fluids. It was found that the CHF of the nano-fluids was enhanced and the CHF of the SiC/water nano-fluid was more enhanced than that of the Al{sub 2}O{sub 3}/water nano-fluid. (authors)

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

    SciTech Connect (OSTI)

    Not Available

    1994-01-15

    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.

  6. Estimating boiling water reactor decommissioning costs. A user`s manual for the BWR Cost Estimating Computer Program (CECP) software: Draft report for comment

    SciTech Connect (OSTI)

    Bierschbach, M.C.

    1994-12-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the U.S. Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personal computer, provides estimates for the cost of decommissioning BWR power stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  7. Thailand: Energy Resources | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Country Profile Name Thailand Population Unavailable GDP Unavailable Energy Consumption Quadrillion Btu 2-letter ISO code TH 3-letter ISO code THA Numeric ISO code...

  8. Introduction to computed microtomography and applications in...

    Office of Scientific and Technical Information (OSTI)

    Resource Relation: Related Information: CMS Workshop Lectures, Advanced Applications of Synchrotron Radiation in Clay Science Publisher: 2014; Tha Clay Minerals Society ;Urbana, IL ...

  9. California Institute of Technology Caltech | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    of higher learning tha investigates the most challenging, fundamental problems in science and technology. Coordinates: 29.690847, -95.196308 Show Map Loading map......

  10. State Nuclear Profiles 2010

    U.S. Energy Information Administration (EIA) (indexed site)

    Data for 2010 BWR Boiling Water Reactor. Source: Form EIA-860, "Annual Electric ... Data for 2010 BWR Boiling Water Reactor. Source: Form EIA-860, "Annual Electric ...

  11. Microsoft Word - TR07-27.doc

    Office of Legacy Management (LM)

    Boiling Nuclear Superheat (BONUS), Site, Rincn, Puerto Rico July 2010 Page 1 2010 Inspection and Status Report for the Former Boiling Nuclear Superheater (BONUS) Reactor...

  12. Flourescent Pigments for High-Performance Cool Roofing and Facades...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    After boiling and then combustion, a fluorescent red pigment is formed. Image: University ... After boiling and then combustion, a fluorescent red pigment is formed. Image: University ...

  13. Fluorescent Pigments for High-Performance Cool Roofing and Facades...

    Energy Savers

    After boiling and then combustion, a fluorescent red pigment is formed. Image: University ... After boiling and then combustion, a fluorescent red pigment is formed. Image: University ...

  14. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    SciTech Connect (OSTI)

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  15. 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 (OSTI)

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

    1996-07-01

    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.

  16. SSRL Accelerator Phycics Home Page

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    at.gif (15297 bytes) BeamOptics.gif (29047 bytes) ICFA2000t.gif (31362 bytes) Home Page LCLS Accelerator Physics at SSRL The field tha t can be covered by the Accelerator Physics...

  17. Building America Puts Residential Research Results To Work; Building America Research That Works (Fact Sheet)

    SciTech Connect (OSTI)

    2009-01-18

    Residential buildings use more than 20% of the energy consumed annually in the United States. To help reduce that energy use, the Department of Energy (DOE) and its Building America partners conduct research to develop advanced building energy systems tha

  18. Indigenous Sustainability

    Energy Savers

    Indigenous S ustainability Image by Jonathan Thunder Climate Change Impacts Tar Sands 86% Food Economy Food Dollars Spent On the Food Dollars Spent On the "the staEsEcs showed tha ...

  19. MASSACHUSETTS AVENUE ChMBRIDGE'39, MASSACHUSETTS TELEPHONE UNrvn...

    Office of Legacy Management (LM)

    Strauch: With reference to Mr. K. E. Field's confidential memorandum of August 22, 1956, this is to advise tha.t Nuclea,r l,':etals, Inc., has no facilities for scrap recovery. ...

  20. TO J. A. QuigUy, M.D. NATIONALLPADCW~

    Office of Legacy Management (LM)

    DcPaaio *that further teats bb conducted for tha plocporre of deter- mining*ether sonic energy cleaning will reduce the alpha radiatiar couut within the limits set forth by BBC". ...

  1. WHC-MR-0519 WHC Significant Lessons Learned

    Office of Scientific and Technical Information (OSTI)

    e or reflect those of tha United Stntss G o v e m n t or any ... Chapter 4 of the DOE Radiological Control Manual states: ... method to control the rate o f steam introduction( ...

  2. Eh:2,' %9'j-& : : _.i

    Office of Legacy Management (LM)

    at roan tmqmatwe with a +HP (Torrington size 6) eager and two piece die. ihendons and ... Tbe Torrbqtan no- 4 machine nith 5-S motor sould not be prite laqe enough to do tha.job of ...

  3. Cosan | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    subsidiary of Cosan Limited. Tha company is one of the largest ethanol producers in the world. References: Cosan1 This article is a stub. You can help OpenEI by expanding it....

  4. NREL: Learning - Concentrating Solar Power Basics

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    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 electricity. ...

  5. Minimizing corrosion in coal liquid distillation

    DOE Patents [OSTI]

    Baumert, Kenneth L.; Sagues, Alberto A.; Davis, Burtron H.

    1985-01-01

    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.

  6. Concentrating solar power | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    the southwestern United States and other sunbelts worldwide. Many power plants today use fossil fuels as a heat source to boil water. The steam from the boiling water spins a...

  7. EIA - State Nuclear Profiles

    Gasoline and Diesel Fuel Update

    84.9 BWR 7281975 10172034 855 6,361 84.9 Data for 2010 BWR Boiling Water Reactor. ... The 900-acre site is also the location of two other General Electric boiling water ...

  8. Supercritical CO2-Brayton Cycle

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    Nuclear Energy Defense Waste Management Programs Advanced Nuclear Energy Nuclear Energy Safety Technologies Facilities Battery Abuse Testing Laboratory Cylindrical Boiling ...

  9. Renewable Energy and Distributed Systems Integration

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    Nuclear Energy Defense Waste Management Programs Advanced Nuclear Energy Nuclear Energy Safety Technologies Facilities Battery Abuse Testing Laboratory Cylindrical Boiling ...

  10. Consortium for Advanced Simulation of Light Water Reactors (CASL...

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    Nuclear Energy Defense Waste Management Programs Advanced Nuclear Energy Nuclear Energy Safety Technologies Facilities Battery Abuse Testing Laboratory Cylindrical Boiling ...

  11. Pratt Whitney Rocketdyne Testing

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    Battery Abuse Testing Laboratory Cylindrical Boiling Facility Distributed Energy Technology Lab Microsystems and Engineering Sciences Applications National Solar Thermal Test ...

  12. Sandia's Continuously Recirculating Falling-Particle Receiver...

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    Battery Abuse Testing Laboratory Cylindrical Boiling Facility Distributed Energy Technology Lab Microsystems and Engineering Sciences Applications National Solar Thermal Test ...

  13. Sandia Researchers Win CSP:ELEMENTS Funding Award

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    Battery Abuse Testing Laboratory Cylindrical Boiling Facility Distributed Energy Technology Lab Microsystems and Engineering Sciences Applications National Solar Thermal Test ...

  14. MELCOR Model of the Spent Fuel Pool of Fukushima Dai-ichi Unit...

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    ALUMINIUM; BOILING; DIMENSIONS; EARTHQUAKES; EXPLOSIONS; FUEL ASSEMBLIES; FUEL RACKS; HYDROGEN; NUCLEAR POWER PLANTS; OXIDATION; OXYGEN; RADIOISOTOPES; REACTOR ACCIDENTS;...

  15. L3:PHI.VCS.P9.02 CTF Validation Robert K. Salko Oak Ridge National...

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    ... . . . . . . . . . 62 4.6 Heat Transfer . . . . . . . . . . . . ... BFBT cases with loss coefficient for each channel as ... (evaporationcondensation) and boiling models in ...

  16. Direct production of fractionated and upgraded hydrocarbon fuels from biomass

    DOE Patents [OSTI]

    Felix, Larry G.; Linck, Martin B.; Marker, Terry L.; Roberts, Michael J.

    2014-08-26

    Multistage processing of biomass to produce at least two separate fungible fuel streams, one dominated by gasoline boiling-point range liquids and the other by diesel boiling-point range liquids. The processing involves hydrotreating the biomass to produce a hydrotreatment product including a deoxygenated hydrocarbon product of gasoline and diesel boiling materials, followed by separating each of the gasoline and diesel boiling materials from the hydrotreatment product and each other.

  17. The dynamics of two-phase (gas

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    dynamics of two-phase (gas/ liquid) bubbly flows are complex: bubbles deform and disperse; large latent heats and heat capacity differ- entials influence local boiling; and relatively small changes in heated surface temperatures yield order of magnitude changes in boiling com- plexity. Because the local void vol- ume has a direct feedback effect on reactor neutron flux and fuel rod power production, prediction of local boiling rates and bulk boiling effects in nuclear reactors is key in achiev-

  18. Consortium for Advanced Simulation of Light Water Reactors

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    An essential part of developing a closed form set of equations (closures) for prediction of two-phase flow with computational fluid dynamics (CFD) is understanding how the bubbles generat- ed by boiling interact. An accurate prediction of moderator and fuel performance once boiling has begun is needed to simulate CASL Challenge Problems related to boiling water reactors (BWRs), departure from nucleate boiling (DNB) behavior in pressurized water reactors (PWRs), loss of coolant accidents (LOCAs),

  19. E P GPT collectively denotes new ...

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    ... different, the research team postulated that ... an application of advanced statistical modeling ... Boiling Flows, International Conference on Mathematics and Computational ...

  20. CASL - Lift Forces in Bubbly Flows

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Lift Forces in Bubbly Flows The dynamics of two-phase (gas/liquid) bubbly flows are complex: bubbles deform and disperse; large latent heats and heat capacity differentials influence local boiling; and relatively small changes in heated surface temperatures yield order of magnitude changes in boiling complexity. Because the local void volume has a direct feedback effect on reactor neutron flux and fuel rod power production, prediction of local boiling rates and bulk boiling effects in nuclear

  1. Coal liquefaction process with increased naphtha yields

    DOE Patents [OSTI]

    Ryan, Daniel F.

    1986-01-01

    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.

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

    SciTech Connect (OSTI)

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

    2008-10-01

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

  3. I:

    Office of Legacy Management (LM)

    : 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

  4. CASL-U-2015-0113-000 RPI Milestone:

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    13-000 RPI Milestone: Development of a Mechanistic Subcooled Boiling Model for PWR Assemblies M.Z. Podowski Rensselaer Polytechnic Institute (RPI) August 31, 2014 CASL-U-2015-0113-000 August 31, 2014 RPI Milestone: Development of a mechanistic subcooled boiling model for PWR assemblies 1 TOP FOCUS AREA ACHIEVEMENTS * Top achievement 1 - The formulation of a mechanistic multidimensional model of vapor condensation in subcooled boiling. The new model allows to separately capture each: the

  5. CASL-U-2015-0253-000

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    3-000 Treatment of Nucleation and Bubble Dynamics in High Heat Flux Boiling Yang Liu, Nam Dinh North Carolina State University July 7, 2015 CASL-U-2015-0253-000 TREATMENT OF NUCLEATION AND BUBBLE DYNAMICS IN HIGH HEAT FLUX BOILING Yang Liu, Department of Nuclear Engineering, North Carolina State University Instructor: Dr. Nam Dinh, Department of Nuclear Engineering, North Carolina State University Nucleate boiling is a highly efficient and desirable cooling mechanism in high- power-density

  6. MEMS sensor measurement of surface temperature response during subcooled

    Office of Scientific and Technical Information (OSTI)

    flow boiling in a rectangular flow channel (Journal Article) | DOE PAGES MEMS sensor measurement of surface temperature response during subcooled flow boiling in a rectangular flow channel Title: MEMS sensor measurement of surface temperature response during subcooled flow boiling in a rectangular flow channel Authors: Yabuki, Tomohide Search DOE PAGES for author "Yabuki, Tomohide" Search DOE PAGES for ORCID "0000000163741912" Search orcid.org for ORCID

  7. Vehicle Technologies Office Merit Review 2014: Thermal Control of Power

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Electronics of Electric Vehicles with Small Channel Coolant Boiling | Department of Energy Thermal Control of Power Electronics of Electric Vehicles with Small Channel Coolant Boiling Vehicle Technologies Office Merit Review 2014: Thermal Control of Power Electronics of Electric Vehicles with Small Channel Coolant Boiling Presentation given by Argonne National Laboratory at 2014 DOE Hydrogen and Fuel Cells Program and Vehicle Technologies Office Annual Merit Review and Peer Evaluation

  8. 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 (OSTI)

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

    1994-09-01

    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.

  9. Microsoft Word - NURETH-14_Paper_397_rev June 15- no review track...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    NUCLEATE POOL BOILING X. Duan, J. Buongiorno and T. McKrell Massachusetts Institute of Technology, Cambridge, Massachusetts, USA Abstract High-resolution data of nucleate pool...

  10. Mutton Rogan Josh | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    boil mutton Ingredients (this section does not need to be done here) Ingredient Alternative Mutton Haldi Ginger Garlic Paste Chicken none Ginger Garlic grinded Instructions...

  11. EXPERIMENTAL AND THEORETICAL DETERMINATION OF HEAVY OIL VISCOSITY...

    Office of Scientific and Technical Information (OSTI)

    Language: English Subject: 02 PETROLEUM; 04 OIL SHALES AND TAR SANDS; API GRAVITY; BOILING POINTS; MOLECULAR WEIGHT; OIL SANDS; PETROLEUM; PHYSICAL PROPERTIES; STEAM; TESTING; ...

  12. Office of Science

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    Nuclear Energy Defense Waste Management Programs Advanced Nuclear Energy Nuclear Energy Safety Technologies Facilities Battery Abuse Testing Laboratory Cylindrical Boiling Facility ...

  13. National Supervisory Control and Data Acquisition (SCADA)

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  14. NSTTF

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  15. Photovoltaic Systems Evaluation Laboratory (PSEL)

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  16. Partnership

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  17. Test Site Operations & Maintenance Safety

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  18. CRF

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  19. Materials Science

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  20. Highlights - Energy Research

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  1. CINT

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  2. Photovoltaic Regional Testing Center (PV RTC)

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    Nuclear Energy Defense Waste Management Programs Advanced Nuclear Energy Nuclear Energy Safety Technologies Facilities Battery Abuse Testing Laboratory Cylindrical Boiling Facility ...

  3. Facilities

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  4. Research & Capabilities

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  5. Biomass

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  6. Fuel Options

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  7. Biofuels

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  8. Grid Integration

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  9. SunShot

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  10. Systems Engineering

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  11. Safety, Security & Resilience of Energy Infrastructure

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  12. News

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  14. Facilities, Partnerships, and Resources

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  16. Sensors & Optical Diagnostics

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  17. Infrastructure Security

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  18. Solar Newsletter

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  19. National Solar Thermal Test Facility

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  20. Analysis

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  1. Systems Analysis

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  2. JBEI

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  4. Phenomenological Modeling

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  6. Request for Testing

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  7. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    ... a finite line source, a series of point sources, and a series of parallel infinite line ... Self-Sustaining Thorium Boiling Water Reactors Greenspan, Ehud ; Gorman, Phillip M. ; ...

  8. U.S. Spent Nuclear Fuel Data as of December 31,2002 Table 3

    Gasoline and Diesel Fuel Update

    permanently discharged in previous years, the historical totals change. BWR Boiling-water reactor; PWR Pressurized-water reactor; HTGR High-temperature gas cooled reactor....

  9. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    ... Center, Morgantown, WV (United States) Mound Area Office, Miamisburg, OH (United ... Self-Sustaining Thorium Boiling Water Reactors Greenspan, Ehud ; Gorman, Phillip M. ; ...

  10. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    ... between, RPE cells influences growth factor expression leading to the initiation and ... control over heat transfer in high power density systems utilizing boiling phenomena. ...

  11. Nanolubricants to Improve Chiller Performance

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Technologies Office eere.energy.gov Approach * Pool-boiling rig used to measure heat transfer performance of R134ananolubricant mixtures with nanoparticle of varied ...

  12. State Nuclear Profiles 2010

    U.S. Energy Information Administration (EIA) (indexed site)

    Data for 2010 PWR Pressurized Light Water Reactor. Source: Form EIA-860, "Annual ... Data for 2010 BWR Boiling Water Reactor. Source: Form EIA-860, "Annual Electric ...

  13. State Nuclear Profiles 2010

    U.S. Energy Information Administration (EIA) (indexed site)

    Data for 2010 BWR Boiling Water Reactor. Source: Form EIA-860, "Annual Electric Generator Report," and Form EIA-923, "Power Plant Operations Report." Type Commercial operation ...

  14. State Nuclear Profiles 2010

    U.S. Energy Information Administration (EIA) (indexed site)

    Data for 2010 BWR Boiling Water Reactor. Source: Form EIA-860, "Annual Electric ... Data for 2010 PWR Pressurized Light Water Reactor. Source: Form EIA-860, "Annual ...

  15. Sandia Energy - First-Ever Asian MELCOR User Group Meeting

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    nuclear power plants. Sandia developed MELCOR for the US Nuclear Regulatory Commission (NRC)-to treat a broad spectrum of phenomena in both boiling and pressurized water reactors...

  16. DOE - Office of Legacy Management -- Bonus

    Office of Legacy Management (LM)

    Puerto Rico Boiling Nuclear Superheater (BONUS), Puerto Rico, Decommissioned Reactor Site Key Documents and Links All documents are Adobe Acrobat files. pdficon Key Documents Fact...

  17. BONUS, Puerto Rico, Decommissioned Reactor Site Fact Sheet

    Office of Legacy Management (LM)

    information about the Defense Decontamination and Decommissioning Program Boiling Nuclear Superheater (BONUS) reactor located northwest of Rincn, Puerto Rico. The site is...

  18. Interim Action Determination Flexible ...

    Office of Environmental Management (EM)

    ... be modified to allow for processing fuel pellets and fuel rods of different diameters and ... Both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel pellets, rods ...

  19. DOE - Office of Legacy Management -- Bonus

    Office of Legacy Management (LM)

    of the DOE Defense Decontamination and Decommissioning (D&D) Program, the Office of Legacy Management manages the Boiling Nuclear Superheater (BONUS) Decommissioned Reactor Site...

  20. New research, publications and videos

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    boiling can result in undesirable local hotspots and the accumulation of excess corrosion and contaminants. A simulation demonstrates the volume fraction of a bubble phase in...

  1. State Nuclear Profiles 2010

    U.S. Energy Information Administration (EIA) (indexed site)

    Data for 2010 BWR Boiling Water Reactor. Source: Form EIA-860, "Annual Electric Generator Report," and Form EIA-923, "Power Plant Operations Report." Type Commercial operation date ...

  2. Sambar | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Ingredients (this section does not need to be done here) Ingredient Alternative Bengal Gram water sambar Powder Turmeric Powder Any Dal alternative Instructions Boil Bengal...

  3. Fluid-inclusion evidence for past temperature fluctuations in...

    Open Energy Information (Open El) [EERE & EIA]

    of the inclusion fluids range from dilute meteoric water to highly modified sea water concentrated by boiling. Comparison of measured drill-hole temperatures with...

  4. Search results | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    groups conduct an investigation into the similarities and differences between solar tea and tea brewed by boiling water. Students will compare their two samples on four...

  5. Inverted fractionation apparatus and use in a heavy oil catalytic...

    Office of Scientific and Technical Information (OSTI)

    cycle oil boiling range hydrocarbons and mixtures thereof into liquid product fractions, ... Subject: 02 PETROLEUM; PETROLEUM; CATALYTIC CRACKING; PETROLEUM FRACTIONS; VISCOSITY; ...

  6. Brunswick County, North Carolina: Energy Resources | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    Island, North Carolina Belville, North Carolina Boiling Spring Lakes, North Carolina Bolivia, North Carolina Calabash, North Carolina Carolina Shores, North Carolina Caswell...

  7. Final Surplus Plutonium Disposition Supplemental Environmental...

    Energy Savers

    ... and new information based in part on comments ... and Integrated Extraction System BWR boiling water reactor CAM continuous ... for further examination in an offsite hot cell. ...

  8. DOE/EIS-0283-S2

    Energy Savers

    ... and new information based in part on comments ... Extraction System Browns Ferry Browns Ferry Nuclear Plant BMP best management practice BWR boiling water reactor ...

  9. Preliminary Technology Readiness Assessment (TRA) for the Calcine...

    Office of Environmental Management (EM)

    ... Hot isostatic pressing A manufacturing process that subjects ... repository requirements based on analyses conducted by ... when heated to remove water and other low boiling volatiles. ...

  10. Nureth-15 paper

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    surface-averaged value), heat transfer coefficient, nucleation site density, and bubble wait time. ... convection, nucleate boiling, condensation, sliding bubbles), two phase flow ...

  11. Nureth-15 paper

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    change, both near the wall (boiling) and far in the core flow (condensation). ... with global parameters including the heat transfer coefficient, the onset of nucleation, the ...

  12. Microsoft PowerPoint - 7_LAUREN_GIBSON_NMMSS Presentation Gibson...

    National Nuclear Security Administration (NNSA)

    Venting System Applies to boiling water reactors with certain designs (Mark III) ... Requires installation of water level instrumentation to indicate: 1 - Normal fuel pool ...

  13. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    CASL's Latest Research CIPS Simulation Capability Implemented in VERA Posted on October 28, 2015 Departure from Nucleate Boiling (DNB) Multi-Physics Approach & Applications using...

  14. Macon County, Tennessee: Energy Resources | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Number 4 Climate Zone Subtype A. Places in Macon County, Tennessee Lafayette, Tennessee Red Boiling Springs, Tennessee Retrieved from "http:en.openei.orgwindex.php?titleMacon...

  15. Integrated External Aerodynamic and Underhood Thermal Analysis...

    Energy.gov (indexed) [DOE]

    Cooling Boiling in Head Region - PACCAR Integrated Underhood Thermal and External Aerodynamics- Cummins Cummins SuperTruck Program - Technology and System Level Demonstration of ...

  16. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    ... Self-Sustaining Thorium Boiling Water Reactors Greenspan, Ehud ; Gorman, Phillip M. ... of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; ...

  17. Modeling of coupled heat transfer and reactive transport processesin...

    Office of Scientific and Technical Information (OSTI)

    heating and boiling, and through local convection. In cooler regions, the vapor condenses on fracture walls, where it drains through the fracture network. Slow imbibition of water ...

  18. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    Energy Technology Engineering Center (ETEC), Canoga ... systems (6) tuff (6) water (6) boiling (5) chemical ... and (2) CO2 disposal in a deep saline aquifer. less Full ...

  19. Fluid Inclusion Evidence for Rapid Formation of the Vapor-Dominated...

    Open Energy Information (Open El) [EERE & EIA]

    geohydrology and not just from simple boiling. Authors Masakatsu Sasada and Fraser E. Goff Published Journal Journal of Volcanology and Geothermal Research, 1995 DOI Not Provided...

  20. MEMO TO:

    Office of Legacy Management (LM)

    21 September 2011 To: Madeline Ramos, Puerto Rico Electric Power Authority (PREPA) Copy: Boiling Nuclear Superheat (BONUS) File and Gunseli Shareef, URS (Program Manager) From:...

  1. Applications of the thermogravimetric analysis in the study of fossil fuels

    SciTech Connect (OSTI)

    Huang, He; Wang, Keyu; Wang, Shaojie

    1996-12-31

    Development and applications of thermogravimetric analysis techniques are reported. Applications include: coal structure, coal liquefaction reactions, hydroprocessing of coal-derived resids, and determination of boiling points.

  2. Search for: All records | DOE PAGES

    Office of Scientific and Technical Information (OSTI)

    temperature response during subcooled flow boiling in a rectangular flow channel Yabuki, Tomohide ; Samaroo, Randy ; Nakabeppu, Osamu ; Kawaji, Masahiro October 2015 , Elsevier

  3. Nureth-15 paper

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Advancements on Wall Boiling Modeling in CFD: Leveraging New Understanding from MIT Flow Boiling Facility Massachusetts Institute of Technology L Gilman and E. Baglietto May 12-17, 2013: CASL-U-2013-0099-000 The 15 th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, NURETH-15 NURETH15-633 Pisa, Italy, May 12-17, 2013 ADVANCEMENTS ON WALL BOILING MODELING IN CFD: LEVERAGING NEW UNDERSTANDING FROM MIT FLOW BOILING FACILITY L. Gilman 1 and E. Baglietto 1 1 Massachusetts

  4. Vehicle Technologies Office Merit Review 2014: Thermal Control...

    Energy Savers

    Thermal Control of Power Electronics of Electric Vehicles with Small Channel Coolant Boiling Vehicle Technologies Office Merit Review 2014: Thermal Control of Power Electronics of ...

  5. EA-1394: Final Environmental Assessment

    Office of Energy Efficiency and Renewable Energy (EERE)

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

  6. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    ... Our experiments show that such a nonlinear cantilever resonator, described analytically as ... control over heat transfer in high power density systems utilizing boiling phenomena. ...

  7. Thermal Storage and Advanced Heat Transfer Fluids (Fact Sheet...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    measure the thermophysical properties of heat transfer fluids and storage materials to ... measure the melting point, boiling point, heat capacity, density, viscosity, and phase- ...

  8. Search results | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Download What's Cooking Students in small groups conduct an investigation into the similarities and differences between solar tea and tea brewed by boiling water. Students will...

  9. DOE - Office of Legacy Management -- Elk River Reactor - MN 01

    Office of Legacy Management (LM)

    Designated Name: Not Designated Alternate Name: None Location: Elk River , Minnesota MN.01-1 Evaluation Year: 1985 MN.01-1 Site Operations: Boiling water reactor demonstration, ...

  10. Water injection as a means for reducing non-condensible andcorrosive...

    Office of Scientific and Technical Information (OSTI)

    Two effects have beenrecognized and discussed in the literature as contributing to ... Country of Publication: United States Language: English Subject: 54; ADSORPTION; BOILING; ...

  11. State Nuclear Profiles 2010

    Gasoline and Diesel Fuel Update

    ... Data for 2010 BWR Boiling Water Reactor. Source: Form EIA-860, "Annual Electric ... Data for 2010 PWR Pressurized Light Water Reactor. Source: Form EIA-860, "Annual ...

  12. DOE Solicits Feedback on Subsurface Characterization to Commercialize...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    for Administration of the Wave Energy Converter Prize Geothermal energy below the boiling point is now being harnessed to generate electricity and a host of applications....

  13. Hydrologic and Geochemical Monitoring in Long Valley Caldera...

    Open Energy Information (Open El) [EERE & EIA]

    Differences since 1982 in fluid chemistry of springs has been minor except at Casa Diablo, where rapid fluctuations in chemistry result from near surface boiling and...

  14. Sticky foam

    DOE Patents [OSTI]

    Rand, Peter B.

    1980-01-01

    Access to a space is impeded by the generation of a sticky foam from a tacky polymeric resin and a low boiling solvent.

  15. Microsoft Word - 2014PHYSOR_Powers_FCM_final rev2.docx

    Office of Scientific and Technical Information (OSTI)

    ... could be pursued for boiling water reactor (BWR) cores. ... Analysis work can rely on sound engineering judgment in the ... Fuel for the Deep Burn Management of ...

  16. Imision, Sohmso~ operatlonr Offloe

    Office of Legacy Management (LM)

    3ayslde, Long fElti, 8. Y. Sylvaaia'o prooorr for pa-oduofng unmiunr pwdor by rtoaimtloa ir rtU1 .kr tha amly mpertal rtqpr and ot thllr tlma it la nof poirible to supply pou with...

  17. Effect of Narrow Cut Oil Shale Distillates on HCCI Engine Performance

    SciTech Connect (OSTI)

    Eaton, Scott J; Bunting, Bruce G; Lewis Sr, Samuel Arthur; Fairbridge, Craig

    2009-01-01

    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.

  18. Analysis of plume following ultraviolet laser ablation of doped polymers: Dependence on polymer molecular weight

    SciTech Connect (OSTI)

    Rebollar, Esther; Oujja, Mohamed; Bounos, Giannis; Kolloch, Andreas; Georgiou, Savas; Castillejo, Marta

    2007-02-01

    This work investigates the effect of polymer molecular weight M{sub W} on the plume characteristics of poly(methyl methacrylate) (PMMA) and polystyrene (PS) films doped with iodonaphthalene (NapI) and iodophenanthrene (PhenI) following irradiation in vacuum at 248 nm. Laser-induced fluorescence probing of the plume reveals the presence of ArH products (NapH and PhenH from, respectively, NapI- and PhenI-doped films). While a bimodal translational distribution of these products is observed in all cases, on average, a slower translational distribution is observed in the low M{sub W} system. The extent of the observed dependence is reduced as the optical absorption coefficient of the film increases, i.e., in the sequence NapI/PMMA, PhenI/PMMA, and PS-doped films. Further confirmation of the bimodal translational distributions is provided by monitoring in situ the temporally resolved attenuation by the plume as it expands in vacuum of a continuous wave helium-neon laser propagating parallel to the substrate. Results are discussed in the framework of the bulk photothermal model, according to which ejection requires that a critical number of bonds are broken.

  19. Fast estimation of reboiler reliability

    SciTech Connect (OSTI)

    Durand, A.A.; Bonilla, M.A.O.

    1995-08-01

    The problems one faces in evaluating the reliability of a reboiler design, or in judging the effect of modifications of process conditions on reboiler operation can be complex. To carry out such evaluations, it is necessary for engineers to perform some calculations to determine: heat transfer coefficients in convection boiling; temperature difference, for the onset of nucleate boiling; heat transfer coefficients in the nucleate boiling region; critical heat flux or critical temperature difference; minimum {Delta}T for film boiling; and heat transfer coefficients for the film boiling region. There are a number of correlations, graphs, and computer programs that can be used to make these calculations. However, besides being laborious, it is still difficult to get a suitable picture of the overall problem from just this data. To simplify the process, and to have a better understanding of the problem, a map of the different boiling regions and their boundaries is presented here. With this map it is possible to locate the design or operating point of a specific kettle reboiler among all the boiling regions, enabling one to make a clearer analysis of its behavior. The parameters used to develop this map are described.

  20. Coal liquefaction process

    DOE Patents [OSTI]

    Wright, C.H.

    1986-02-11

    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.

  1. Quenching phenomena in natural circulation loop

    SciTech Connect (OSTI)

    Umekawa, Hisashi; Ozawa, Mamoru; Ishida, Naoki

    1995-09-01

    Quenching phenomena has been investigated experimentally using circulation loop of liquid nitrogen. During the quenching under natural circulation, the heat transfer mode changes from film boiling to nucleate boiling, and at the same time flux changes with time depending on the vapor generation rate and related two-phase flow characteristics. Moreover, density wave oscillations occur under a certain operating condition, which is closely related to the dynamic behavior of the cooling curve. The experimental results indicates that the occurrence of the density wave oscillation induces the deterioration of effective cooling of the heat surface in the film and the transition boiling regions, which results in the decrease in the quenching velocity.

  2. Coal liquefaction process

    DOE Patents [OSTI]

    Wright, Charles H.

    1986-01-01

    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.

  3. Software Based DIAS and FACET Training Guide

    Energy Science and Technology Software Center (OSTI)

    2005-03-01

    The Dynamic Information Architecture System (DIAS) is an ANL- developed framework fo developing multidisciplinary simulation systems. It is copyrighted (ANL-SF-96-130 and ANL-SF-98-127) and licensed software and the invention is patented (ANL-IN-95-146). This Developer's Guide and accompanying Farm Tax tutorial provides software based system and documentation tha supports the DIAS licenses.

  4. Interim Safety Basis for Fuel Supply Shutdown Facility

    SciTech Connect (OSTI)

    BENECKE, M.W.

    2000-09-07

    This ISB, in conjunction with the IOSR, provides the required basis for interim operation or restrictions on interim operations and administrative controls for the facility until a SAR is prepared in accordance with the new requirements or the facility is shut down. It is concluded that the risks associated with tha current and anticipated mode of the facility, uranium disposition, clean up, and transition activities required for permanent closure, are within risk guidelines.

  5. Heat Transfer Fluids for Solar Water Heating Systems | Department...

    Energy.gov (indexed) [DOE]

    a high boiling point. Viscosity and thermal capacity determine the amount of pumping energy required. A fluid with low viscosity and high specific heat is easier to pump, because...

  6. Search results | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    tea brewed by boiling water. Students will compare their two samples on four criteria-color, clarity, smell and taste-rate which they prefer, and graph the results of the...

  7. L3:VUQ.SAUQ.P2-2.01 Brian Williams LANL

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    distributions for DBCoeff based on two calibrations to crud index from F71 and F22 (green) and F71, F22, and F88 (blue). Figure 4 . Calibrated V IPRE---W boiling index...

  8. BWR Assembly Optimization for Minor Actinide Recycling

    SciTech Connect (OSTI)

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

    2010-03-22

    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).

  9. CX-004148: Categorical Exclusion Determination

    Energy.gov [DOE]

    Online Monitoring Implementation in Boiling Water ReactorsCX(s) Applied: B3.6, B5.1Date: 09/17/2010Location(s): Knoxville, TennesseeOffice(s): Energy Efficiency and Renewable Energy

  10. CASL-U-2015-0054-000 COBRA-TF Subchannel Thermal-Hydraulics...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Coolant-Boiling in Rod Arrays-Two Fluids (COBRA-TF) is a Thermal Hydraulic (TH) ... sponsorship of the NRC 68, began as a TH rod-bundle analysis code, but has been ...

  11. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    Nextra Energy Duane Arnold LLC 1 Plant 1 Reactor 601 4,451 100.0 Note: Totals may not ... 84.5 BWR 211975 2212014 601 4,451 84.5 Data for 2010 BWR Boiling Water Reactor. ...

  12. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    96.7 BWR 6301971 982030 554 4,695 96.7 Data for 2010 BWR Boiling Water Reactor. ... 520 full-time and contract employees. Reactor Descriptions: The nuclear generating unit ...

  13. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    85.5 BWR 1211969 492029 615 4,601 85.5 Data for 2010 BWR Boiling Water Reactor. ... not including security personnel. Reactor Descriptions: The nuclear generating unit ...

  14. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    Entergy Nuclear Generation Co 1 Plant 1 Reactor 685 5,918 100.0 Note: Totals may not ... 98.7 BWR 1211972 682012 685 5,918 98.7 Data for 2010 BWR Boiling Water Reactor. ...

  15. State Nuclear Profiles 2010

    U.S. Energy Information Administration (EIA) (indexed site)

    Entergy Nuclear Generation Co 1 Plant 1 Reactor 685 5,918 100.0 Owner Note: Totals may ... Data for 2010 BWR Boiling Water Reactor. Source: Form EIA-860, "Annual Electric ...

  16. Minnesota Nuclear Profile - Monticello

    U.S. Energy Information Administration (EIA) (indexed site)

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

  17. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    Entergy Nuclear Vermont Yankee 1 Plant 1 Reactor 620 4,782 100.0 Note: Totals may not ... 88.0 BWR 11301972 3212012 620 4,782 88.0 Data for 2010 BWR Boiling Water Reactor. ...

  18. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    99.3 BWR 1131975 12272034 1,858 14,808 91.0 Data for 2010 BWR Boiling Water Reactor. ... Construction Cost: 2.490 billion (2007 USD)2 Reactor Descriptions: Both units are General ...

  19. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    81.2 BWR 311977 722036 3,309 24,771 85.4 Data for 2010 BWR Boiling Water Reactor. ... Construction Cost: 3.259 billion (2007 USD)2 Reactor Descriptions: All units are General ...

  20. State Nuclear Profiles 2010

    U.S. Energy Information Administration (EIA) (indexed site)

    Share of State nuclear net generation (percent) Cooper 1 767 6,793 101.1 BWR 711974 1182014 767 6,793 101.1 Data for 2010 BWR Boiling Water Reactor. Source: Form EIA-860, ...

  1. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    95.6 BWR 951979 6132038 1,759 13,902 90.2 Data for 2010 BWR Boiling Water Reactor. ... Construction Cost: 3.214 billion (2007 USD)2 Reactor Descriptions: The Hatch plant has ...

  2. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    98.0 BWR 6161986 8292025 974 8,363 98.0 Data for 2010 BWR Boiling Water Reactor. ... Staffing: The plant has approximately 560 full-time employees. Reactor Descriptions: The ...

  3. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    84.9 BWR 7281975 10172034 855 6,361 84.9 Data for 2010 BWR Boiling Water Reactor. ... Staffing: There are about 650 employees at this facility. Reactor Descriptions: ...

  4. Mississippi Nuclear Profile - Grand Gulf

    U.S. Energy Information Administration (EIA) (indexed site)

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

  5. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    101.1 BWR 711974 1182014 767 6,793 101.1 Data for 2010 BWR Boiling Water Reactor. ... Construction Cost: 1.152 billion (2007 USD)2 Reactor Descriptions: The Cooper unit is a ...

  6. Mamba_ptacFeb2014.ppt

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Cladding surface z (cm) 100.0 r (microns) 50.0 Coolant flow Thermodynamics and chemical kinetics computed at each node and time step MAMBA example grid: Heat flux Boiling...

  7. EERE Success Story-Geothermal Technology Breakthrough in Alaska...

    Energy.gov (indexed) [DOE]

    A binary process mixes geothermal brine with a working fluid that has a lower boiling point than water. This fluid is compressed into steam to turn a turbine and generate ...

  8. -.. IJNDSAY CHFXtCAL COMFAKY IND'J?ZRIAL HYGIEFE SJRVEY PART1

    Office of Legacy Management (LM)

    ... shoveling Fthe cake into barrels or into tanks and.cleanup of areaffter press :dumping ... 0.5 Between Red Mud Boiling & Dissolving Tanks Floor Level Ir . . . 3' Level 1 BBIS. ...

  9. Exploratory Well At Raft River Geothermal Area (1950) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    and Crank wells, encountered boiling water. References Diek, A.; White, L.; Roegiers, J.-C.; Moore, J.; McLennan, J. D. (1 January 2012) BOREHOLE PRECONDITIONING OF GEOTHERMAL...

  10. Search results | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    boiling water. Students will compare their two samples on four criteria-color, clarity, smell and taste-rate which they prefer, and graph the results of the experiment as a class....

  11. Search results | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    boiling water. Students will compare their two samples on four criteria-color, clarity, smell and taste-rate which they prefer, and graph the results of the experiment as a...

  12. Microsoft Word - Final WIPP Rad Release Phase 1 04 22 2014 r2...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    ... the potential boiling liquid expanding vapor explosions ... that damage to the outer metal drum had occurred but the ... all the major components, i.e., bearings, shaft, rotor, etc. ...

  13. PROGRESS IN RESEARCH

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    ... and Negative Heat Capacity ... normalization coefficient (ANC) for the 7 Be+p 8 B using the transfer reaction 14 N( 7 ... the boiling and condensation and noted the ...

  14. Aluto-Langano Geothermal Field, Ethiopian Rift Valley- Physical...

    Open Energy Information (Open El) [EERE & EIA]

    region of the Ethiopian Rift Valley. The upflow zone for the system lies along a deep, young NNE trending fault and is characterized by boiling. As a result, the deep upflow zone...

  15. Microsoft Word - Pu Disposition Red Team Report.docx

    National Nuclear Security Administration (NNSA)

    ... baseline c hange p roposal BWR boiling---water r eactor CCO Criticality C ontrol O ... D isposition A greement PWG Plutonium D isposition W orking G roup PWR pressurized---water ...

  16. Volatilization of iodine from nitric acid using peroxide

    DOE Patents [OSTI]

    Cathers, G.I.; Shipman, C.J.

    1975-10-21

    A method for removing radioactive iodine from nitric acid solution by adding hydrogen peroxide to the solution while concurrently holding the solution at the boiling point and distilling hydrogen iodide from the solution is reported.

  17. Savannah River Ecology Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    At present, Boiling Springs is the only known intact, old-growth community that exists on the SRS. Ironically, this remnant stand is both the oldest plant community recorded for ...

  18. Consortium for Advanced Simulation ...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Departure from nucleate boiling (DNB) serves as a critical pa- rameter in nuclear power plant operational and safety analysis. It occurs when a fuel rod clad surface is overheated ...

  19. Microsoft Word - 504_FinalPaper.doc

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    ... boiling and two-phase flow models in the Eulerian Multiphase CFD code STAR-CD 7. ... a sequence of three mesh resolutions. 2. CFD Two Phase Models and Constitutive Relations ...

  20. A Review of Stress Corrosion Cracking/Fatigue Modeling for Light Water Reactor Cooling System Components

    Energy.gov [DOE]

    In the United States currently there are approximately 104 operating light water reactors. Of these, 69 are pressurized water reactors (PWRs) and 35 are boiling water reactors (BWRs). In 2007, the...

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

    DOE Patents [OSTI]

    Nizamoff, Alan J.

    1980-01-01

    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.

  2. Property:Building/SPPurchasedEngyNrmlYrMwhYrOil-FiredBoiler ...

    Open Energy Information (Open El) [EERE & EIA]

    rmlYrMwhYrOil-FiredBoiler Jump to: navigation, search This is a property of type String. Oil-fired boiler Pages using the property "BuildingSPPurchasedEngyNrmlYrMwhYrOil-FiredBoil...

  3. Geothermal Energy (5 Activities) | Department of Energy

    Energy.gov (indexed) [DOE]

    rock to water? How does energy transferred between fluids in a binary geothermal power plant work? How does salinity affect the boiling point of water? How do the emissions...

  4. Lesson 6 - Atoms to Electricity | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    6 - Atoms to Electricity Lesson 6 - Atoms to Electricity 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 Heat Pressure Water Fission Control Fuel assemblies Control rods Coolant Pressure vessel Electricity Generation Generator Condenser Cooling tower Lesson 6 - Atoms to Electricity.pptx (9.7 MB) More

  5. PowerPoint Presentation

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    9-000 CASL Program Highlights June 2015 Jess C. Gehin Oak Ridge National Laboratory June 30, 2015 1 First CASL Simulation of Departure from Nucleate Boiling Challenge Problem with VERA Completed * PWR Streamline Break (SLB) Event Simulated - Broken main steam pipe in one loop at hot zero power (HZP) - Core return to power with high peaking factor could lead to departure from nucleate boiling (DNB) on fuel rods * CASL Modeling Approach - Core boundary condition from system transient code - Core

  6. Process for the conversion of alcohols and oxygenates to hydrocarbons in a turbulent fluid bed reactor

    SciTech Connect (OSTI)

    Avidan, A. A.; Kam, A. Y.

    1985-04-23

    Improvements in converting C/sub 1/-C/sub 3/ monohydric alcohols, particularly methanol, related oxygenates of said alcohols and/or oxygenates produced by Fischer-Tropsch synthesis to light olefins, gasoline boiling range hydrocarbons and/or distillate boiling range hydrocarbons are obtained in a fluidized bed of ZSM-5 type zeolite catalyst operating under conditions effective to provide fluidization in the turbulent regime.

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

    DOE Patents [OSTI]

    Pinnau, Ingo; Lokhandwala, Kaaeid; Nguyen, Phuong; Segelke, Scott

    1997-11-18

    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.

  8. Collecting and Characterizing Validation Data

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Characterizing Validation Data to Support Advanced Simulation of Nuclear Reactor Hydraulics Nam Dinh North Carolina State University Anh Bui Idaho National Laboratory Hyung Lee Bettis Laboratory ASME 2013 Verification and Validation Symposium Las Vegas, NV, May 22-24, 2013 Multi-Physics, Multi-Scale Problem Validation Hierarchy (Validation Pyramid) of Subcooled Boiling Flow Model Bayesian Framework for Data Integration Nuclear System Analysis - Subcooled Boiling Flow Example * Underlying physics

  9. SUBCOOLING DETECTOR

    DOE Patents [OSTI]

    McCann, J.A.

    1963-12-17

    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)

  10. Coal Liquefaction desulfurization process

    DOE Patents [OSTI]

    Givens, Edwin N. (Bethlehem, PA)

    1983-01-01

    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.

  11. Prediction of critical heat flux in water-cooled plasma facing components using computational fluid dynamics.

    SciTech Connect (OSTI)

    Bullock, James H.; Youchison, Dennis Lee; Ulrickson, Michael Andrew

    2010-11-01

    Several commercial computational fluid dynamics (CFD) codes now have the capability to analyze Eulerian two-phase flow using the Rohsenow nucleate boiling model. Analysis of boiling due to one-sided heating in plasma facing components (pfcs) is now receiving attention during the design of water-cooled first wall panels for ITER that may encounter heat fluxes as high as 5 MW/m2. Empirical thermalhydraulic design correlations developed for long fission reactor channels are not reliable when applied to pfcs because fully developed flow conditions seldom exist. Star-CCM+ is one of the commercial CFD codes that can model two-phase flows. Like others, it implements the RPI model for nucleate boiling, but it also seamlessly transitions to a volume-of-fluid model for film boiling. By benchmarking the results of our 3d models against recent experiments on critical heat flux for both smooth rectangular channels and hypervapotrons, we determined the six unique input parameters that accurately characterize the boiling physics for ITER flow conditions under a wide range of absorbed heat flux. We can now exploit this capability to predict the onset of critical heat flux in these components. In addition, the results clearly illustrate the production and transport of vapor and its effect on heat transfer in pfcs from nucleate boiling through transition to film boiling. This article describes the boiling physics implemented in CCM+ and compares the computational results to the benchmark experiments carried out independently in the United States and Russia. Temperature distributions agreed to within 10 C for a wide range of heat fluxes from 3 MW/m2 to 10 MW/m2 and flow velocities from 1 m/s to 10 m/s in these devices. Although the analysis is incapable of capturing the stochastic nature of critical heat flux (i.e., time and location may depend on a local materials defect or turbulence phenomenon), it is highly reliable in determining the heat flux where boiling instabilities begin

  12. CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling

    SciTech Connect (OSTI)

    Fan-Bill Cheung; Joy L. Rempe

    2004-06-01

    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.

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

    DOE Patents [OSTI]

    Turner, Terry D.; Beller, Laurence S.; Clark, Michael L.; Klingler, Kerry M.

    1997-01-01

    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.

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

    DOE Patents [OSTI]

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

    1997-10-14

    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.

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

    DOE Patents [OSTI]

    MacArthur, James B.; Comolli, Alfred G.; McLean, Joseph B.

    1989-01-01

    A process 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.degree.-750.degree. 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.degree. 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.degree.-800.degree. F. temperature, 1000-4000 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.degree.-860.degree. F. temperature for further hydrogenation and hydroconversion reactions. A 600.degree.-750.degree. 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.

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

    DOE Patents [OSTI]

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

    1989-10-17

    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.

  17. Oceanographic effects of the 1992 Point Loma sewage pipe spill

    SciTech Connect (OSTI)

    Casey, R.; Ciccateri, A.; Dougherty, K.; Gacek, L.; Lane, S.; Liponi, K.; Leeds, R.; Walsh, F. )

    1992-01-01

    Early in early 1992, 180 million gallons of advanced primarily treated sewage emptied into 10 meters of water from the broken Point Loma sewage pipe, San Diego. For about two months a sewage boil about the size of a football field existed at the surface and within the Point Loma kelp bed. Sampling and observations taken during the spill indicated the surface waters at the spill site were grayish and smelling of sewage. The sewage water had mixed with the marine waters reducing salinity to about one-half normal (or 15 ppt.). The sediment load of the sewage coated the blades of the giant kelp and the kelp was limp and withdrawn from the surface. At the site of the main boil the kelp appeared to have dropped to the bottom. Sediments on the bottom in the boil area were mainly coarse sands as compared to the surrounding sandy-muds. Preliminary results using laboratory analysis suggest: one month into the spill no infauna were observed in the sediments or planktons in the water of the boil area, but were in the surrounding sediments and water; the observed phytoplankton were dominated by dinoflagellates and suggested red tide conditions surrounding the boil. The site has been monitored monthly since the spill to observe further impact and recovery.

  18. Transetherification method

    DOE Patents [OSTI]

    Hearn, D.

    1985-04-09

    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.

  19. Transetherification method

    DOE Patents [OSTI]

    Hearn, Dennis

    1985-01-01

    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.

  20. Method for conducting exothermic reactions

    DOE Patents [OSTI]

    Smith, Jr., Lawrence (Bellaire, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

    1993-01-01

    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.degree. to 300.degree. 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.

  1. Oligomerization process

    DOE Patents [OSTI]

    Smith, Jr., Lawrence A. (Bellaire, TX); hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

    1991-01-01

    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.degree. 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. Etherification process

    DOE Patents [OSTI]

    Smith, Jr., Lawrence A. (Houston, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

    1990-01-01

    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.degree. to 300.degree. 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.

  3. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, Jr., Lawrence A. (Bellaire, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

    1993-01-01

    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.degree. to 300.degree. 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.

  4. Methods of cracking a crude product to produce additional crude products

    DOE Patents [OSTI]

    Mo, Weijian; Roes, Augustinus Wilhelmus Maria; Nair, Vijay

    2009-09-08

    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.

  5. Reactor for exothermic reactions

    DOE Patents [OSTI]

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

    1993-03-02

    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.

  6. Generic Issue 87: Flexible wedge gate valve test program

    SciTech Connect (OSTI)

    Steele, R. Jr.; DeWall, K.G.; Watkins, J.C. )

    1991-01-01

    Qualification and flow isolation tests were conducted to analyze the ability of selected boiling water reactor process valves to perform their containment isolation functions at high energy pipe break conditions and other more normal flow conditions. Numerous parameters were measured to assess valve and motor-operator performance at various valve loadings and to assess industry practices for predicting valve and motor operator requirements. The valves tested were representative of those used in reactor water cleanup systems in boiling water reactors and those used in boiling water reactor high-pressure coolant injection (HPCI) steam lines. These tests will provide further information for the US Nuclear Regulatory Commission Generic Issue-87, Failure of the HPCI Steam Line Without Isolation,'' and Generic Letter 89--10, Safety-related Motor Operated Valve Testing and Surveillance.'' 6 refs., 54 figs., 4 tabs.

  7. Method for conducting exothermic reactions

    DOE Patents [OSTI]

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

    1993-01-05

    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.

  8. Etherification process

    DOE Patents [OSTI]

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

    1990-08-21

    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 figs.

  9. Oligomerization process

    DOE Patents [OSTI]

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

    1991-03-26

    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.

  10. Summary - Small Column Ion Exchange (SCIX)Technology at the SRS

    Office of Environmental Management (EM)

    ETR R Un Baseline The Sm being The SC operat which Sr, and waste critical the SC deploy Specif exchan [CST]) CST, a (mono and so (RMF) maturi readin design moving The pu techni projec Site: S roject: S E Report Date: F ited States Sma Why DOE e SCIX System Pr mall Column Io developed at S CIX system is tions (ion excha function to rem d actinides) fro and prepare th l technology ele CIX system tha yment and thes fically the critica nge on a selec ) housed in an actinide and Sr osodium titanat

  11. L I II C

    Office of Legacy Management (LM)

    -- - L I II C rr u c c c 7 i' :- ' r' ' 7 i ' -- A' t i ()lL.H~ ORAU 89/i-29 Prepared by Oak Ridge Associated Universities Prepared for Division of Facility and Site Decommissioning Projects U.S. Department of Energy VERIFICATION OF REMEDIAL ACTIONS ALBANYRESEARCHCENTER ALBANY, OREGON P. R. C O lTEN Environmental Survey and Site Assessment Program Energy/Environment Systems Division FINAL REPORT OCTOBER 1989 NOTICES Tha opiniona l xprSaaJd harJln do not n acoaa~rlly ranKI thy oplnioru of thJ l

  12. Further Notice of 230kV Circuit Planned Outages | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Further Notice of 230kV Circuit Planned Outages Further Notice of 230kV Circuit Planned Outages Docket No. EO-05-01. Order No. 202-05-03: Pursuant 10 the United States Department of Energy "DOE") Order No. 102-05-3, issued December 20, 2005 ("DOE Potomac River Order''), Pepco hereby files this Further Notice Of 230kV Circuit Planned Outages serving the Potomac River Substation, and through thaI station, the District of Columbia. Further Notice of 230kV Circuit Planned Outages

  13. 3210T3 Risk Code Assignment;

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    210 Appendix T3 Risk Code Assignment 1.0 Purpose Identification of work hazards and understanding their risks is an essential part of Jefferson Lab's work process as defined in ES&H Manual Chapter 3210 Appendix T1 Work Planning, Control, and Authorization Procedure. The purpose of this document is to demonstrate the accepted method for determining the Risk Code for task steps identified while completing Task Hazard Analysis (THA) Worksheet. 2.0 Scope This appendix describes the procedure for

  14. Screening Prosopis (mesquite) species for biofuel production on semi-arid lands. Final report, April 1, 1978-March 30, 1981

    SciTech Connect (OSTI)

    Felker, P; Cannell, G H; Clark, P R; Osborn, J F; Nash, P

    1985-01-01

    Arid adapted nitrogen fixing trees and shrubs of the genus Prosopis (mesquite) have been examined for woody biomass production on semi-arid lands of southwestern United States. A germ-plasm collection of 900 accessions from North and South America and Africa was assembled. Field studies screening for biomass production, frost tolerance, response to irrigation, pod production and heat/drought tolerance involved a total of 80 accessions. Selections made from survivors of coal/frost screening trial had more frost tolerance and biomass productivity than prostrate selections from the ranges of Arizona, New Mexico and west Texas. Thirteen Prosopis species were found to nodulate, reduce acetylene to ethylene, and grow on a nitrogen free media in greenhouse experiments. The salinity tolerance of six Prosopis species was examined on a nitrogen free media in greenhouse experiments. No reduction in growth occurred for any species tested at a salinity of 6000 mg NaC1/L which is considered too saline for normal agricultural crops. Individual trees have grown 5 to 7 cm in basal diameter, and 2.0 to 3.7 meters in height per year and have achieved 50 kg oven dry weight per tree in 2 years with 600 mm water application per year. Vegetative propagation techniques have been developed and clones of these highly productive trees have been made. Small pilots on 1.5 x 1.5 m spacing in the California Imperial Valley had a first and second season dry matter production of 11.7 and 16.9 T/ha for P. chilensis (0009), 7.1 and 6.9 T/ha for P. glandulosa var. torreyana (0001), 9.8 and 19.2 T/ha for P. alba (0039) and 7.9 and 14.5 T/ha for progency of a California ornamental (0163). The projected harvested costs of $25.00 per oven dry ton or $1.50 per million Btu's compare favorable with coal and other alternative fuel sources in South Texas.

  15. Characteristic evaluation of cooling technique using liquid nitrogen and metal porous media

    SciTech Connect (OSTI)

    Tanno, Yusuke; Ito, Satoshi; Hashizume, Hidetoshi

    2014-01-29

    A remountable high-temperature superconducting magnet, whose segments can be mounted and demounted repeatedly, has been proposed for construction and maintenance of superconducting magnet and inner reactor components of a fusion reactor. One of the issues in this design is that the performance of the magnet deteriorates by a local temperature rise due to Joule heating in jointing regions. In order to prevent local temperature rise, a cooling system using a cryogenic coolant and metal porous media was proposed and experimental studies have been carried out using liquid nitrogen. In this study, flow and heat transfer characteristics of cooling system using subcooled liquid nitrogen and bronze particle sintered porous media are evaluated through experiments in which the inlet degree of subcooling and flow rate of the liquid nitrogen. The flow characteristics without heat input were coincided with Erguns equation expressing single-phase flow in porous materials. The obtained boiling curve was categorized into three conditions; convection region, nucleate boiling region and mixed region with nucleate and film boiling. Wall superheat did not increase drastically with porous media after departure from nucleate boiling point, which is different from a situation of usual boiling curve in a smooth tube. The fact is important characteristic to cooling superconducting magnet to avoid its quench. Heat transfer coefficient with bronze particle sintered porous media was at least twice larger than that without the porous media. It was also indicated qualitatively that departure from nucleate boiling point and heat transfer coefficient depends on degree of subcooling and mass flow rate. The quantitative evaluation of them and further discussion for the cooling system will be performed as future tasks.

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

    SciTech Connect (OSTI)

    Buongiorno, Jacopo; Hu, Lin-wen

    2009-07-31

    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

  17. Process and catalyst for converting synthesis gas to liquid hydrocarbon mixture

    DOE Patents [OSTI]

    Rao, V. Udaya S.; Gormley, Robert J.

    1987-01-01

    Synthesis gas containing CO and H.sub.2 is converted to a high-octane hydrocarbon liquid in the gasoline boiling point range by bringing the gas into contact with a heterogeneous catalyst including, in physical mixture, a zeolite molecular sieve, cobalt at 6-20% by weight, and thoria at 0.5-3.9% by weight. The contacting occurs at a temperature of 250.degree.-300.degree. C., and a pressure of 10-30 atmospheres. The conditions can be selected to form a major portion of the hydrocarbon product in the gasoline boiling range with a research octane of more than 80 and less than 10% by weight aromatics.

  18. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    SciTech Connect (OSTI)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made.

  19. Temperature dependent droplet impact dynamics on flat and textured surfaces

    SciTech Connect (OSTI)

    Azar Alizadeh; Vaibhav Bahadur; Sheng Zhong; Wen Shang; Ri Li; James Ruud; Masako Yamada; Liehi Ge; Ali Dhinojwala; Manohar S Sohal

    2012-03-01

    Droplet impact dynamics determines the performance of surfaces used in many applications such as anti-icing, condensation, boiling and heat transfer. We study impact dynamics of water droplets on surfaces with chemistry/texture ranging from hydrophilic to superhydrophobic and across a temperature range spanning below freezing to near boiling conditions. Droplet retraction shows very strong temperature dependence especially for hydrophilic surfaces; it is seen that lower substrate temperatures lead to lesser retraction. Physics-based analyses show that the increased viscosity associated with lower temperatures can explain the decreased retraction. The present findings serve to guide further studies of dynamic fluid-structure interaction at various temperatures.

  20. MHTGR steam generator on-line heat balance, instrumentation and function

    SciTech Connect (OSTI)

    Klapka, R.E.; Howard, W.W.; Etzel, K.T. ); Basol, M.; Karim, N.U. )

    1991-09-01

    Instrumentation is used to measure the Modular High Temperature Gas-Cooled Reactor (MHTGR) steam generator dissimilar metal weld temperature during start-up testing. Additional instrumentation is used to determine an on-line heat balance which is maintained during the 40 year module life. In the process of calibrating the on-line heat balance, the helium flow is adjusted to yield the optimum boiling level in the steam generator relative to the dissimilar metal weld. After calibration is complete the weld temperature measurement is non longer required. The reduced boiling level range results in less restrictive steam generator design constraints.

  1. Synthesis and Optical Properties of Sulfide Nanoparticles Prepared in Dimethylsulfoxide

    SciTech Connect (OSTI)

    Li, Yuebin; Ma, Lun; Zhang, Xing; Joly, Alan G.; Liu, Zuli; Chen, Wei

    2008-11-01

    Many methods have been reported for the formation of sulfide nanoparticles by the reaction of metallic salts with sulfide chemical sources in aqueous solutions or organic solvents. Here, we report the formation of sulfide nanoparticles in dimethylsulfoxide (DMSO) by boiling metallic salts without sulfide sources. The sulfide sources are generated from the boiling of DMSO and react with metallic salts to form sulfide nanoparticles. In this method DMSO functions as a solvent and a sulfide source as well as a stabilizer for the formation of the nanoparticles. The recipe is simple and economical making sulfide nanoparticles formed in this way readily available for many potential applications.

  2. Superconducting magnet cooling system

    DOE Patents [OSTI]

    Vander Arend, Peter C.; Fowler, William B.

    1977-01-01

    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.

  3. Small break LOCA analysis of the ONRL high flux isotope reactor

    SciTech Connect (OSTI)

    Wilson, T.L. Jr.; Cook, D.H.; Sozer, A.

    1988-01-01

    A digital simulation program, HFIRSYS, was developed using MMS to analyze small break loss of coolant events in the ORNL High Flux Isotope Reactor. The code evaluates the response of the primary reactor system including automatic controls actions resulting from breaks in auxiliary piping connected to the primary. The primary output of the code is the margin to the onset of nucleate boiling expressed as a ratio of heat flux which would cause boiling to the current hot channel heat flux. A description of the model, validation results and a sample transient are presented.

  4. Upgrading heavy oils by non-catalytic treatment with hydrogen and hydrogen transfer solvent

    SciTech Connect (OSTI)

    Derbyshire, F.J.; Mitchell, T.O.; Whitehurst, D.D.

    1981-09-29

    Heavy liquid hydrocarbon oil, such as petroleum derived tars, predominantly boiling over 425/sup 0/C, are upgraded to products boiling below 425/sup 0/C, without substantial formation of insoluble char, by heating the heavy oil with hydrogen and a hydrogen transfer solvent in the absence of hydrogenation catalyst at temperatures of about 320/sup 0/C to 500/sup 0/C, and a pressure of 20 to 180 bar for 3 to 30 minutes. The hydrogen transfer solvents polycyclic compounds free of carbonyl groups, e.g., pyrene, and have a polarographic reduction potential which is less negative than phenanthrene and equal to or more negative than azapyrene.

  5. Process for upgrading heavy hydrocarbonaceous oils

    SciTech Connect (OSTI)

    Fisher, I.P.; Souhrada, F.; Woods, H.J.

    1981-10-13

    An integrated upgrading process is disclosed which can be used to lower the specific gravity, viscosity and boiling range of heavy, viscous hydrocarbonaceous oil . The process consists of fractionally distilling the oil, treating its residuum with a hydrogen donor material under hydrocracking conditions, fractionally distilling the effluent from the hydrocracking zone and rehydrogenating that portion boiling from about 180/sup 0/ C to 350/sup 0/ C for recycling to the hydrocracking zone. The liquid portion of the oil not recycled can be recombined into a reconstituted crude suitable for transporting by normal crude pipelines.

  6. Four different shale oils processed into jet fuel

    SciTech Connect (OSTI)

    Not Available

    1987-03-01

    Crude shale oils produced by (a) Geokinetics, (b) Occidental, (c) Paraho, and (d) Tosco II processes have each been catalytically hydroprocessed to produce jet fuel fractions. The shale oil hydroprocessing was performed at low, medium and high hydroprocessing severities. Hydroprocessing severity was changed mainly by varying the temperature. Full boiling range (121-300/sup 0/C) jet fuel was produced from the hydroprocessed product of the raw oil distillates boiling below 343/sup 0/C. This paper describes the shale oil properties and hydroprocessing, gives the results of sulfur removal and hydrogenated shale oil distillation, and lists the physical and chemical properties of the jet fuels. 2 figures, 3 tables.

  7. Straight Vegetable Oil as a Vehicle Fuel? (Fact Sheet), Energy Efficiency & Renewable Energy (EERE), Vehicle Technologies Office (VTO)

    Alternative Fuels and Advanced Vehicles Data Center

    Performance of SVO Research has shown that there are sev- eral technical barriers to widespread use of SVO as a vehicle fuel. The published engineering literature strongly indicates that the use of SVO leads to reduced engine life, 1 caused by the buildup of carbon deposits inside the engine and the buildup of SVO in the engine lubricant. These issues are attributable to SVO's high viscosity and high boiling point relative to the required boiling range for diesel fuel. The carbon buildup doesn't

  8. CASL-U-2015-0016-000

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    6-000 Advanced Calibration and Validation of a Mechanistic Model of Subcooled Boiling Two-Phase Flow Anh Bui, Idaho National Laboratory Brian Williams, Los Alamos National Laboratory Nam Dinh, North Carolina State University April 6, 2014 CASL-U-2015-0016-000 Proceedings of ICAPP 2014 Charlotte, USA, April 6-9, 2014 Paper 14257 Advanced Calibration and Validation of a Mechanistic Model of Subcooled Boiling Two-Phase Flow Anh Bui 1 , Brian Williams 2 , Nam Dinh 3,* 1 Idaho National Laboratory

  9. CNS 2008 Template

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Sensitivity Study Of Eulerian Multiphase Boiling Models: NPHASE-CMFD University of Michigan I. Asher, K. Fidkowski, T. Drzewiecki, T. Grunloh, V. Petrov, A. Manera, T Downar May 12-17, 2013: CASL-U-2013-0095-000 The 15 th International Topical Meeting on Nuclear Reactor Thermal - Hydraulics, NURETH-15 NURETH15-215 Pisa, Italy, May 12-17, 2013 SENSITIVITY STUDY OF EULERIAN MULTIPHASE BOILING MODELS: NPHASE-CMFD I. Asher 1 , K. Fidkowski 1 , T. Drzewiecki 2 , T. Grunloh 2 , V. Petrov 2 , A. Manera

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

    SciTech Connect (OSTI)

    Nibur, Kevin A.

    2010-11-01

    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.

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

    SciTech Connect (OSTI)

    Dadfarnia, Mohsen; Nibur, Kevin A.; San Marchi, Christopher W.; Sofronis, Petros; Somerday, Brian P.; Foulk, James W., III; Hayden, Gary A.

    2010-07-01

    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.

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

    SciTech Connect (OSTI)

    1997-06-01

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

  13. Anhydrous hydrogen fluoride electrolyte battery. [Patent application

    DOE Patents [OSTI]

    Not Available

    1972-06-26

    It is an object of the invention to provide a primary cell or battery using ammonium fluoride--anhydrous hydrogen fluoride electrolyte having improved current and power production capabilities at low temperatures. It is operable at temperatures substantially above the boiling point of hydrogen fluoride. (GRA)

  14. Cyclic thermochemical process for producing hydrogen using cerium-titanium compounds

    DOE Patents [OSTI]

    Bamberger, C.E.

    A thermochemical cyclic process for producing hydrogen employs the reaction between ceric oxide and titanium dioxide to form cerium titanate and oxygen. The titanate is treated with an alkali metal hydroxide to give hydrogen, ceric oxide, an alkali metal titanate and water. Alkali metal titanate and water are boiled to give titanium dioxide which, along with ceric oxide, is recycled.

  15. Cyclic thermochemical process for producing hydrogen using cerium-titanium compounds

    DOE Patents [OSTI]

    Bamberger, Carlos E.

    1980-01-01

    A thermochemical cyclic process for producing hydrogen employs the reaction between ceric oxide and titanium dioxide to form cerium titanate and oxygen. The titanate is treated with an alkali metal hydroxide to give hydrogen, ceric oxide, an alkali metal titanate and water. Alkali metal titanate and water are boiled to give titanium dioxide which, along with ceric oxide, is recycled.

  16. Conditioning of BWR Control - Elements Using the New MOSAIK 80T/SWR-SE Cask - Concept

    SciTech Connect (OSTI)

    Oldiges, O.; Blenski, H.-J.; Engelage, H.; Behrens, W.; Majunke, J.; Schwarz, W.; Hallfarth, Dr.

    2002-02-27

    During the operation of Boiling Water Reactors, Control - Elements are used to control the neutron flux inside the reactor vessel. After the end of the lifetime, the Control - Elements are usually stored in the fuel - elements - pool of the reactor. Up to now, in Germany no conditioning of Control - Elements has been done in a BWR under operation.

  17. What's Cooking

    Education Teach & Learn

    Students in small groups conduct an investigation into the similarities and differences between solar tea and tea brewed by boiling water. Students will compare their two samples on four criteria—color, clarity, smell and taste—rate which they prefer, and graph the results of the experiment as a class.

  18. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    9,241 100.0 Energy Northwest 1 Plant 1 Reactor 1,097 9,241 100.0 Note: Totals may not ... 96.2 BWR 12131984 12202033 1,097 9,241 96.2 Data for 2010 BWR Boiling Water Reactor. ...

  19. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    Wolf Creek Nuclear Optg Corp 1 Plant 1 Reactor 1,160 9,556 100.0 Note: Totals may not ... 3112045 1,160 9,566 94.0 Data for 2010 PWR Pressurized Light Boiling Water Reactor. ...

  20. EIA - State Nuclear Profiles

    U.S. Energy Information Administration (EIA) (indexed site)

    System Energy Resources Inc 1 Plant 1 Reactor 1,251 9,643 100.0 Note: Totals may not ... 88.0 BWR 711985 1112024 1,251 9,643 88.0 Data for 2010 BWR Boiling Water Reactor. ...

  1. NUCLEAR POWER PLANT

    DOE Patents [OSTI]

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

    1963-05-14

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

  2. A I'J' ANCR-1115 UC-78 I S

    Office of Scientific and Technical Information (OSTI)

    p A I'J' ANCR-1115 UC-78 I S . BOILING WATER REACTOR-FULL LENGTH EMERGENCI CORE COOLING ... i : v a t n l y omned riphf's. . Light- Water Reactor Technology TID-4500 I - N O T I C ...

  3. RECOVERY OF ACTINIDES FROM AQUEOUS NITRIC ACID SOLUTIONS

    DOE Patents [OSTI]

    Ader, M.

    1963-11-19

    A process of recovering actinides is presented. Tetravalent actinides are extracted from rare earths in an aqueous nitric acid solution with a ketone and back-extracted from the ketone into an aqueous medium. The aqueous actinide solution thus obtained, prior to concentration by boiling, is sparged with steam to reduce its ketone to a maximum content of 3 grams per liter. (AEC)

  4. Method for hydrocracking a heavy polynuclear hydrocarbonaceous feedstock in the presence of a molten metal halide catalyst

    DOE Patents [OSTI]

    Gorin, Everett

    1981-01-01

    A method for hydrocracking a heavy polynuclear hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the feedstock with hydrogen in the presence of a molten metal halide catalyst, the method comprising: mixing the feedstock with a heavy naphtha fraction which has an initial boiling point from about 100.degree. to about 160.degree. C. with a boiling point difference between the initial boiling point and the final boiling point of no more than about 50.degree. C. to produce a mixture; thereafter contacting the mixture with partially spent molten metal halide and hydrogen under temperature and pressure conditions so that the temperature is near the critical temperature of the heavy naphtha fraction; separating at least a portion of the heavy naphtha fraction and lighter hydrocarbon fuels from the partially spent molten metal halide, unreacted feedstock and reaction products; thereafter contacting the partially spent molten metal halide, unreacted feedstock and reaction products with hydrogen and fresh molten metal halide in a hydrocracking zone to produce additional lighter hydrocarbon fuels and separating at least a major portion of the lighter hydrocarbon fuels from the spent molten metal halide.

  5. Fuel and fuel blending components from biomass derived pyrolysis oil

    SciTech Connect (OSTI)

    McCall, Michael J.; Brandvold, Timothy A.; Elliott, Douglas C.

    2012-12-11

    A process for the conversion of biomass derived pyrolysis oil to liquid fuel components is presented. The process includes the production of diesel, aviation, and naphtha boiling point range fuels or fuel blending components by two-stage deoxygenation of the pyrolysis oil and separation of the products.

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

    DOE Patents [OSTI]

    Tokarz, R.D.

    1981-10-27

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

  7. Apparatus for upgrading heavy hydrocarbons employing a diluent

    SciTech Connect (OSTI)

    Calderon, J. L.; Layrisse, I.

    1985-06-04

    Apparatus for upgrading heavy hydrocarbonaceous materials for making coke suitable for metallurgical purposes comprises mixing the heavy hydrocarbonaceous materials with a diluent having a closely controlled boiling range so as to facilitate transport, dehydration and desalting of the crude oil. In addition, the diluent aids in controlling temperature and residence time of the crude thereby avoiding premature decomposition.

  8. Combined process for heavy oil, upgrading and synthetic fuel production

    SciTech Connect (OSTI)

    Polomski, R.E.

    1984-06-05

    A process for upgrading heavy oil to fuel products comprises deasphalting the heavy oil with an oxygenated solvent and simultaneously converting the oxygenated solvent and deasphalted oil over a ZSM-5 type catalyst to produce gasoline and distillate boiling range hydrocarbons.

  9. Process for upgrading heavy hydrocarbons employing a diluent

    SciTech Connect (OSTI)

    Calderon, J.L.; Layrisse, I.

    1984-06-19

    A process for upgrading heavy hydrocarbonaceous materials for making coke suitable for metallurgical purposes comprises mixing the heavy hydrocarbonaceous materials with a diluent having a closely controlled boiling range so as to facilitate transport, dehydration and desalting of the crude oil. In addition, the diluent aids in controlling temperature and residence time of the crude thereby avoiding premature decomposition.

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

    DOE Patents [OSTI]

    Lloyd, Milton H.; Collins, Jack L.; Shell, Sam E.

    1985-01-01

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

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

    DOE Patents [OSTI]

    Lloyd, M.H.; Collins, J.L.; Shell, S.E.

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

  12. Heat reclaiming method and apparatus

    DOE Patents [OSTI]

    Jardine, Douglas M.

    1984-01-01

    Method and apparatus to extract heat by transferring heat from hot compressed refrigerant to a coolant, such as water, without exceeding preselected temperatures in the coolant and avoiding boiling in a water system by removing the coolant from direct or indirect contact with the hot refrigerant.

  13. Enhanced shell-and-tube heat eschangers for the power and process industries. Final report

    SciTech Connect (OSTI)

    Bergles, A.E.; Jensen, M.K.; Somerscales, E.F.; Curcio, L.A. Jr.; Trewin, R.R.

    1994-08-01

    Single-tube pool boiling tests were performed with saturated pure refrigerants and binary mixtures of refrigerants. Generally, with pure refrigerants, the High Flux surface performed better at the higher heat fluxes compared to the Turbo-B tube, and both enhanced surfaces performed significantly better than smooth surface. In tests of R-11/R-113 mixtures, the enhanced surfaces had much less degradation in heat transfer coefficient due to mixture effects compared to smooth tubes; the largest degradation occurred at a mixture of 25% R-11/75% R-113. Under boiling in saturated aqueous solution of calcium sulfate, with a single tube, effects of fouling were more pronounced at the higher heat fluxes for all surfaces. Two staggered tube bundles were tested with tube pitch-diameter ratios of 1.17 and 1.50. For the pure refrigerant, tests on the smooth-tube bundle indicated that the effects on the heat transfer coefficient of varying mass flux, quality, and tube-bundle geometry were small, except at low heat fluxes. Neither enhanced surface showed any effect with changing mass flux or quality. The binary mixture bundle-boiling tests had results that were very similar to those obtained with the pure refrigerants. When boiling a refrigerant-oil mixture, all three surfaces (smooth, High Flux, and Turbo-B) experienced a degradation in its heat transfer coefficient; no surface studied was found to be immune or vulnerable to the presence of oil than another surface.

  14. Studies of stress corrosion cracking in steels used for reactor pressure vessels

    SciTech Connect (OSTI)

    Van Der Sluys, W.A.; Pathania, R.

    1992-12-31

    This paper reviews the state of technology concerning stress corrosion crack growth in LWR environments and reports the results from a series of experiments that attempted to duplicate the results obtained from the literature. These experiments include one conducted in a steam environment representative of the condition in the upper head of a boiling-water reactor (BWR).

  15. Loss of spent fuel pool cooling PRA: Model and results

    SciTech Connect (OSTI)

    Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

    1996-09-01

    This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 {times} 10{sup {minus}5} and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 {times} 10{sup {minus}3}. Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible.

  16. Spent Nuclear Fuel

    Gasoline and Diesel Fuel Update

    1968 through 2002 1968 through June 30, 2013 Increase Boiling-water reactor 89,843 136,821 46,978 Pressurized-water reactor 69,352 104,647 35,295 Total 159,195 241,468 82,273 ...

  17. BWR containment failure analysis during degraded-core accidents

    SciTech Connect (OSTI)

    Yue, D.D.

    1982-06-06

    This paper presents a containment failure mode analysis during a spectrum of postulated degraded core accident sequences in a typical 1000-MW(e) boiling water reactor (BWR) with a Mark-I wetwell containment. Overtemperature failure of containment electric penetration assemblies (CEPAs) has been found to be the major failure mode during such accidents.

  18. Carbon-13 variations in fluids from the Cerro Prieto geothermal system

    SciTech Connect (OSTI)

    Janik, C.J.; Nehring, N.L.; Huebner, M.A.; Truesdell, A.H.

    1982-08-10

    The carbon isotope compositions of CO/sub 2/ in steam from Cerro Prieto production well have been measured for 1977, 1979, and 1982. Variations in the delta/sup 13/C values are caused by production-related changes in the chemical and physical parameters of the geothermal system. In 1977, most CO/sub 2/ in the reservoir was isotopically light (delta/sup 13/C = -6.4 +/- 0.4). Heavier CO/sub 2/ was produced from wells in the center of the field (M5,M26,M27) due to deposition of isotopically light calcite caused by near-well boiling. In 1979 nearly all well showed relatively heavy CO/sub 2/, probably due to expansion of aquifer boiling and calcite precipitation. In 1982, many wells in the central part of the field were shut in. The amount of drawndown decreased and as temperatures and pressures near the wells increased, the boiling zones collapsed. The CO/sub 2/ in the fluid then exchanged with the precipitated calcite and became isotopically lighter. The sensitivity of carbon isotopes to calcite precipitations caused by aquifer boiling and to reequilibration with this deposited calcite upon decrease of boiling suggests use as an indicator of these aquifer processes. Surficial CO/sub 2/ of thermal origin was collected in 1981. Generally, the carbon-13 contents were close to CO/sub 2/ from production wells except for high-temperature mud pots and fumaroles containing isotopically light CO/sub 2/ derived from near surface alteration of organic matter.

  19. Sodium reflux pool-boiler solar receiver on-sun test results

    SciTech Connect (OSTI)

    Andraka, C E; Moreno, J B; Diver, R B; Moss, T A

    1992-06-01

    The efficient operation of a Stirling engine requires the application of a high heat flux to the relatively small area occupied by the heater head tubes. Previous attempts to couple solar energy to Stirling engines generally involved directly illuminating the heater head tubes with concentrated sunlight. In this study, operation of a 75-kW{sub t} sodium reflux pool-boiler solar receiver has been demonstrated and its performance characterized on Sandia's nominal 75-kW{sub t} parabolic-dish concentrator, using a cold-water gas-gap calorimeter to simulate Stirling engine operation. The pool boiler (and more generally liquid-metal reflux receivers) supplies heat to the engine in the form of latent heat released from condensation of the metal vapor on the heater head tubes. The advantages of the pool boiler include uniform tube temperature, leading to longer life and higher temperature available to the engine, and decoupling of the design of the solar absorber from the engine heater head. The two-phase system allows high input thermal flux, reducing the receiver size and losses, therefore improving system efficiency. The receiver thermal efficiency was about 90% when operated at full power and 800{degree}C. Stable sodium boiling was promoted by the addition of 35 equally spaced artificial cavities in the wetted absorber surface. High incipient boiling superheats following cloud transients were suppressed passively by the addition of small amounts of xenon gas to the receiver volume. Stable boiling without excessive incipient boiling superheats was observed under all operating conditions. The receiver developed a leak during performance evaluation, terminating the testing after accumulating about 50 hours on sun. The receiver design is reported here along with test results including transient operations, steady-state performance evaluation, operation at various temperatures, infrared thermography, x-ray studies of the boiling behavior, and a postmortem analysis.

  20. Superconducting cable cooling system by helium gas and a mixture of gas and liquid helium

    DOE Patents [OSTI]

    Dean, John W.

    1977-01-01

    Thermally contacting, oppositely streaming cryogenic fluid streams in the same enclosure in a closed cycle that changes from a cool high pressure helium gas to a cooler reduced pressure helium fluid comprised of a mixture of gas and boiling liquid so as to be near the same temperature but at different pressures respectively in go and return legs that are in thermal contact with each other and in thermal contact with a longitudinally extending superconducting transmission line enclosed in the same cable enclosure that insulates the line from the ambient at a temperature T.sub.1. By first circulating the fluid in a go leg from a refrigerator at one end of the line as a high pressure helium gas near the normal boiling temperature of helium; then circulating the gas through an expander at the other end of the line where the gas becomes a mixture of reduced pressure gas and boiling liquid at its boiling temperature; then by circulating the mixture in a return leg that is separated from but in thermal contact with the gas in the go leg and in the same enclosure therewith; and finally returning the resulting low pressure gas to the refrigerator for compression into a high pressure gas at T.sub.2 is a closed cycle, where T.sub.1 >T.sub.2, the temperature distribution is such that the line temperature is nearly constant along its length from the refrigerator to the expander due to the boiling of the liquid in the mixture. A heat exchanger between the go and return lines removes the gas from the liquid in the return leg while cooling the go leg.

  1. Cryogenic Current Lead Analysis Model Program

    Energy Science and Technology Software Center (OSTI)

    1992-01-01

    CCLAMP was developed to provide a tool for tha analysis of superconducting or normal current leads used to supply electricity from a warm interface (usually room temperature) to a device at cryogenic temperatures. It determines the heat leak to the cryogenic connection and the mass flow of the cryogen (typically helium) for the lead and installation modelled. It may be used to thermally optimize a lead design for a particular application. The user provides relevantmore » geometry details to model the electrical (length, diameter, superconducting length) and heat exchanger design of the lead (heat transfer coefficient, heat transfer area). It has a transient analysis capability so that lead transients such as cool down, current ramping, flow disruptions, and control simulations can be performed.« less

  2. I

    Office of Legacy Management (LM)

    ?am-3 . ,' .*. . - yp: -.* : .- ., ._ ' Yi * <. ? :+". thfa prcbputir. 80,UUU lb. of tmmiuu, J.m,cDu lb. of 3wukdlw crper' tiwu 5.8 t&i8 l atr:irur ral u&d i.Wttd&?# Bir;n8 i;orammant end rUl rid nrtrlcial by uo&utboFlwd putqlm. ). The ~&&a, ' 8m ;altielJ 79 p-rmlt arrgora ted and ttw tap t.ha aikalini~, . L pokotlal brlf)r, bU88M 8-i .ii.i co# sat8 awtaet wltb the mBtmtl8a. aada q*iast fb a8v0-*..u @ow +.ta p-?Y h&al. . .; . ' 6 G.. ..*... . ,,z.. ,. ..*,::

  3. Ou,I~

    Office of Legacy Management (LM)

    ' $1 ;- / J. s. Cuidor, t?lrsckor, AbfnFstr&iqe L"~rz~fons 317ision t. .?, ' GwkwrJ, Chief, ?mp?rty "or??r.c.h 33-l i L!C3?~~~CH - !ST.dJ. r;c.pi, f:i!.!:&L: il<:-[,TC: Ij$ ';cto?w? X2, l?~- NY. 17 - *. . . -'p&. - _ _: Ou,I~ QMt w : -_ . .- . _I --_ . . - B ! /- 4 Id : ji&Y;~ .__ . I .m:;qy& * k Cctcber 1, and 5, 1-351, tha f~3ltirz~ xere jw33ent d&icg 32 . ~~3,?2~tifXt of Ytecl a.nd izm 3wn_p ~c~~3~lsti~3s et L4ka CCt.zJ."fQ XOr3,~9 Atua: ;-Ir. .i,llpb

  4. Method of detecting genetic deletions identified with chromosomal abnormalities

    DOE Patents [OSTI]

    Gray, Joe W; Pinkel, Daniel; Tkachuk, Douglas

    2013-11-26

    Methods and compositions for staining based upon nucleic acid sequence that employ nucleic acid probes are provided. Said methods produce staining patterns that can be tailored for specific cytogenetic analyzes. Said probes are appropriate for in situ hybridization and stain both interphase and metaphase chromosomal material with reliable signals. The nucleic acids probes are typically of a complexity greater tha 50 kb, the complexity depending upon the cytogenetic application. Methods and reagents are provided for the detection of genetic rearrangements. Probes and test kits are provided for use in detecting genetic rearrangements, particlularly for use in tumor cytogenetics, in the detection of disease related loci, specifically cancer, such as chronic myelogenous leukemia (CML) and for biological dosimetry. Methods and reagents are described for cytogenetic research, for the differentiation of cytogenetically similar ut genetically different diseases, and for many prognostic and diagnostic applications.

  5. Biology Division progress report, October 1, 1984-September 30, 1985

    SciTech Connect (OSTI)

    Not Available

    1986-01-01

    The body of this report provides summaries of the aims, scope and progress of the research by groups of investigators in the Division during the period of October 1, 1984, through September 30, 1985. At the end of each summary is a list of publications covering the same period. For convenience, the summaries are assembled under Sections in accordance with the current organizational structure of the Biology Division; each Section begins with an overview. It will be apparent, however, tha crosscurrents run throughout the Division and that the various programs support and interact with each other. In addition, this report includes information on the Division's educational activities, Advisory Committee, seminar program, and international interactions, as well as extramural activities of staff members, abstracts for technical meetings, and funding and personnel levels.

  6. To study of different level of nitrogen manure and density on yield and yield component of variety of K.S.C 704 in dry region of sistan

    SciTech Connect (OSTI)

    Dahmardeh, M.; Forghani, F.; Khammari, E.

    2008-01-30

    Out of three grain of the world, Corn is one of the best, About 7 to 10 thousand years ago in south of Mexico corn become domesticated. In the year 1995 culfivation of corn in the world was 130 mil/ha, and to Total production of the world of corn is 507 M/Tons. Average yield of corn in the year 1995 Among Producer countries was 7.78 To 7.60 t/ha in fance and united state was state was 2.36 To 2.20 t/ha, but in Brazil and Mexico Production of corn was different. With this regards, special manner has been arranged for the suitable cultivation or suitable density plants in one heactar on cultivation variety of K.S.C 704 corn. Also suitable level of Nitrogen manure, this Protect in climatic condition of Sistan region done, sith complete block design with 3 replication. Experiment has been selected as split plot, the main plot with 4 different concentration level such as (200-250-3500 and 350 Kg/ha) and sub plot density with 3 different level such as 111000,83000 and 66000 plan/ha respectively. From stage growth up to harvesting of corn in this reache having Data for each treat. ment, After harvesting Analysis of variance and companion of Average of each treatment has been done by DunKan method. Results has been shown, Measurment of characteristics (yield component) seed yield effected different density level of manure, with increasing of manure weight of one thousand seed yield and also in high density showed high significant differente amoung each other. These are with suitable climatic condition of sistan region if enough water will be available ed using Amount of 350 ks/ha Nitrogen manure and with density 111000 plants/ha we can product suitable seed yield Biological yield.

  7. The effect of fuel thermal conductivity on the behavior of LWR cores during loss-of-coolant accidents

    SciTech Connect (OSTI)

    Terrani, Kurt A.; Wang, Dean; Ott, Larry J.; Montgomery, Robert O.

    2014-05-01

    The effect of variation in thermal conductivity of light water reactor fuel elements on core response during loss-of-coolant accident scenarios is examined. Initially, a simplified numerical analysis is utilized to determine the time scales associated with dissipation of stored energy from the fuel into the coolant once the fission reaction is stopped. The analysis is then followed by full reactor system thermal-hydraulics analysis of a typical boiling and pressurized water reactor subjected to a large break loss-of-coolant accident scenario using the TRACE code. Accordingly, sensitivity analyses to examine the effect of an increase in fuel thermal conductivity, up to 500%, on fuel temperature evolution during these transients are performed. Given the major differences in thermal-hydraulics design aspects of boiling and pressurized water reactors, different fuel and temperature responses during the simulated loss-of-coolant transients are observed.

  8. RAMONA-3B calculations for Browns Ferry ATWS (Anticipated Transient Without Scram) study

    SciTech Connect (OSTI)

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

    1987-02-01

    Several aspects of the Anticipated Transient Without Scram (ATWS) initiated by an inadvertent closure of all Main Steam Isolation Valves (MSIV) in a typical BWR/4 are analyzed in the report. The analysis is performed using the Brookhaven National Laboratory code, RAMONA-3B, which employs a three-dimensional neutron kinetics model coupled with a parallel-channel thermal hydraulics in representing a Boiling Water Reactor (BWR) Core. Four different transient scenarios have been investigated: (a) downcomer water level and reactor pressure control, (b) manual control rod insertion transient, (c) high pressure boil-off, and (d) recirculation pump trip failure. Results of these calculations should provide better understanding of mitigative effects of operator actions during ATWS, thus helping in the development of adequate Emergency Procedure Guidelines (EPG) required for the BWR plant safety. A few unresolved questions subject to future investigations are also discussed.

  9. The coolability limits of a reactor pressure vessel lower head

    SciTech Connect (OSTI)

    Theofanous, T.G.; Syri, S.

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  10. Catalytic two-stage coal liquefaction process having improved nitrogen removal

    DOE Patents [OSTI]

    Comolli, Alfred G.

    1991-01-01

    A process for catalytic multi-stage hydrogenation and liquefaction of coal to produce high yields of low-boiling hydrocarbon liquids containing low concentrations of nitogen compounds. First stage catalytic reaction conditions are 700.degree.-800.degree. F. temperature, 1500-3500 psig hydrogen partial pressure, with the space velocity maintained in a critical range of 10-40 lb coal/hr ft.sup.3 catalyst settled volume. The first stage catalyst has 0.3-1.2 cc/gm total pore volume with at least 25% of the pore volume in pores having diameters of 200-2000 Angstroms. Second stage reaction conditions are 760.degree.-870.degree. F. temperature with space velocity exceeding that in the first stage reactor, so as to achieve increased hydrogenation yield of low-boiling hydrocarbon liquid products having at least 75% removal of nitrogen compounds from the coal-derived liquid products.

  11. Fission gas induced fuel swelling in low and medium burnup fuel during high temperature transients. [PWR

    SciTech Connect (OSTI)

    Vinjamuri, K.

    1980-01-01

    The behavior of light water reactor fuel elements under postulated accident conditions is being studied by the EG and G Idaho, Inc., Thermal Fuels Behavior Program for the Nuclear Regulatory Commission. As a part of this program, unirradiated and previously irradiated, pressurized-water-reactor type fuel rods were tested under power-cooling-mismatch (PCM) conditions in the Power Burst Facility (PBF). During these integral in-reactor experiments, film boiling was produced on the fuel rods which created high fuel and cladding temperatures. Fuel rod diameters increased in the film boiling region to a greater extent for irradiated rods than for unirradiated rods. The purpose of the study was to investigate and assess the fuel swelling which caused the fuel rod diameter increases and to evaluate the ability of an analytical code, the Gas Release and Swelling Subroutine - Steady-State and Transient (GRASS-SST), to predict the results.

  12. ATOMIC POWER PLANT

    DOE Patents [OSTI]

    Daniels, F.

    1957-11-01

    This patent relates to neutronic reactor power plants and discloses a design of a reactor utilizing a mixture of discrete units of a fissionable material, such as uranium carbide, a neutron moderator material, such as graphite, to carry out the chain reaction. A liquid metal, such as bismuth, is used as the coolant and is placed in the reactor chamber with the fissionable and moderator material so that it is boiled by the heat of the reaction, the boiling liquid and vapors passing up through the interstices between the discrete units. The vapor and flue gases coming off the top of the chamber are passed through heat exchangers, to produce steam, for example, and thence through condensers, the condensed coolant being returned to the chamber by gravity and the non- condensible gases being carried off through a stack at the top of the structure.

  13. Methods of producing transportation fuel

    DOE Patents [OSTI]

    Nair, Vijay; Roes, Augustinus Wilhelmus Maria; Cherrillo, Ralph Anthony; Bauldreay, Joanna M.

    2011-12-27

    Systems, methods, and heaters for treating a subsurface formation are described herein. At least one method for producing transportation fuel is described herein. The method for producing transportation fuel may include providing formation fluid having a boiling range distribution between -5.degree. C. and 350.degree. C. from a subsurface in situ heat treatment process to a subsurface treatment facility. A liquid stream may be separated from the formation fluid. The separated liquid stream may be hydrotreated and then distilled to produce a distilled stream having a boiling range distribution between 150.degree. C. and 350.degree. C. The distilled liquid stream may be combined with one or more additives to produce transportation fuel.

  14. Thermal treatment of low permeability soils using electrical resistance heating

    SciTech Connect (OSTI)

    Udell, K.S.

    1996-08-01

    The acceleration of recovery rates of second phase liquid contaminants from the subsurface during gas or water pumping operations is realized by increasing the soil and ground water temperature. Electrical heating with AC current is one method of increasing the soil and groundwater temperature and has particular applicability to low permeability soils. Several mechanisms have been identified that account for the enhanced removal of the contaminants during electrical heating. These are vaporization of liquid contaminants with low boiling points, temperature-enhanced evaporation rates of semi-volatile components, and removal of residual contaminants by the boiling of residual water. Field scale studies of electrical heating and fluid extraction show the effectiveness of this technique and its applicability to contaminants found both above and below the water table and within low permeability soils. 10 refs., 8 figs.

  15. Effectiveness of decanter modifications on organic removal

    SciTech Connect (OSTI)

    Lambert, D.P.

    1992-08-20

    A series of runs were planned in the Precipitate Hydrolysis Experimental Facility (PHEF) at the Savannah River Plant to determine the effectiveness of equipment and process modifications on the PHEF decanter organic removal efficiency. Runs 54-59 were planned to test the effectiveness of spray recirculation, a new decanter, heated organic recirculation and aqueous drawoff on organic removal efficiency in the revised HAN flowsheet. Runs 60-63 were planned to provide a comparison of the original and new decanter designs on organic removal efficiency in the late wash flowsheet without organic recirculation. Operational problems were experienced in both the PHEF and IDMS pilot facilities because of the production of high boiling organics and the low organic removal efficiency of the PHEF decanters. To prevent these problems in the DWPF Salt and Chemical Cells, modifications were proposed to the decanter and flowsheet to maximize the organic removal efficiency and minimize production of high boiling organics.

  16. Generation of energy

    DOE Patents [OSTI]

    Kalina, Alexander I.

    1984-01-01

    A method of generating energy which comprises utilizing relatively lower temperature available heat to effect partial distillation of at least portion of a multicomponent working fluid stream at an intermediate pressure to generate working fluid fractions of differing compositions. The fractions are used to produce at least one main rich solution which is relatively enriched with respect to the lower boiling component, and to produce at least one lean solution which is relatively improverished with respect to the lower boiling component. The pressure of the main rich solution is increased whereafter it is evaporated to produce a charged gaseous main working fluid. The main working fluid is expanded to a low pressure level to release energy. The spent low pressure level working fluid is condensed in a main absorption stage by dissolving with cooling in the lean solution to regenerate an initial working fluid for reuse.

  17. Upgrading of heavy hydrocarbonaceous feeds

    SciTech Connect (OSTI)

    Bhattacharya, A.K.; Storm, D.A.; DeRosa, T.F.

    1995-12-31

    This paper is based on our work in the area of upgrading of heavy hydrocarbonaceous feedstocks. The work involves the development of a method of catalytically hydroconverting a hydrocarbon feed stream containing a substantial quantity of components boiling above about 538{degrees}C to a substantial portion thereof to components boiling below 538{degrees}C. More particularly, an oil-miscible or oil-soluble poly(ether)diol or a derivative thereof and an aromatic additive oil, such as Heavy Cycle Gas oil, are added to a heavy hydrocarbon feed and the mixed stream is contacted at elevated temperatures with a solid catalyst such as a sulfided nickel molybdenum oxide on alumina in the presence of hydrogen under pressure. This method advantageously affords higher conversion to more valuable liquid products containing lower amounts of heteroatoms such as sulfur, substantially eliminates plugging of the hydroconversion reactor, and reduces the amount of insolubles in the total liquid product.

  18. Process for catalytic cracking of heavy hydrocarbon feed to lighter products

    SciTech Connect (OSTI)

    Herbst, J.A.; Owen, H.; Schipper, P.H.

    1990-05-29

    This patent describes a process for catalytic cracking of a feed of hydrocarbons boiling in the gas oil and heavier boiling range to lighter products by contacting the feed at catalytic cracking conditions and catalytically cracking the feed to lighter products with a cracking catalyst. It comprises: a mixture of separate particles of: a bulk conversion cracking catalyst containing at least one component with an equivalent pore size of at least about 7 angstroms in a matrix, the bulk conversion cracking catalyst having fluidization properties which permit use in a fluidized or moving bed catalytic cracking reactor; a light paraffin upgrading catalyst comprising at least one zeolite having a constraint index of 1--12 and paraffin cracking/isomerization activity; and, a light paraffin upgrading catalyst comprising at least one zeolite having a constraint index of 1--12 and paraffin aromatization activity; and wherein the upgrading catalysts have substantially the same fluidization properties as the bulk conversion cracking catalyst.

  19. Study on release and transport of aerial radioactive materials in reprocessing plants

    SciTech Connect (OSTI)

    Amano, Y.; Tashiro, S.; Uchiyama, G.; Abe, H.; Yamane, Y.; Yoshida, K.; Kodama, T.

    2013-07-01

    The release and transport characteristics of radioactive materials at a boiling accident of the high active liquid waste (HALW) in a reprocessing plant have been studied for improving experimental data of source terms of the boiling accident. In the study, a heating test and a thermogravimetry and differential thermal analysis (TG-DTA) test were conducted. In the heating test using a simulated HALW, it was found that ruthenium was mainly released into the air in the form of gas and that non-volatile elements were released into the air in the form of mist. In the TG-DTA test, the rate constants and reaction heat of thermal decomposition of ruthenium nitrosyl nitrate were obtained from TG and DTA curves. (authors)

  20. Three-stage hydrocracking process

    SciTech Connect (OSTI)

    Bachtel, R.W.; Cash, D.R.

    1983-09-13

    A hydrocracking process is disclosed of improved flexibility comprises three hydroprocessing stages and a fractionating zone such that the effluent from the first hydroprocessing stage is fractionated into a first fraction consisting of gasoline and lower boiling components, a second fraction, the residuum, having a normal boiling point greater than about 700/sup 0/ F. (370/sup 0/ C.), and a third fraction, the middle and heavy distillate, consisting of the remainder of the effluent of the first hydroprocessing stage. The second fraction undergoes hydrocracking in a second hydroprocessing stage with recycle of the effluent from said second hydroprocessing stage to the fractionation zone, and the third fraction undergoes hydrocracking in a third hydrocracking stage.

  1. Viscosity stabilization of SRC residual oil. Final technical report

    SciTech Connect (OSTI)

    Tewari, K.C.

    1984-05-01

    The use of SRC residual oils for No. 6 Fuel Oil substitutes has been proposed. The oils exhibit viscosity characteristics at elevated temperatures that allow this substitution with only minor modifications to the existing fuel oil infrastructure. However, loss of low-boiling materials causes an increase in the viscosity of the residual oils that is greater than expected from concentration changes. A process has been developed that minimizes the loss of volatiles and thus maintains the viscosity of these materials. The use of an additive (water, phenol, or an SRC light oil cut rich in low-boiling phenols in amounts up to 2.0 wt %) accomplishes this and hence stabilizes the pumping and atomizing characteristics for an extended period. During the course of the work, the components of the volatiles lost were identified and the viscosity change due to this loss was quantified. 3 references, 6 figures, 9 tables.

  2. Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

    SciTech Connect (OSTI)

    Arave, A.E.; Rodman, G.R.

    1992-12-01

    The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D&D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D&D plans for the turbine building were prepared from 1979 through 1990. D&D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and the absence of loose contamination, the D&D activities were completed with no radiation exposure to the workers. The D&D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.

  3. Synthetic crude oils carcinogenicity screening tests. Progress report, September 15, 1979-March 15, 1980

    SciTech Connect (OSTI)

    Calkins, W.H.; Deye, J.F.; King, C.F.; Hartgrove, R.W.; Krahn, D.F.

    1980-01-01

    Four crude oils (H Coal-Fuel Oil Mode, Occidental in situ Shale Oil, Exxon Donor Solvent Liquid, and SRC II) which were distilled into four fractions (naphtha, mid-distillate, gas oil and residue) for analysis and biological screening testing during the last report period were tested for mutagenicity by the Ames test and for tumor initiating activity by an initiation/promotion (skin painting) test. Substantial agreement exists between Ames and skin painting results. Low boiling naphtha fractions of the 4 crude oils showed little or no mutagenicity or tumor initiating activity by the two tests used. The higher boiling fractions (gas oils and residues) and the crude oils themselves were mutagenic and exhibited tumor initiation activity. The coal derived fractions were more active by both tests than the shale oil fractions.

  4. Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

    SciTech Connect (OSTI)

    Arave, A.E.; Rodman, G.R.

    1992-12-01

    The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D D plans for the turbine building were prepared from 1979 through 1990. D D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and the absence of loose contamination, the D D activities were completed with no radiation exposure to the workers. The D D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.

  5. System transient response to loss of off-site power

    SciTech Connect (OSTI)

    Sozer, A.

    1990-01-01

    A simultaneous trip of the reactor, main circulation pumps, secondary coolant pumps, and pressurizer pump due to loss of off-site power at the High Flux Isotope Reactor (HFIR) located at the Oak Ridge National Laboratory (ORNL) has been analyzed to estimate available safety margin. A computer model based on the Modular Modeling System code has been used to calculate the transient response of the system. The reactor depressurizes from 482.7 psia down to about 23 psia in about 50 seconds and remains stable thereafter. Available safety margin has been estimated in terms of the incipient boiling heat flux ratio. It is a conservative estimate due to assumed less than available primary and secondary flows and higher than normal depressurization rate. The ratio indicates no incipient boiling conditions at the hot spot. No potential damage to the fuel is likely to occur during this transient. 2 refs., 6 figs.

  6. Enrichment of light hydrocarbon mixture

    DOE Patents [OSTI]

    Yang, Dali; Devlin, David; Barbero, Robert S.; Carrera, Martin E.; Colling, Craig W.

    2011-11-29

    Light hydrocarbon enrichment is accomplished using a vertically oriented distillation column having a plurality of vertically oriented, nonselective micro/mesoporous hollow fibers. Vapor having, for example, both propylene and propane is sent upward through the distillation column in between the hollow fibers. Vapor exits neat the top of the column and is condensed to form a liquid phase that is directed back downward through the lumen of the hollow fibers. As vapor continues to ascend and liquid continues to countercurrently descend, the liquid at the bottom of the column becomes enriched in a higher boiling point, light hydrocarbon (propane, for example) and the vapor at the top becomes enriched in a lower boiling point light hydrocarbon (propylene, for example). The hollow fiber becomes wetted with liquid during the process.

  7. Enrichment of light hydrocarbon mixture

    DOE Patents [OSTI]

    Yang; Dali; Devlin, David; Barbero, Robert S.; Carrera, Martin E.; Colling, Craig W.

    2010-08-10

    Light hydrocarbon enrichment is accomplished using a vertically oriented distillation column having a plurality of vertically oriented, nonselective micro/mesoporous hollow fibers. Vapor having, for example, both propylene and propane is sent upward through the distillation column in between the hollow fibers. Vapor exits neat the top of the column and is condensed to form a liquid phase that is directed back downward through the lumen of the hollow fibers. As vapor continues to ascend and liquid continues to countercurrently descend, the liquid at the bottom of the column becomes enriched in a higher boiling point, light hydrocarbon (propane, for example) and the vapor at the top becomes enriched in a lower boiling point light hydrocarbon (propylene, for example). The hollow fiber becomes wetted with liquid during the process.

  8. NREL: Transportation Research - Power Electronics Thermal Management

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Power Electronics Thermal Management A photo of water boiling in liquid cooling lab equipment. Power electronics thermal management research aims to help lower the cost and improve the performance of electric-drive vehicles. Photo by Dennis Schroeder, NREL NREL investigates and develops thermal management strategies for power electronics systems that use wide-bandgap technology, which enables the development of devices that are smaller than those based on other materials, demonstrating

  9. ARM - Temperature Converter

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    CalculatorsTemperature Converter Outreach Home Room News Publications Traditional Knowledge Kiosks Barrow, Alaska Tropical Western Pacific Site Tours Contacts Students Study Hall About ARM Global Warming FAQ Just for Fun Meet our Friends Cool Sites Teachers Teachers' Toolbox Lesson Plans Temperature Converter The Fahrenheit scale, invented by German physicist Daniel Gabriel Fahrenheit (1686-1736), is based on 32 °F for the freezing point of water and 212 °F for the boiling point of water. The

  10. Mixed feed evaporator

    DOE Patents [OSTI]

    Vakil, Himanshu B. (Schenectady, NY); Kosky, Philip G. (Ballston Lake, NY)

    1982-01-01

    In the preparation of the gaseous reactant feed to undergo a chemical reaction requiring the presence of steam, the efficiency of overall power utilization is improved by premixing the gaseous reactant feed with water and then heating to evaporate the water in the presence of the gaseous reactant feed, the heating fluid utilized being at a temperature below the boiling point of water at the pressure in the volume where the evaporation occurs.

  11. Numerical Studies of Fluid Leakage from a Geologic DisposalReservoir for CO2 Show Self-Limiting Feedback between Fluid Flow and HeatTransfer

    SciTech Connect (OSTI)

    Pruess, Karsten

    2005-03-22

    Leakage of CO2 from a hypothetical geologic storage reservoir along an idealized fault zone has been simulated, including transitions between supercritical, liquid, and gaseous CO2. We find strong non-isothermal effects due to boiling and Joule-Thomson cooling of expanding CO2. Leakage fluxes are limited by limitations in conductive heat transfer to the fault zone. The interplay between multiphase flow and heat transfer effects produces non-monotonic leakage behavior.

  12. Control system for fluid heated steam generator

    DOE Patents [OSTI]

    Boland, J.F.; Koenig, J.F.

    1984-05-29

    A control system for controlling the location of the nucleate-boiling region in a fluid heated steam generator comprises means for measuring the temperature gradient (change in temperature per unit length) of the heating fluid along the steam generator; means for determining a control variable in accordance with a predetermined function of temperature gradients and for generating a control signal in response thereto; and means for adjusting the feedwater flow rate in accordance with the control signal.

  13. Control system for fluid heated steam generator

    DOE Patents [OSTI]

    Boland, James F.; Koenig, John F.

    1985-01-01

    A control system for controlling the location of the nucleate-boiling region in a fluid heated steam generator comprises means for measuring the temperature gradient (change in temperature per unit length) of the heating fluid along the steam generator; means for determining a control variable in accordance with a predetermined function of temperature gradients and for generating a control signal in response thereto; and means for adjusting the feedwater flow rate in accordance with the control signal.

  14. Conflict Between Economic

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Conflict Between Economic Growth and Environmental Protection Dr. Brian Czech Advancement - Steady State Economy Monday, Jan 9, 2012 - 4:15PM MBG AUDITORIUM Refreshments at 4:00PM The confict between economic growth and environmental protection may not be reconciled via technological progress. The fundamentality of the confict ultimately boils down to laws of thermodynamics. Physicists and other scholars from the physical sciences are urgently needed for helping the public and policy makers

  15. Combined cold compressor/ejector helium refrigerator

    DOE Patents [OSTI]

    Brown, Donald P.

    1985-01-01

    A refrigeration apparatus having an ejector operatively connected with a cold compressor to form a two-stage pumping system. This pumping system is used to lower the pressure, and thereby the temperature of a bath of boiling refrigerant (helium). The apparatus as thus arranged and operated has substantially improved operating efficiency when compared to other processes or arrangements for achieving a similar low pressure.

  16. METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Reed, G.A.

    1961-01-10

    A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

  17. Liquid metal heat exchanger for efficient heating of soils and geologic formations

    DOE Patents [OSTI]

    DeVault, Robert C [Knoxville, TN; Wesolowski, David J [Kingston, TN

    2010-02-23

    Apparatus for efficient heating of subterranean earth includes a well-casing that has an inner wall and an outer wall. A heater is disposed within the inner wall and is operable within a preselected operating temperature range. A heat transfer metal is disposed within the outer wall and without the inner wall, and is characterized by a melting point temperature lower than the preselected operating temperature range and a boiling point temperature higher than the preselected operating temperature range.

  18. News | Solid State Solar Thermal Energy Conversion

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    News Sponge creates steam using ambient sunlight MIT engineers have invented a bubble-wrapped, sponge-like device that soaks up natural sunlight and heats water to boiling temperatures, generating steam through its pores. Read full news Physicists predict previously unseen phenomena in exotic materials MIT News highlighted work in the S3TEC center led by Liang Fu investigating topological semimetals. Read full news A nanophotonic comeback for incandescent bulbs? MIT News highlighted work in the

  19. Page

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Page 1 .........CIPS Simulation Capability Implemented in VERA 1 .........Watts Bar Operating Cycles Simulated to Present 1 .........VERA Training Workshop Held 2 .........Departure from Nucleate Boiling (DNB) Multi-Physics Approach & Applications using VERA 5 .........ITM/DNS Simulations for Closure Development 7 .........EPRI Test Stand Concluded 8 .........Verification and Validation Supporting VERA 12 .........Illustration of VERA Capability to Model a Typical SMR 14 .........VERA

  20. Integrated coal liquefaction process

    DOE Patents [OSTI]

    Effron, Edward

    1978-01-01

    In a process for the liquefaction of coal in which coal liquids containing phenols and other oxygenated compounds are produced during the liquefaction step and later hydrogenated, oxygenated compounds are removed from at least part of the coal liquids in the naphtha and gas oil boiling range prior to the hydrogenation step and employed as a feed stream for the manufacture of a synthesis gas or for other purposes.