National Library of Energy BETA

Sample records for reactor cooling system

  1. Liquid metal cooled nuclear reactors with passive cooling system

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Fanning, Alan W. (San Jose, CA)

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

  2. Reactor core isolation cooling system

    DOE Patents [OSTI]

    Cooke, F.E.

    1992-12-08

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

  3. Reactor core isolation cooling system

    DOE Patents [OSTI]

    Cooke, Franklin E. (San Jose, CA)

    1992-01-01

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

  4. Liquid metal cooled nuclear reactor plant system

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  5. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1990-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  6. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA); Hui, Marvin M. (Sunnyvale, CA); Berglund, Robert C. (Saratoga, CA)

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  7. Cooling system for a nuclear reactor

    DOE Patents [OSTI]

    Amtmann, Hans H. (Rancho Santa Fe, CA)

    1982-01-01

    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  8. Measurement of Temperature Profile in the Reactor Cavity Cooling System 

    E-Print Network [OSTI]

    Alhashimi, Tariq Yaqoob Sayed

    2014-12-02

    -cooled Reactor Cavity Cooling System (RCCS) based on General Electric Modular High Temperature Gas Cooled Reactor (MHTGR) design to study the thermal hydraulic phenomena occurring in the upper plenum. The facility consists of four vertical parallel riser ducts...

  9. Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Busboom, Herbert J. (San Jose, CA)

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

  10. Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1993-01-01

    A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

  11. Nuclear reactor cooling system decontamination reagent regeneration

    DOE Patents [OSTI]

    Anstine, Larry D. (San Jose, CA); James, Dean B. (Saratoga, CA); Melaika, Edward A. (Berkeley, CA); Peterson, Jr., John P. (Livermore, CA)

    1985-01-01

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  12. Passive cooling system for top entry liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Boardman, Charles E. (Saratoga, CA); Hunsbedt, Anstein (Los Gatos, CA); Hui, Marvin M. (Cupertino, CA)

    1992-01-01

    A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

  13. Passive cooling system for nuclear reactor containment structure

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Wade, Gentry E. (Saratoga, CA)

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  14. Natural circulating passive cooling system for nuclear reactor containment structure

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Wade, Gentry E. (Saratoga, CA)

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  15. System Study: Reactor Core Isolation Cooling 1998–2012

    SciTech Connect (OSTI)

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  16. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    SciTech Connect (OSTI)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  17. Monitoring system for a liquid-cooled nuclear fission reactor

    DOE Patents [OSTI]

    DeVolpi, Alexander (Bolingbrook, IL)

    1987-01-01

    A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

  18. THERMAL STRESS CALCULATIONS FOR HEATPIPE-COOLED REACTOR POWER SYSTEMS.

    SciTech Connect (OSTI)

    Kapernick, R. J. (Richard J.); Guffee, R. M. (Ray M.)

    2001-01-01

    A heatpipe-cooled fast reactor concept has been under development at Los Alamos National Laboratory for the past several years, to be used as a power source for nuclear electric propulsion (NEP) or as a planetary surface power system. The reactor core consists of an array of modules that are held together by a core lateral restraint system. Each module comprises a single heatpipe surrounded by 3-6 clad fuel pins. As part of the design development and performance assessment activities for these reactors, specialized methods and models have been developed to perform thermal and stress analyses of the core modules. The methods have been automated so that trade studies can be readily performed, looking at design options such as module size, heatpipe and clad thickness, use of sleeves to contain the fuel, material type, etc. This paper describes the methods and models that have been developed, and presents thermal and stress analysis results for a Mars surface power system and a NEP power source.

  19. Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

    DOE Patents [OSTI]

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

    1980-06-06

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  20. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    SciTech Connect (OSTI)

    Hassan, Yassin; Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  1. Method and apparatus for enhancing reactor air-cooling system performance

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA)

    1996-01-01

    An enhanced decay heat removal system for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer.

  2. Method and apparatus for enhancing reactor air-cooling system performance

    DOE Patents [OSTI]

    Hunsbedt, A.

    1996-03-12

    An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

  3. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    SciTech Connect (OSTI)

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions between the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very effective in removing decay heat. By removing the limit on the decay heat removal capability due to the limited available surface area as in a RVACS, the reactor power and power density can be significantly increased, without losing the passive heat removal feature. This paper will introduce the concept of using DRACS to enhance VHTR passive safety and economics. Three design options will be discussed, depending on the cooling pipe locations. Analysis results from a lumped volume based model and CFD simulations will be presented.

  4. Experimental and Computational Study of a Scaled Reactor Cavity Cooling System 

    E-Print Network [OSTI]

    Vaghetto, Rodolfo

    2013-11-25

    ), producing fission power in a circulating molten salt fuel mixture with an epithermal-spectrum reactor and a full actinide recycling fuel cycle. 7 Figure 4. Generation IV Nuclear Systems [7] (1: GFR; 2: VHTR; 3: SCWR; 4: SFR; 5: LFR; 6: MSR... Of Energy GFR Gas-cooled Fast Reactor GT-MHR Gas Turbine-Modular Helium Reactor HS Heat Structure IAEA International Atomic Energy Agency ID Inner Diameter INL Idaho National Laboratory IRUG International RELAP5 Users Group L Length LFR Lead...

  5. Passive decay heat removal system for water-cooled nuclear reactors

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1991-01-01

    A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

  6. Design of passive decay heat removal system for the lead cooled flexible conversion ratio fast reactor

    E-Print Network [OSTI]

    Whitman, Joshua (Joshua J.)

    2007-01-01

    The lead-cooled flexible conversion ratio fast reactor shows many benefits over other fast-reactor designs; however, the higher power rating and denser primary coolant present difficulties for the design of a passive decay ...

  7. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

    Corletti, Michael M. (New Kensington, PA); Lau, Louis K. (Monroeville, PA); Schulz, Terry L. (Murrysville Boro, PA)

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  8. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

    Corletti, M.M.; Lau, L.K.; Schulz, T.L.

    1993-12-14

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

  9. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    SciTech Connect (OSTI)

    Lv, Quiping; Sun, Xiaodong; Chtistensen, Richard; Blue, Thomas; Yoder, Graydon; Wilson, Dane

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  10. KEY DESIGN REQUIREMENTS FOR THE HIGH TEMPERATURE GAS-COOLED REACTOR NUCLEAR HEAT SUPPLY SYSTEM

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01

    Key requirements that affect the design of the high temperature gas-cooled reactor nuclear heat supply system (HTGR-NHSS) as the NGNP Project progresses through the design, licensing, construction and testing of the first of a kind HTGR based plant are summarized. These requirements derive from pre-conceptual design development completed to-date by HTGR Suppliers, collaboration with potential end users of the HTGR technology to identify energy needs, evaluation of integration of the HTGR technology with industrial processes and recommendations of the NGNP Project Senior Advisory Group.

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

    DOE Patents [OSTI]

    Hill, Paul R. (Tucson, AZ)

    1994-01-01

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

  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. Heat Transfer Simulation of Reactor Cavity Cooling System Experimental Facility using RELAP5-3D and Generation of View Factors using MCNP 

    E-Print Network [OSTI]

    Wu, Huali

    2013-08-08

    As one of the most attractive reactor types, The High Temperature Gas-cooled Reactor (HTGR) is designed to be passively safe with the incorporation of Reactor Cavity Cooling System (RCCS). In this paper, a RELAP5-3D simulation model is set up based...

  14. Update on Small Modular Reactors Dynamics System Modeling Tool -- Molten Salt Cooled Architecture

    SciTech Connect (OSTI)

    Hale, Richard Edward; Cetiner, Sacit M.; Fugate, David L.; Qualls, A L.; Borum, Robert C.; Chaleff, Ethan S.; Rogerson, Doug W.; Batteh, John J.; Tiller, Michael M.

    2014-08-01

    The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

  15. Initial Requirements for Gas-Cooled Fast Reactor (GFR) System Design, Performance, and Safety Analysis Models

    SciTech Connect (OSTI)

    Kevan D. Weaver; Thomas Y. C. Wei

    2004-08-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  16. Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors

    E-Print Network [OSTI]

    Zweibaum, Nicolas

    2015-01-01

    test SFR – Sodium-cooled fast reactor SWU – Separative workvalues than sodium-cooled fast reactors ( SFRs) and HTGRs.

  17. Monitoring system for a liquid-cooled nuclear fission reactor. [PWR

    DOE Patents [OSTI]

    DeVolpi, A.

    1984-07-20

    The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

  18. Passive containment cooling system

    DOE Patents [OSTI]

    Conway, Lawrence E. (Robinson Township, Allegheny County, PA); Stewart, William A. (Penn Hills Township, Allegheny County, PA)

    1991-01-01

    A containment cooling system utilizes a naturally induced air flow and a gravity flow of water over the containment shell which encloses a reactor core to cool reactor core decay heat in two stages. When core decay heat is greatest, the water and air flow combine to provide adequate evaporative cooling as heat from within the containment is transferred to the water flowing over the same. The water is heated by heat transfer and then evaporated and removed by the air flow. After an initial period of about three to four days when core decay heat is greatest, air flow alone is sufficient to cool the containment.

  19. Cooling water distribution system

    DOE Patents [OSTI]

    Orr, Richard (Pittsburgh, PA)

    1994-01-01

    A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

  20. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    SciTech Connect (OSTI)

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  1. Thermal analysis for fuel handling system for sodium cooled reactor considering minor actinide-bearing metal fuel.

    SciTech Connect (OSTI)

    Chikazawa, Y.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The Advanced Burner Reactor (ABR) is one of the components of the Global Nuclear Energy Partnership (GNEP) used to close the fuel cycle. ABR is a sodium-cooled fast reactor that is used to consume transuranic elements resulting from the reprocessing of light water reactor spent nuclear fuel. ABR-1000 [1000 MW(thermal)] is a fast reactor concept created at Argonne National Laboratory to be used as a reference concept for various future trade-offs. ABR-1000 meets the GNEP goals although it uses what is considered base sodium fast reactor technology for its systems and components. One of the considerations of any fast reactor plant concept is the ability to perform fuel-handling operations with new and spent fast reactor fuel. The transmutation fuel proposed as the ABR fuel has a very little experience base, and thus, this paper investigates a fuel-handling concept and potential issues of handling fast reactor fuel containing minor actinides. In this study, two thermal analyses supporting a conceptual design study on the ABR-1000 fuel-handling system were carried out. One analysis investigated passive dry spent fuel storage, and the other analysis investigated a fresh fuel shipping cask. Passive dry storage can be made suitable for the ABR-1000 spent fuel storage with sodium-bonded metal fuel. The thermal analysis shows that spent fast reactor fuel with a decay heat of 2 kW or less can be stored passively in a helium atmosphere. The 2-kW value seems to be a reasonable and practical level, and a combination of reasonably-sized in-sodium storage followed by passive dry storage could be a candidate for spent fuel storage for the next-generation sodium-cooled reactor with sodium-bonded metal fuel. Requirements for the shipping casks for minor actinide-bearing fuel with a high decay heat level are also discussed in this paper. The shipping cask for fresh sodium-cooled-reactor fuel should be a dry type to reduce the reaction between residual moisture on fresh fuel and the sodium coolant. The cladding temperature requirement is maintained below the creep temperature limit to avoid any damage before core installation. The thermal analysis shows that a helium gas-filled cask can accommodate ABR-1000 fresh minor actinide-bearing fuel with 700-W decay heat. The above analysis results revealed the overall requirement for minor actinide-bearing metal fuel handling. The information is thought to be helpful in the design of the ABR-1000 and future sodium-cooled-reactor fuel-handling system.

  2. Analysis of fission product revaporization in a BWR reactor cooling system during a station blackout accident

    SciTech Connect (OSTI)

    Yang, J.W.; Schmidt, E.; Cazzoli, E.; Khatib-Rahbar, M.

    1988-01-01

    This report presents a preliminary analysis of fission product revaporization in the Reactor Cooling System (RCS) after the vessel failure. The station blackout transient for BWR Mark I Power Plant is considered. The TRAPMELT3 models of evaporization, chemisorption, and the decay heating of RCS structures and gases are adopted in the analysis. The RCS flow models based on the density-difference between the RCS and containment pedestal region are developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP is developed for the analysis. The REVAP is incorporated with the MARCH, TRAPMELT3 and NAUA codes of the Source Term Code Pack Package (STCP). The NAUA code is used to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors determining the magnitude of revaporization and subsequent release of the volatile fission product. 8 figs., 1 tab.

  3. Development of Materials for Supercritical-Water-Cooled Reactor

    Office of Energy Efficiency and Renewable Energy (EERE)

    Supercritical-Water-Cooled Reactor (SCWR) was selected as one of the promising candidates in Generation IV reactors for its prominent advantages; those are the high thermal efficiency, the system...

  4. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  5. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  6. Power conversion system design for supercritical carbon dioxide cooled indirect cycle nuclear reactors

    E-Print Network [OSTI]

    Gibbs, Jonathan Paul

    2008-01-01

    The supercritical carbon dioxide (S-CO?) cycle is a promising advanced power conversion cycle which couples nicely to many Generation IV nuclear reactors. This work investigates the power conversion system design and ...

  7. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    SciTech Connect (OSTI)

    Not Available

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  8. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    SciTech Connect (OSTI)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  9. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    L. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,R. J. Neuhold, Introductury Nuclear Reactor Dynamics. ANSL. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,

  10. Gas-cooled nuclear reactor

    DOE Patents [OSTI]

    Peinado, Charles O. (La Jolla, CA); Koutz, Stanley L. (San Diego, CA)

    1985-01-01

    A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

  11. Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

    SciTech Connect (OSTI)

    Marcille, T. F.; Poston, D. I.; Kapernick, R. J. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Dixon, D. D. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Fischer, G. A. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Doherty, S. P. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Engineering, Trinity College, Hartford, CT 06106 (United States)

    2006-01-20

    In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

  12. Stability analysis of supercritical water cooled reactors

    E-Print Network [OSTI]

    Zhao, Jiyun, Ph. D. Massachusetts Institute of Technology

    2005-01-01

    The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500°C average core exit). The high coolant temperature as it leaves the ...

  13. A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

    E-Print Network [OSTI]

    Billings, Jay Jay; Hull, S Forest; Lingerfelt, Eric J; Wojtowicz, Anna

    2014-01-01

    Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately available in the community and need to be developed. This paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It describes requirements to build such a system and the technical challenges faced...

  14. Risk-informed design guidance for a Generation-IV gas-cooled fast reactor emergency core cooling system

    E-Print Network [OSTI]

    Delaney, Michael J. (Michael James), 1979-

    2004-01-01

    Fundamental objectives of sustainability, economics, safety and reliability, and proliferation resistance, physical protection and stakeholder relations must be considered during the design of an advanced reactor. However, ...

  15. Emergency core cooling system

    DOE Patents [OSTI]

    Schenewerk, William E. (Sherman Oaks, CA); Glasgow, Lyle E. (Westlake Village, CA)

    1983-01-01

    A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.

  16. Passive containment cooling system

    DOE Patents [OSTI]

    Billig, P.F.; Cooke, F.E.; Fitch, J.R.

    1994-01-25

    A passive containment cooling system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel and is vented to the drywell. An isolation pool is disposed above the GDCS pool and includes an isolation condenser therein. The condenser has an inlet line disposed in flow communication with the drywell for receiving the non-condensable gas along with any steam released therein following a loss-of-coolant accident (LOCA). The condenser also has an outlet line disposed in flow communication with the drywell for returning to the drywell both liquid condensate produced upon cooling of the steam and the non-condensable gas for reducing pressure within the containment vessel following the LOCA. 1 figure.

  17. Passive containment cooling system

    DOE Patents [OSTI]

    Billig, Paul F. (San Jose, CA); Cooke, Franklin E. (San Jose, CA); Fitch, James R. (San Jose, CA)

    1994-01-01

    A passive containment cooling system includes a containment vessel surrounding a reactor pressure vessel and defining a drywell therein containing a non-condensable gas. An enclosed wetwell pool is disposed inside the containment vessel, and a gravity driven cooling system (GDCS) pool is disposed above the wetwell pool in the containment vessel and is vented to the drywell. An isolation pool is disposed above the GDCS pool and includes an isolation condenser therein. The condenser has an inlet line disposed in flow communication with the drywell for receiving the non-condensable gas along with any steam released therein following a loss-of-coolant accident (LOCA). The condenser also has an outlet line disposed in flow communication with the drywell for returning to the drywell both liquid condensate produced upon cooling of the steam and the non-condensable gas for reducing pressure within the containment vessel following the LOCA.

  18. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    high temperature reactors (VHTR), 3) supercritical waterup to 760 ° C. The DOE VHTR program is also sponsoring a

  19. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, R.M.

    1983-11-08

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  20. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, Robert M. (Macungie, PA)

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  1. Alternative cooling resource for removing the residual heat of reactor

    SciTech Connect (OSTI)

    Park, H. C.; Lee, J. H.; Lee, D. S.; Jung, C. Y.; Choi, K. Y. [Korea Hydro and Nuclear Power Co., Ltd., 260 Naa-ri Yangnam-myeon Gyeongju-si, Gyeonasangbuk-do, 780-815 (Korea, Republic of)

    2012-07-01

    The Recirculated Cooling Water (RCW) system of a Candu reactor is a closed cooling system which delivers demineralized water to coolers and components in the Service Building, the Reactor Building, and the Turbine Building and the recirculated cooling water is designed to be cooled by the Raw Service Water (RSW). During the period of scheduled outage, the RCW system provides cooling water to the heat exchangers of the Shutdown Cooling System (SDCS) in order to remove the residual heat of the reactor, so the RCW heat exchangers have to operate at all times. This makes it very hard to replace the inlet and outlet valves of the RCW heat exchangers because the replacement work requires the isolation of the RCW. A task force was formed to prepare a plan to substitute the recirculated water with the chilled water system in order to cool the SDCS heat exchangers. A verification test conducted in 2007 proved that alternative cooling was possible for the removal of the residual heat of the reactor and in 2008 the replacement of inlet and outlet valves of the RCW heat exchangers for both Wolsong unit 3 and 4 were successfully completed. (authors)

  2. Reactor physics design of supercritical CO?-cooled fast reactors

    E-Print Network [OSTI]

    Pope, Michael A. (Michael Alexander)

    2004-01-01

    Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO? (S-CO?) as a Brayton cycle working fluid in a direct ...

  3. Experimental Study of the Thermal-Hydraulic Phenomena in the Reactor Cavity Cooling System and Analysis of the Effects of Graphite Dispersion 

    E-Print Network [OSTI]

    Vaghetto, Rodolfo

    2012-07-16

    An experimental activity was performed to observe and study the effects of graphite dispersion and deposition on thermal hydraulic phenomena in a Reactor Cavity Cooling System (RCCS). The small scale RCCS experimental facility (16.5cm x 16.5cm x 30...

  4. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    SciTech Connect (OSTI)

    Corradin, Michael; Anderson, M.; Muci, M.; Hassan, Yassin; Dominguez, A.; Tokuhiro, Akira; Hamman, K.

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  5. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    electric heaters, jet pumps, turbines, separators, annuli, pressurizers, accumulators, and power control system components. In addition, special process models

  6. Radiant vessel auxiliary cooling system

    DOE Patents [OSTI]

    Germer, John H. (San Jose, CA)

    1987-01-01

    In a modular liquid-metal pool breeder reactor, a radiant vessel auxiliary cooling system is disclosed for removing the residual heat resulting from the shutdown of a reactor by a completely passive heat transfer system. A shell surrounds the reactor and containment vessel, separated from the containment vessel by an air passage. Natural circulation of air is provided by air vents at the lower and upper ends of the shell. Longitudinal, radial and inwardly extending fins extend from the shell into the air passage. The fins are heated by radiation from the containment vessel and convect the heat to the circulating air. Residual heat from the primary reactor vessel is transmitted from the reactor vessel through an inert gas plenum to a guard or containment vessel designed to contain any leaking coolant. The containment vessel is conventional and is surrounded by the shell.

  7. Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors

    E-Print Network [OSTI]

    Zweibaum, Nicolas

    2015-01-01

    Systems for Advanced Nuclear Reactors,” Ph.D. Dissertation,Handbook for Nuclear Reactor Safety,” Luxembourg: Commissiondevelopment of advanced nuclear reactor technology requires

  8. Reliability analysis of a passive cooling system using a response surface with an application to the Flexible Conversion Ratio Reactor

    E-Print Network [OSTI]

    Fong, Christopher J. (Christopher Joseph)

    2008-01-01

    A comprehensive risk-informed methodology for passive safety system design and performance assessment is presented and demonstrated on the Flexible Conversion Ratio Reactor (FCRR). First, the methodology provides a framework ...

  9. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    Potential Safety Issues – Regulatory Design Criteria3-3 Regulatory design criteria for safety Table 3-4 Input3-4 Regulatory Design Criteria for safety The DRACS system

  10. Cooling Water System Optimization 

    E-Print Network [OSTI]

    Aegerter, R.

    2005-01-01

    During summer months, many manufacturing plants have to cut back in rates because the cooling water system is not providing sufficient cooling to support higher production rates. There are many low/no-cost techniques available to improve tower...

  11. Data center cooling system

    DOE Patents [OSTI]

    Chainer, Timothy J; Dang, Hien P; Parida, Pritish R; Schultz, Mark D; Sharma, Arun

    2015-03-17

    A data center cooling system may include heat transfer equipment to cool a liquid coolant without vapor compression refrigeration, and the liquid coolant is used on a liquid cooled information technology equipment rack housed in the data center. The system may also include a controller-apparatus to regulate the liquid coolant flow to the liquid cooled information technology equipment rack through a range of liquid coolant flow values based upon information technology equipment temperature thresholds.

  12. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    SciTech Connect (OSTI)

    David E. Shropshire

    2004-04-01

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

  13. Gas turbine cooling system

    DOE Patents [OSTI]

    Bancalari, Eduardo E. (Orlando, FL)

    2001-01-01

    A gas turbine engine (10) having a closed-loop cooling circuit (39) for transferring heat from the hot turbine section (16) to the compressed air (24) produced by the compressor section (12). The closed-loop cooling system (39) includes a heat exchanger (40) disposed in the flow path of the compressed air (24) between the outlet of the compressor section (12) and the inlet of the combustor (14). A cooling fluid (50) may be driven by a pump (52) located outside of the engine casing (53) or a pump (54) mounted on the rotor shaft (17). The cooling circuit (39) may include an orifice (60) for causing the cooling fluid (50) to change from a liquid state to a gaseous state, thereby increasing the heat transfer capacity of the cooling circuit (39).

  14. Rotary engine cooling system

    SciTech Connect (OSTI)

    Jones, C.

    1988-07-26

    A rotary internal combustion engine is described comprising: a rotor housing forming a trochoidal cavity therein; an insert of refractory material received in the recess, an element of a fuel injection and ignition system extending through the housing and insert bores, and the housing having cooling passages extending therethrough. The cooling passages are comprised of drilled holes.

  15. Startup of the FFTF sodium cooled reactor. [Acceptance Test Program

    SciTech Connect (OSTI)

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed.

  16. Candidate Materials Evaluation for Supercritical Water-Cooled Reactor

    SciTech Connect (OSTI)

    T. R. Allen and G. S. Was

    2008-12-12

    Final technical report on the corrosion, stress corrosion cracking, and radiation response of candidate materials for the supercritical water-cooled reactor concept.

  17. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01

    safety design criteria separate effects test steam generators small modular reactor San Onofre Nuclear

  18. Hydronic rooftop cooling systems

    DOE Patents [OSTI]

    Bourne, Richard C. (Davis, CA); Lee, Brian Eric (Monterey, CA); Berman, Mark J. (Davis, CA)

    2008-01-29

    A roof top cooling unit has an evaporative cooling section that includes at least one evaporative module that pre-cools ventilation air and water; a condenser; a water reservoir and pump that captures and re-circulates water within the evaporative modules; a fan that exhausts air from the building and the evaporative modules and systems that refill and drain the water reservoir. The cooling unit also has a refrigerant section that includes a compressor, an expansion device, evaporator and condenser heat exchangers, and connecting refrigerant piping. Supply air components include a blower, an air filter, a cooling and/or heating coil to condition air for supply to the building, and optional dampers that, in designs that supply less than 100% outdoor air to the building, control the mixture of return and ventilation air.

  19. Superconductor rotor cooling system

    DOE Patents [OSTI]

    Gamble, Bruce B.; Sidi-Yekhlef, Ahmed; Schwall, Robert E.; Driscoll, David I.; Shoykhet, Boris A.

    2004-11-02

    A system for cooling a superconductor device includes a cryocooler located in a stationary reference frame and a closed circulation system external to the cryocooler. The closed circulation system interfaces the stationary reference frame with a rotating reference frame in which the superconductor device is located. A method of cooling a superconductor device includes locating a cryocooler in a stationary reference frame, and transferring heat from a superconductor device located in a rotating reference frame to the cryocooler through a closed circulation system external to the cryocooler. The closed circulation system interfaces the stationary reference frame with the rotating reference frame.

  20. Superconductor rotor cooling system

    DOE Patents [OSTI]

    Gamble, Bruce B. (Wellesley, MA); Sidi-Yekhlef, Ahmed (Framingham, MA); Schwall, Robert E. (Northborough, MA); Driscoll, David I. (South Euclid, OH); Shoykhet, Boris A. (Beachwood, OH)

    2002-01-01

    A system for cooling a superconductor device includes a cryocooler located in a stationary reference frame and a closed circulation system external to the cryocooler. The closed circulation system interfaces the stationary reference frame with a rotating reference frame in which the superconductor device is located. A method of cooling a superconductor device includes locating a cryocooler in a stationary reference frame, and transferring heat from a superconductor device located in a rotating reference frame to the cryocooler through a closed circulation system external to the cryocooler. The closed circulation system interfaces the stationary reference frame with the rotating reference frame.

  1. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect (OSTI)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  2. Relap5-3d model validation and benchmark exercises for advanced gas cooled reactor application 

    E-Print Network [OSTI]

    Moore, Eugene James Thomas

    2006-08-16

    HTTR High Temperature engineering Test Reactor INET Institute of Nuclear Energy Technology LWR Light Water Reactor OKBM Test Design Bureau for Machine Building ORNL Oak Ridge National Laboratory RCCS Reactor Cavity Cooling System.... This code used a card-based input deck and was designed to simulate transients in light water reactors (LWR) including, but not limited to, a loss of coolant accident or a station blackout. In the thirty years since, the code has been continuously...

  3. Cooling molten salt reactors using “gas-lift”

    SciTech Connect (OSTI)

    Zitek, Pavel E-mail: klimko@kke.zcu.cz; Valenta, Vaclav E-mail: klimko@kke.zcu.cz; Klimko, Marek E-mail: klimko@kke.zcu.cz

    2014-08-06

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a “Two-phase flow demonstrator” (TFD) used for experimental study of the “gas-lift” system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for “gas-lift” (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

  4. Ground Water Cooling System 

    E-Print Network [OSTI]

    Greaves, K.; Chave, G. H.

    1984-01-01

    has a total shop area of 128,000 square feet and the majority of the machine tools are equipped with computerized numerical controls. The cooling system was designed around five (5) floor mounted, 50,000 CFM, air handling units which had been...

  5. Diesel lubrication and cooling systems

    SciTech Connect (OSTI)

    NONE

    1994-12-31

    The film describes the parts of diesel lubricating and cooling systems and how they work in relation to each other.

  6. Diesel lubrication and cooling systems

    SciTech Connect (OSTI)

    Not Available

    1994-01-01

    The film describes the parts of diesel lubricating and cooling systems and how they work in relation to each other.

  7. CONTROL SYSTEM FOR SOLAR HEATING and COOLING

    E-Print Network [OSTI]

    Dols, C.

    2010-01-01

    operating solar heating and cooling systems covering a widepractical heating and cooling system configurations andexperimental heating and cooling system, the main purpose of

  8. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  9. Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor

    SciTech Connect (OSTI)

    Cheng, Lap Y.; Ludewig, Hans; Jo, Jae [Brookhaven National Laboratory, P.O. Box 5000, Upton, NY 11973-5000 (United States)

    2006-07-01

    A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400 MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645 m{sup 2}) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800 kPa. (authors)

  10. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  11. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01

    reprocessing to recover fissionable material, FHR fuel handling systems must be designed to facilitate the application of IAEA safeguards.

  12. Fuel Development For Gas-Cooled Fast Reactors

    SciTech Connect (OSTI)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  13. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    SciTech Connect (OSTI)

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  14. A Semi-Passive Containment Cooling System Conceptual Design

    E-Print Network [OSTI]

    Liu, H.

    The objective of this project was to investigate a passive containment cooling system (PCCS) for the double concrete containment of the Korean Next Generation Reactor (KNGR). Two conceptual PCCS designs: the thermosyphon ...

  15. TRITIUM PERMEATION AND TRANSPORT IN THE GASOLINE PRODUCTION SYSTEM COUPLED WITH HIGH TEMPERATURE GAS-COOLED REACTORS (HTGRS)

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2011-05-01

    This paper describes scoping analyses on tritium behaviors in the HTGR-integrated gasoline production system, which is based on a methanol-to-gasoline (MTG) plant. In this system, the HTGR transfers heat and electricity to the MTG system. This system was analyzed using the TPAC code, which was recently developed by Idaho National Laboratory. The global sensitivity analyses were performed to understand and characterize tritium behaviors in the coupled HTGR/MTG system. This Monte Carlo based random sampling method was used to evaluate maximum 17,408 numbers of samples with different input values. According to the analyses, the average tritium concentration in the product gasoline is about 3.05×10-3 Bq/cm3, and 62 % cases are within the tritium effluent limit (= 3.7x10-3 Bq/cm3[STP]). About 0.19% of released tritium is finally transported from the core to the gasoline product through permeations. This study also identified that the following four parameters are important concerning tritium behaviors in the HTGR/MTG system: (1) tritium source, (2) wall thickness of process heat exchanger, (3) operating temperature, and (4) tritium permeation coefficient of process heat exchanger. These four parameters contribute about 95 % of the total output uncertainties. This study strongly recommends focusing our future research on these four parameters to improve modeling accuracy and to mitigate tritium permeation into the gasol ine product. If the permeation barrier is included in the future study, the tritium concentration will be significantly reduced.

  16. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  17. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1995-01-01

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

  18. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOE Patents [OSTI]

    Hunsbedt, A.; Boardman, C.E.

    1995-04-11

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

  19. Liquid metal reactor air cooling baffle

    DOE Patents [OSTI]

    Hunsbedt, A.

    1994-08-16

    A baffle is provided between a relatively hot containment vessel and a relatively cold silo for enhancing air cooling performance. The baffle includes a perforate inner wall positionable outside the containment vessel to define an inner flow riser therebetween, and an imperforate outer wall positionable outside the inner wall to define an outer flow riser therebetween. Apertures in the inner wall allow thermal radiation to pass laterally therethrough to the outer wall, with cooling air flowing upwardly through the inner and outer risers for removing heat. 3 figs.

  20. Liquid metal reactor air cooling baffle

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA)

    1994-01-01

    A baffle is provided between a relatively hot containment vessel and a relatively cold silo for enhancing air cooling performance. The baffle includes a perforate inner wall positionable outside the containment vessel to define an inner flow riser therebetween, and an imperforate outer wall positionable outside the inner wall to define an outer flow riser therebetween. Apertures in the inner wall allow thermal radiation to pass laterally therethrough to the outer wall, with cooling air flowing upwardly through the inner and outer risers for removing heat.

  1. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative) applications

    SciTech Connect (OSTI)

    Marshall, A.C.

    1989-08-01

    We have completed a review of multimegawatt gas-cooled reactor concepts proposed for SDI applications. Our study concluded that the principal reason for considering gas-cooled reactors for burst-mode operation was the potential for significant system mass savings over closed-cycle systems if open-cycle gas-cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas-cooled reactors for steady-state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas-cooled reactor concepts were compared to identify the most promising. For burst-mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology has been demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady-State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode. 90 refs., 2 figs., 10 tabs.

  2. Process Cooling Systems 

    E-Print Network [OSTI]

    McCann, C. J.

    1983-01-01

    Cooling towers have been on the scene for more than 50 years. It is because they have proven to be an economic choice for waste heat dissipation. But it seems, for some reason, that after installation very little attention is paid to the cooling...

  3. Emergency cooling system and method

    DOE Patents [OSTI]

    Oosterkamp, W.J.; Cheung, Y.K.

    1994-01-04

    An improved emergency cooling system and method are disclosed that may be adapted for incorporation into or use with a nuclear BWR wherein a reactor pressure vessel (RPV) containing a nuclear core and a heat transfer fluid for circulation in a heat transfer relationship with the core is housed within an annular sealed drywell and is fluid communicable therewith for passage thereto in an emergency situation the heat transfer fluid in a gaseous phase and any noncondensibles present in the RPV, an annular sealed wetwell houses the drywell, and a pressure suppression pool of liquid is disposed in the wetwell and is connected to the drywell by submerged vents. The improved emergency cooling system and method has a containment condenser for receiving condensible heat transfer fluid in a gaseous phase and noncondensibles for condensing at least a portion of the heat transfer fluid. The containment condenser has an inlet in fluid communication with the drywell for receiving heat transfer fluid and noncondensibles, a first outlet in fluid communication with the RPV for the return to the RPV of the condensed portion of the heat transfer fluid and a second outlet in fluid communication with the drywell for passage of the noncondensed balance of the heat transfer fluid and the noncondensibles. The noncondensed balance of the heat transfer fluid and the noncondensibles passed to the drywell from the containment condenser are mixed with the heat transfer fluid and the noncondensibles from the RPV for passage into the containment condenser. A water pool is provided in heat transfer relationship with the containment condenser and is thermally communicable in an emergency situation with an environment outside of the drywell and the wetwell for conducting heat transferred from the containment condenser away from the wetwell and the drywell. 5 figs.

  4. Heat pipe cooled reactors for multi-kilowatt space power supplies

    SciTech Connect (OSTI)

    Ranken, W.A.; Houts, M.G.

    1995-01-01

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFEs) to radiator heat pipes.

  5. Experimental Study on Flow Optimization in Upper Plenum of Reactor Vessel for a Compact Sodium-Cooled Fast Reactor

    SciTech Connect (OSTI)

    Kimura, Nobuyuki; Hayashi, Kenji; Kamide, Hideki; Itoh, Masami; Sekine, Tadashi

    2005-11-15

    An innovative sodium-cooled fast reactor has been investigated in a feasibility study of fast breeder reactor cycle systems in Japan. A compact reactor vessel and a column-type upper inner structure with a radial slit for an arm of a fuel-handling machine (FHM) are adopted. Dipped plates are set in the reactor vessel below the free surface to prevent gas entrainment. We performed a one-tenth-scaled model water experiment for the upper plenum of the reactor vessel. Gas entrainment was not observed in the experiment under the same velocity condition as the reactor. Three vortex cavitations were observed near the hot-leg inlet. A vertical rib on the reactor vessel wall was set to restrict the rotating flow near the hot leg. The vortex cavitation between the reactor vessel wall and the hot leg was suppressed by the rib under the same cavitation factor condition as in the reactor. The cylindrical plug was installed through the hole in the dipped plates for the FHM to reduce the flow toward the free surface. It was effective when the plug was submerged into the middle height in the upper plenum. This combination of two components had a possibility to optimize the flow in the compact reactor vessel.

  6. Desiccant Cooling Systems - A Review 

    E-Print Network [OSTI]

    Kettleborough, C. F.; Ullah, M. R.; Waugaman, D. G.

    1986-01-01

    Desiccant cooling systems have been investigated extensively during the past decade as alternatives to electrically driven vapor compression systems because regeneration temperatures of the desiccant - about 160°F, can be achieved using natural gas...

  7. Temperature initiated passive cooling system

    DOE Patents [OSTI]

    Forsberg, C.W.

    1994-11-01

    A passive cooling system for cooling an enclosure only when the enclosure temperature exceeds a maximum standby temperature comprises a passive heat transfer loop containing heat transfer fluid having a particular thermodynamic critical point temperature just above the maximum standby temperature. An upper portion of the heat transfer loop is insulated to prevent two phase operation below the maximum standby temperature. 1 fig.

  8. Temperature initiated passive cooling system

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1994-01-01

    A passive cooling system for cooling an enclosure only when the enclosure temperature exceeds a maximum standby temperature comprises a passive heat transfer loop containing heat transfer fluid having a particular thermodynamic critical point temperature just above the maximum standby temperature. An upper portion of the heat transfer loop is insulated to prevent two phase operation below the maximum standby temperature.

  9. Combustor liner cooling system

    DOE Patents [OSTI]

    Lacy, Benjamin Paul; Berkman, Mert Enis

    2013-08-06

    A combustor liner is disclosed. The combustor liner includes an upstream portion, a downstream end portion extending from the upstream portion along a generally longitudinal axis, and a cover layer associated with an inner surface of the downstream end portion. The downstream end portion includes the inner surface and an outer surface, the inner surface defining a plurality of microchannels. The downstream end portion further defines a plurality of passages extending between the inner surface and the outer surface. The plurality of microchannels are fluidly connected to the plurality of passages, and are configured to flow a cooling medium therethrough, cooling the combustor liner.

  10. Non-intrusive cooling system

    DOE Patents [OSTI]

    Morrison, Edward F. (Burnt Hills, NY); Bergman, John W. (Barrington, NH)

    2001-05-22

    A readily replaceable heat exchange cooling jacket for applying fluid to a system conduit pipe. The cooling jacket comprises at least two members, separable into upper and lower portions. A chamber is formed between the conduit pipe and cooling jacket once the members are positioned about the pipe. The upper portion includes a fluid spray means positioned above the pipe and the bottom portion includes a fluid removal means. The heat exchange cooling jacket is adaptable with a drain tank, a heat exchanger, a pump and other standard equipment to provide a system for removing heat from a pipe. A method to remove heat from a pipe, includes the steps of enclosing a portion of the pipe with a jacket to form a chamber between an outside surface of the pipe and the cooling jacket; spraying cooling fluid at low pressure from an upper portion of the cooling jacket, allowing the fluid to flow downwardly by gravity along the surface of the pipe toward a bottom portion of the chamber; and removing the fluid at the bottom portion of the chamber.

  11. Core design and reactor physics of a breed and burn gas-cooled fast reactor

    E-Print Network [OSTI]

    Yarsky, Peter

    2005-01-01

    In order to fulfill the goals set forth by the Generation IV International Forum, the current NERI funded research has focused on the design of a Gas-cooled Fast Reactor (GFR) operating in a Breed and Burnm (B&B) fuel cycle ...

  12. Compressor bleed cooling fluid feed system

    DOE Patents [OSTI]

    Donahoo, Eric E; Ross, Christopher W

    2014-11-25

    A compressor bleed cooling fluid feed system for a turbine engine for directing cooling fluids from a compressor to a turbine airfoil cooling system to supply cooling fluids to one or more airfoils of a rotor assembly is disclosed. The compressor bleed cooling fluid feed system may enable cooling fluids to be exhausted from a compressor exhaust plenum through a downstream compressor bleed collection chamber and into the turbine airfoil cooling system. As such, the suction created in the compressor exhaust plenum mitigates boundary layer growth along the inner surface while providing flow of cooling fluids to the turbine airfoils.

  13. Hydraulic tests of emergency cooling system: L-Area

    SciTech Connect (OSTI)

    Hinton, J H

    1988-01-01

    The delay in L-Area startup provided an opportunity to obtain valuable data on the Emergency Cooling System (ECS) which will permit reactor operation at the highest safe power level. ECS flow is a major input to the FLOOD code which calculates reactor ECS power limits. The FLOOD code assesses the effectiveness of the ECS cooling capacity by modeling the core and plenum hydraulics under accident conditions. Presently, reactor power is not limited by the ECS cooling capacity (power limit). However, the manual calculations of ECS flows had been recently updated to include piping changes (debris strainer, valve changes, pressure release systems) and update fitting losses. Both updates resulted in reduced calculated ECS flows. Upon completion of the current program to update, validate, and document, reactor power may be limited under certain situations by ECS cooling capacity for some present reactor charge designs. A series of special hydraulic tests (Reference 1, 3) were conducted in L-Area using all sources of emergency coolant including the ECS pumps (Reference 2). The tests provided empirical hydraulic data on the ECS piping. These data will be used in computer models of the system as well as manual calculations of ECS flows. The improved modeling and accuracy of the flow calculations will permit reactor operation at the highest safe power level with respect to an ECS power limit.

  14. Solar-powered cooling system

    DOE Patents [OSTI]

    Farmer, Joseph C

    2013-12-24

    A solar-powered adsorption-desorption refrigeration and air conditioning system uses nanostructural materials made of high specific surface area adsorption aerogel as the adsorptive media. Refrigerant molecules are adsorbed on the high surface area of the nanostructural material. A circulation system circulates refrigerant from the nanostructural material to a cooling unit.

  15. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Cheng, Lap-Yan; Wei, Thomas Y. C.

    2009-01-01

    The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore »evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

  16. Alternative Passive Decay-Heat Systems for the Advanced High-Temperature Reactor

    SciTech Connect (OSTI)

    Forsberg, Charles W.

    2006-07-01

    The Advanced High-Temperature Reactor (AHTR) is a low-pressure, liquid-salt-cooled high-temperature reactor for the production of electricity and hydrogen. The high-temperature (950 deg C) variant is defined as the liquid-salt-cooled very high-temperature reactor (LS-VHTR). The AHTR has the same safety goals and uses the same graphite-matrix coated particle fuel as do modular high-temperature gas-cooled reactors. However, the large AHTR power output [2400 to 4000 MW(t)] implies the need for a different type of passive decay-heat removal system. Because the AHTR is a low-pressure, liquid-cooled reactor like sodium-cooled reactors, similar types of decay-heat-removal systems can be used. Three classes of passive decay heat removal systems have been identified: the reactor vessel auxiliary cooling system which is similar to that proposed for the General Electric S-PRISM sodium-cooled fast reactor; the direct reactor auxiliary cooling system, which is similar to that used in the Experimental Breeder Reactor-II; and a new pool reactor auxiliary cooling system. These options are described and compared. (author)

  17. Cooling system for electronic components

    DOE Patents [OSTI]

    Anderl, William James; Colgan, Evan George; Gerken, James Dorance; Marroquin, Christopher Michael; Tian, Shurong

    2015-12-15

    Embodiments of the present invention provide for non interruptive fluid cooling of an electronic enclosure. One or more electronic component packages may be removable from a circuit card having a fluid flow system. When installed, the electronic component packages are coincident to and in a thermal relationship with the fluid flow system. If a particular electronic component package becomes non-functional, it may be removed from the electronic enclosure without affecting either the fluid flow system or other neighboring electronic component packages.

  18. Risk-informed design changes for a passive cooling system

    E-Print Network [OSTI]

    Patalano, Giovanbattista

    2007-01-01

    The failure probability of a passive decay heat removal system after a LOCA is evaluated as part of a risk-informed design process for a helium-cooled fast reactor. The system was modeled using RELAP5-3D. The epistemic ...

  19. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    SciTech Connect (OSTI)

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  20. Lamination cooling system formation method

    DOE Patents [OSTI]

    Rippel, Wally E. (Altadena, CA); Kobayashi, Daryl M. (Monrovia, CA)

    2012-06-19

    An electric motor, transformer or inductor having a cooling system. A stack of laminations have apertures at least partially coincident with apertures of adjacent laminations. The apertures define straight or angled cooling-fluid passageways through the lamination stack. Gaps between the adjacent laminations are sealed by injecting a heat-cured sealant into the passageways, expelling excess sealant, and heat-curing the lamination stack. Manifold members adjoin opposite ends of the lamination stack, and each is configured with one or more cavities to act as a manifold to adjacent passageway ends. Complex manifold arrangements can create bidirectional flow in a variety of patterns.

  1. Lamination cooling system formation method

    DOE Patents [OSTI]

    Rippel, Wally E [Altadena, CA; Kobayashi, Daryl M [Monrovia, CA

    2009-05-12

    An electric motor, transformer or inductor having a cooling system. A stack of laminations have apertures at least partially coincident with apertures of adjacent laminations. The apertures define straight or angled cooling-fluid passageways through the lamination stack. Gaps between the adjacent laminations are sealed by injecting a heat-cured sealant into the passageways, expelling excess sealant, and heat-curing the lamination stack. Manifold members adjoin opposite ends of the lamination stack, and each is configured with one or more cavities to act as a manifold to adjacent passageway ends. Complex manifold arrangements can create bidirectional flow in a variety of patterns.

  2. Optimized core design of a supercritical carbon dioxide-cooled fast reactor

    E-Print Network [OSTI]

    Handwerk, Christopher S. (Christopher Stanley), 1974-

    2007-01-01

    Spurred by the renewed interest in nuclear power, Gas-cooled Fast Reactors (GFRs) have received increasing attention in the past decade. Motivated by the goals of the Generation-IV International Forum (GIF), a GFR cooled ...

  3. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    SciTech Connect (OSTI)

    Lee O. Nelson

    2011-04-01

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540°C and the helium coolant was delivered at 7 MPa at 625–925°C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the hightemperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

  4. Thermal hydraulic design of a salt-cooled highly efficient environmentally friendly reactor

    E-Print Network [OSTI]

    Whitman, Joshua (Joshua J.)

    2009-01-01

    A 1 OOOMWth liquid-salt cooled thermal spectrum reactor was designed with a long fuel cycle, and high core exit temperature. These features are desirable in a reactor designed to provide process heat applications such as ...

  5. Gas hydrate cool storage system

    DOE Patents [OSTI]

    Ternes, M.P.; Kedl, R.J.

    1984-09-12

    The invention presented relates to the development of a process utilizing a gas hydrate as a cool storage medium for alleviating electric load demands during peak usage periods. Several objectives of the invention are mentioned concerning the formation of the gas hydrate as storage material in a thermal energy storage system within a heat pump cycle system. The gas hydrate was formed using a refrigerant in water and an example with R-12 refrigerant is included. (BCS)

  6. An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

    SciTech Connect (OSTI)

    Yoder Jr, Graydon L; Aaron, Adam M; Cunningham, Richard Burns; Fugate, David L; Holcomb, David Eugene; Kisner, Roger A; Peretz, Fred J; Robb, Kevin R; Wilgen, John B; Wilson, Dane F

    2014-01-01

    The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

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

    SciTech Connect (OSTI)

    Yang, J.; Dizon, M.B.; Cheung, F.B. [Department of Mechanical and Nuclear Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Rempe, J.L. [Idaho National Engineering and Environmental Laboratory, P.O. Box 1625, Idaho Falls, ID (United States); Suh, K.Y. [Department of Nuclear Engineering, Seoul National University, Seoul (Korea, Republic of); Kim, S.B. [Korea Atomic Energy Research Institute, P.O. Box 105, Yuseoung, Taejon (Korea, Republic of)

    2004-07-01

    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 (Sub-scale 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. (authors)

  8. 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.

  9. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    SciTech Connect (OSTI)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  10. Reactor protection system design alternatives for sodium fast reactors

    E-Print Network [OSTI]

    DeWitte, Jacob D. (Jacob Dominic)

    2011-01-01

    Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a ...

  11. Design Windows for a He Cooled Fusion Reactor* Dai-Kai Sze and Ahmed Hassanein

    E-Print Network [OSTI]

    Harilal, S. S.

    Design Windows for a He Cooled Fusion Reactor* Dai-Kai Sze and Ahmed Hassanein Argonne National Laboratory 9700 South Cass Avenue, Argonne, IL 60439 EQUATIONDERIVATION ABSTRACT A design window concept is developed for a He-cooled fusion reactor blanket and divertor design. This concept allows study

  12. Cooling system for superconducting magnet

    DOE Patents [OSTI]

    Gamble, Bruce B. (Wellesley, MA); Sidi-Yekhlef, Ahmed (Framingham, MA)

    1998-01-01

    A cooling system is configured to control the flow of a refrigerant by controlling the rate at which the refrigerant is heated, thereby providing an efficient and reliable approach to cooling a load (e.g., magnets, rotors). The cooling system includes a conduit circuit connected to the load and within which a refrigerant circulates; a heat exchanger, connected within the conduit circuit and disposed remotely from the load; a first and a second reservoir, each connected within the conduit, each holding at least a portion of the refrigerant; a heater configured to independently heat the first and second reservoirs. In a first mode, the heater heats the first reservoir, thereby causing the refrigerant to flow from the first reservoir through the load and heat exchanger, via the conduit circuit and into the second reservoir. In a second mode, the heater heats the second reservoir to cause the refrigerant to flow from the second reservoir through the load and heat exchanger via the conduit circuit and into the first reservoir.

  13. Cooling system for superconducting magnet

    DOE Patents [OSTI]

    Gamble, B.B.; Sidi-Yekhlef, A.

    1998-12-15

    A cooling system is configured to control the flow of a refrigerant by controlling the rate at which the refrigerant is heated, thereby providing an efficient and reliable approach to cooling a load (e.g., magnets, rotors). The cooling system includes a conduit circuit connected to the load and within which a refrigerant circulates; a heat exchanger, connected within the conduit circuit and disposed remotely from the load; a first and a second reservoir, each connected within the conduit, each holding at least a portion of the refrigerant; a heater configured to independently heat the first and second reservoirs. In a first mode, the heater heats the first reservoir, thereby causing the refrigerant to flow from the first reservoir through the load and heat exchanger, via the conduit circuit and into the second reservoir. In a second mode, the heater heats the second reservoir to cause the refrigerant to flow from the second reservoir through the load and heat exchanger via the conduit circuit and into the first reservoir. 3 figs.

  14. SCALE Code Validation for Prismatic High-Temperature Gas-Cooled Reactors

    SciTech Connect (OSTI)

    Ilas, Dan [ORNL

    2013-01-01

    Using experimental data published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments for the fresh cold core of the High Temperature Engineering Test Reactor, a comprehensive validation study has been carried out to assess the performance of the SCALE code system for analysis of high-temperature gas-cooled reactor configurations. This paper describes part of the results of that effort. The studies performed included criticality evaluations for the full core and for the annular cores realized during the fuel loading, as well as calculations and comparisons for excess reactivity, shutdown margin, control rod worths, temperature coefficient of reactivity, and reaction rate distributions. Comparisons of the SCALE results with both experimental values and MCNP-calculated values are presented. The comparisons show that the SCALE calculated results, obtained with both multigroup and continuous energy cross sections, are in reasonable agreement with the experimental data and the MCNP predictions.

  15. SCALE Code Validation for Prismatic High-Temperature Gas-Cooled Reactors

    SciTech Connect (OSTI)

    Ilas, Dan [ORNL] [ORNL

    2012-01-01

    Using experimental data published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments for the fresh cold core of the High Temperature Engineering Test Reactor, a comprehensive validation study has been carried out to assess the performance of the SCALE code system for analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. This paper describes part of the results of this effort. The studies performed included criticality evaluations for the full core and for the annular cores realized during the fuel loading, as well as calculations and comparisons for excess reactivity, shutdown margin, control rod worths, temperature coefficient of reactivity, and reaction rate distributions. Comparisons of the SCALE results with both the experimental values and MCNP-calculated values are presented. The comparisons show that the SCALE calculated results, obtained with both multigroup and continuous energy cross sections, are in reasonable agreement with the experimental data. The agreement with the MCNP predictions is, in general, very good.

  16. Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

    SciTech Connect (OSTI)

    Holcomb, David Eugene; Flanagan, George F; Mays, Gary T; Pointer, William David; Robb, Kevin R; Yoder Jr, Graydon L

    2013-11-01

    Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

  17. Cooling load differences between radiant and air systems

    E-Print Network [OSTI]

    Feng, Jingjuan Dove; Schiavon, Stefano; Bauman, Fred

    2013-01-01

    radiant heating and cooling systems, in: Proceedings ofSizing of the radiant system cooling equipment is highlycooling rate and air system cooling rate in this section. To

  18. Design and Control of Hydronic Radiant Cooling Systems

    E-Print Network [OSTI]

    Feng, Jingjuan

    2014-01-01

    embedded heating and cooling systems. Brussels, Belgium,radiant heating and cooling systems. Proceedings of Climaradiant heating and cooling systems: integration with a

  19. Critical review of water based radiant cooling system design methods

    E-Print Network [OSTI]

    Feng, Jingjuan Dove; Bauman, Fred; Schiavon, Stefano

    2014-01-01

    shown that radiant system cooling capacity could be enhancedof trends regarding radiant system cooling load analysis andEmbedded Radiant Heating and Cooling Systems, International

  20. Design and Control of Hydronic Radiant Cooling Systems

    E-Print Network [OSTI]

    Feng, Jingjuan Dove

    2014-01-01

    embedded heating and cooling systems. Brussels, Belgium,radiant heating and cooling systems. Proceedings of Climaof Slab Heating and Cooling Systems Studied by Dynamic

  1. Attrition reactor system

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  2. Attrition reactor system

    DOE Patents [OSTI]

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  3. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  4. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    SciTech Connect (OSTI)

    Steven J. Piet; Samuel E. Bays; Michael A. Pope; Gilles J. Youinou

    2010-11-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

  5. Thermal Storage with Conventional Cooling Systems 

    E-Print Network [OSTI]

    Kieninger, R. T.

    1994-01-01

    simple thermal energy storage system that already exists in almost every structure - concrete. Thermal storage calculations simulate sub-cooling of a building's structure during unoccupied times. During occupied times, the sub-cooled concrete reduces peak...

  6. Thermal-Hydraulic Analyses of Transients in an Actinide-Burner Reactor Cooled by Forced Convection of Lead Bismuth

    SciTech Connect (OSTI)

    Davis, Cliff Bybee

    2003-09-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead–bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed.

  7. Evaporative cooling enhanced cold storage system

    DOE Patents [OSTI]

    Carr, P.

    1991-10-15

    The invention provides an evaporatively enhanced cold storage system wherein a warm air stream is cooled and the cooled air stream is thereafter passed into contact with a cold storage unit. Moisture is added to the cooled air stream prior to or during contact of the cooled air stream with the cold storage unit to effect enhanced cooling of the cold storage unit due to evaporation of all or a portion of the added moisture. Preferably at least a portion of the added moisture comprises water condensed during the cooling of the warm air stream. 3 figures.

  8. Evaporative cooling enhanced cold storage system

    DOE Patents [OSTI]

    Carr, Peter (Cary, NC)

    1991-01-01

    The invention provides an evaporatively enhanced cold storage system wherein a warm air stream is cooled and the cooled air stream is thereafter passed into contact with a cold storage unit. Moisture is added to the cooled air stream prior to or during contact of the cooled air stream with the cold storage unit to effect enhanced cooling of the cold storage unit due to evaporation of all or a portion of the added moisture. Preferably at least a portion of the added moisture comprises water condensed during the cooling of the warm air stream.

  9. Performance Evaluation for Modular, Scalable Overhead Cooling Systems In Data Centers

    E-Print Network [OSTI]

    Xu, TengFang T.

    2009-01-01

    overhead cooling system Cooling system. The cooling systemTHE CHARACTERISTICS OF MODULAR, SCALABLE COOLING SYSTEMS ANDmodular, scalable cooling systems and servers 3.1 Modular,

  10. ATWS Transients for the 2400 MWt Gas-Cooled Fast Reactor

    SciTech Connect (OSTI)

    Cheng,L.Y.; Ludewig, H.

    2007-08-05

    Reactivity transients have been analyzed with an updated RELAPS-3D (ver. 2.4.2) system model of the pin core design for the 2400MWt gas-cooled fast reactor (GCFR). Additional reactivity parameters were incorporated in the RELAP5 point-kinetics model to account for reactivity feedbacks due to axial and radial expansion of the core, fuel temperature changes (Doppler effect), and pressure changes (helium density changes). Three reactivity transients without scram were analyzed and the incidents were initiated respectively by reactivity ramp, loss of load, and depressurization. During the course of the analysis the turbine bypass model for the power conversion unit (PCU) was revised to enable a better utilization of forced flow cooling after the PCU is tripped. The analysis of the reactivity transients demonstrates the significant impact of the PCU on system pressure and core flow. Results from the modified turbine bypass model suggest a success path for the GCFR to mitigate reactivity transients without scram.

  11. High temperature cooling system and method

    DOE Patents [OSTI]

    Loewen, Eric P.

    2006-12-12

    A method for cooling a heat source, a method for preventing chemical interaction between a vessel and a cooling composition therein, and a cooling system. The method for cooling employs a containment vessel with an oxidizable interior wall. The interior wall is oxidized to form an oxide barrier layer thereon, the cooling composition is monitored for excess oxidizing agent, and a reducing agent is provided to eliminate excess oxidation. The method for preventing chemical interaction between a vessel and a cooling composition involves introducing a sufficient quantity of a reactant which is reactive with the vessel in order to produce a barrier layer therein that is non-reactive with the cooling composition. The cooling system includes a containment vessel with oxidizing agent and reducing agent delivery conveyances and a monitor of oxidation and reduction states so that proper maintenance of a vessel wall oxidation layer occurs.

  12. Solar-powered cooling system

    DOE Patents [OSTI]

    Farmer, Joseph C.

    2015-07-28

    A solar-powered adsorption-desorption refrigeration and air conditioning system that uses nanostructural materials such as aerogels, zeolites, and sol gels as the adsorptive media. Refrigerant molecules are adsorbed on the high surface area of the nanostructural material while the material is at a relatively low temperature, perhaps at night. During daylight hours, when the nanostructural materials is heated by the sun, the refrigerant are thermally desorbed from the surface of the aerogel, thereby creating a pressurized gas phase in the vessel that contains the aerogel. This solar-driven pressurization forces the heated gaseous refrigerant through a condenser, followed by an expansion valve. In the condenser, heat is removed from the refrigerant, first by circulating air or water. Eventually, the cooled gaseous refrigerant expands isenthalpically through a throttle valve into an evaporator, in a fashion similar to that in more conventional vapor recompression systems.

  13. Tritium production analysis and management strategies for a Fluoride-salt-cooled high-temperature test reactor (FHTR)

    E-Print Network [OSTI]

    Rodriguez, Judy N

    2013-01-01

    The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational ...

  14. Containment system for supercritical water oxidation reactor

    DOE Patents [OSTI]

    Chastagner, P.

    1994-07-05

    A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

  15. Nuclear power reactor instrumentation systems handbook. Volume...

    Office of Scientific and Technical Information (OSTI)

    Nuclear power reactor instrumentation systems handbook. Volume 1 Citation Details In-Document Search Title: Nuclear power reactor instrumentation systems handbook. Volume 1 You...

  16. Cooling load differences between radiant and air systems

    E-Print Network [OSTI]

    Feng, Jingjuan Dove; Schiavon, Stefano; Bauman, Fred

    2013-01-01

    radiant heating and cooling systems, in: Proceedings ofInc, Altanta,GA, 2009. Cooling load differences betweensurface level 24-hour total cooling energy between radiant

  17. Development of a High Temperature Gas-Cooled Reactor TRISO-coated particle fuel chemistry model

    E-Print Network [OSTI]

    Diecker, Jane T

    2005-01-01

    The first portion of this work is a comprehensive analysis of the chemical environment in a High Temperature Gas-Cooled Reactor TRISO fuel particle. Fission product inventory versus burnup is calculated. Based on those ...

  18. Application of the technology neutral framework to sodium cooled fast reactors

    E-Print Network [OSTI]

    Johnson, Brian C. (Brian Carl)

    2010-01-01

    Sodium cooled fast reactors (SFRs) are considered as a novel example to exercise the Technology Neutral Framework (TNF) proposed in NUREG- 1860. One reason for considering SFRs is that they have historically had a licensing ...

  19. Application of the Technology Neutral Framework to Sodium-­Cooled Fast Reactors

    E-Print Network [OSTI]

    Johnson, Brian C.

    Sodium cooled fast reactors (SFRs) are considered as a novel example to exercise the Technology Neutral Framework (TNF) proposed in NUREG-1860. One reason for considering SFRs is that they have historically had a licensing ...

  20. Implementation of vented fuel assemblies in the supercritical CO?-cooled fast reactor

    E-Print Network [OSTI]

    McKee, Stephanie A

    2008-01-01

    Analysis has been undertaken to investigate the utilization of fuel assembly venting in the reference design of the gas-cooled fast reactor under study as part of the larger research effort at MIT under Gen-IV NERI Project ...

  1. Applying risk informed methodologies to improve the economics of sodium-cooled fast reactors

    E-Print Network [OSTI]

    Nitta, Christopher C

    2010-01-01

    In order to support the increasing demand for clean sustainable electricity production and for nuclear waste management, the Sodium-Cooled Fast Reactor (SFR) is being developed. The main drawback has been its high capital ...

  2. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    reactor pas- sive reactivity control systems: LEM and LIM”. In: Nuclearnuclear reactor. Also it did not feature oxygen content controlControl Rods Reactor Concept and Plant Dynamics Analyses”. In: Journal of Nuclear

  3. Evaluation of cooling performance of thermally activated building system with evaporative cooling source for typical United States climates

    E-Print Network [OSTI]

    Feng, Jingjuan; Bauman, Fred

    2013-01-01

    embedded surface cooling system and TABS systems THERMALwas simulated. The radiant cooling system was an exposedcooling + radiant cooling system alone may not be able to

  4. Gas-cooled fast breeder reactor. Quarterly progress report, November 1, 1979 through January 31, 1980

    SciTech Connect (OSTI)

    Not Available

    1980-02-01

    Information is presented concerning the nuclear steam supply system; reactor core; systems engineering; safety and reliability; and circulator test facility.

  5. Hot gas path component cooling system

    DOE Patents [OSTI]

    Lacy, Benjamin Paul; Bunker, Ronald Scott; Itzel, Gary Michael

    2014-02-18

    A cooling system for a hot gas path component is disclosed. The cooling system may include a component layer and a cover layer. The component layer may include a first inner surface and a second outer surface. The second outer surface may define a plurality of channels. The component layer may further define a plurality of passages extending generally between the first inner surface and the second outer surface. Each of the plurality of channels may be fluidly connected to at least one of the plurality of passages. The cover layer may be situated adjacent the second outer surface of the component layer. The plurality of passages may be configured to flow a cooling medium to the plurality of channels and provide impingement cooling to the cover layer. The plurality of channels may be configured to flow cooling medium therethrough, cooling the cover layer.

  6. Corium Retention for High Power Reactors by An In-Vessel Core Catcher in Combination with External Reactor Vessel Cooling

    SciTech Connect (OSTI)

    Joy L. Rempe; D. L. Knudson; K. G. Condie; K. Y. Suh; F. -B. Cheung; S. -B. Kim

    2004-05-01

    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 lower 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 Pressurized Water Reactor (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for in-vessel retention (IVR), resulted in the United States Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). Accordingly, 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. However, it is not clear that currently-proposed methods to achieve ERVC will provide sufficient heat removal for higher power reactors. A US–Korean International Nuclear Energy Research Initiative (INERI) project has been initiated 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) will determine if IVR is feasible for reactors up to 1500 MWe. This paper summarizes results from the first year of this 3-year project.

  7. Reactor User Interface Technology Development Roadmaps for a High Temperature Gas-Cooled Reactor Outlet Temperature of 750 degrees C

    SciTech Connect (OSTI)

    Ian Mckirdy

    2010-12-01

    This report evaluates the technology readiness of the interface components that are required to transfer high-temperature heat from a High Temperature Gas-Cooled Reactor (HTGR) to selected industrial applications. This report assumes that the HTGR operates at a reactor outlet temperature of 750°C and provides electricity and/or process heat at 700°C to conventional process applications, including the production of hydrogen.

  8. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    SciTech Connect (OSTI)

    Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental results show similar trends as the computational fluid dynamics (CFD) results presented in this report; however, some differences exist that will need to be assessed in future studies. The results of this testing will be used to improve the diode design to be tested in the liquid salt loop system.

  9. Adaptive polynomial chaos techniques for uncertainty quantification of a gas cooled fast reactor transient

    SciTech Connect (OSTI)

    Perko, Z.; Gilli, L.; Lathouwers, D.; Kloosterman, J. L.

    2013-07-01

    Uncertainty quantification plays an increasingly important role in the nuclear community, especially with the rise of Best Estimate Plus Uncertainty methodologies. Sensitivity analysis, surrogate models, Monte Carlo sampling and several other techniques can be used to propagate input uncertainties. In recent years however polynomial chaos expansion has become a popular alternative providing high accuracy at affordable computational cost. This paper presents such polynomial chaos (PC) methods using adaptive sparse grids and adaptive basis set construction, together with an application to a Gas Cooled Fast Reactor transient. Comparison is made between a new sparse grid algorithm and the traditionally used technique proposed by Gerstner. An adaptive basis construction method is also introduced and is proved to be advantageous both from an accuracy and a computational point of view. As a demonstration the uncertainty quantification of a 50% loss of flow transient in the GFR2400 Gas Cooled Fast Reactor design was performed using the CATHARE code system. The results are compared to direct Monte Carlo sampling and show the superior convergence and high accuracy of the polynomial chaos expansion. Since PC techniques are easy to implement, they can offer an attractive alternative to traditional techniques for the uncertainty quantification of large scale problems. (authors)

  10. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, James E. (Greensburg, PA); Meuschke, Robert E. (Pittsburgh, PA)

    1995-01-01

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  11. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  12. Thermal hydraulic design of a 2400 MW t?h? direct supercritical CO?-cooled fast reactor

    E-Print Network [OSTI]

    Pope, Michael A. (Michael Alexander)

    2006-01-01

    The gas cooled fast reactor (GFR) has received new attention as one of the basic concepts selected by the Generation-IV International Forum (GIF) for further investigation. Currently, the reference GFR is a helium-cooled ...

  13. Final report-passive safety optimization in liquid sodium-cooled reactors.

    SciTech Connect (OSTI)

    Cahalana, J. E.; Hahn, D.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2007-08-13

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety Implications of Advanced Technology Power Conversion and Design Innovations and Simplifications: Investigations of supercritical CO{sub 2} gas turbine Brayton cycles coupled to the sodium-cooled reactors and innovative concepts for sodium-to-CO{sub 2} heat exchangers were performed to discover new designs for high efficiency electricity production. The objective of the analyses was to characterize the design and safety performance of equipment needed to implement the new power cycle. The project included considerations of heat transfer and power conversion systems arrangements and evaluations of systems performance. Task 4--Post Accident Heat Removal and In-Vessel Retention: Test plans were developed to evaluate (1) freezing and plugging of molten metallic fuel in subassembly geometry, (2) retention of metallic fuel core melt debris within reactor vessel structures, and (3) consequences of intermixing of high pressure CO{sub 2} and sodium. The objective of the test plan development was to provide planning for measurements of data needed to characterize the consequences of very low probability accident sequences unique to metallic fuel and CO{sub 2} Brayton power cycles. The project produced three test plans ready for execution.

  14. Effectiveness-weighted control method for a cooling system

    DOE Patents [OSTI]

    Campbell, Levi A.; Chu, Richard C.; David, Milnes P.; Ellsworth Jr., Michael J.; Iyengar, Madhusudan K.; Schmidt, Roger R.; Simons, Robert E.

    2015-12-15

    Energy efficient control of cooling system cooling of an electronic system is provided based, in part, on weighted cooling effectiveness of the components. The control includes automatically determining speed control settings for multiple adjustable cooling components of the cooling system. The automatically determining is based, at least in part, on weighted cooling effectiveness of the components of the cooling system, and the determining operates to limit power consumption of at least the cooling system, while ensuring that a target temperature associated with at least one of the cooling system or the electronic system is within a desired range by provisioning, based on the weighted cooling effectiveness, a desired target temperature change among the multiple adjustable cooling components of the cooling system. The provisioning includes provisioning applied power to the multiple adjustable cooling components via, at least in part, the determined control settings.

  15. Effectiveness-weighted control of cooling system components

    DOE Patents [OSTI]

    Campbell, Levi A.; Chu, Richard C.; David, Milnes P.; Ellsworth Jr., Michael J.; Iyengar, Madhusudan K.; Schmidt, Roger R.; Simmons, Robert E.

    2015-12-22

    Energy efficient control of cooling system cooling of an electronic system is provided based, in part, on weighted cooling effectiveness of the components. The control includes automatically determining speed control settings for multiple adjustable cooling components of the cooling system. The automatically determining is based, at least in part, on weighted cooling effectiveness of the components of the cooling system, and the determining operates to limit power consumption of at least the cooling system, while ensuring that a target temperature associated with at least one of the cooling system or the electronic system is within a desired range by provisioning, based on the weighted cooling effectiveness, a desired target temperature change among the multiple adjustable cooling components of the cooling system. The provisioning includes provisioning applied power to the multiple adjustable cooling components via, at least in part, the determined control settings.

  16. Preventing fuel failure for a beyond design basis accident in a fluoride salt cooled high temperature reactor

    E-Print Network [OSTI]

    Minck, Matthew J. (Matthew Joseph)

    2013-01-01

    The fluoride salt-cooled high-temperature reactor (FHR) combines high-temperature coated-particle fuel with a high-temperature salt coolant for a reactor with unique market and safety characteristics. This combination can ...

  17. GentleCool: Cooling Aware Proactive Workload Scheduling in Multi-Machine Systems

    E-Print Network [OSTI]

    Simunic, Tajana

    GentleCool: Cooling Aware Proactive Workload Scheduling in Multi-Machine Systems Raid Ayoub characteristics of the workload running on the system. We propose a scheduling framework called GentleCool, a proactive multi-tier approach for significantly lowering the fan cooling costs in highly utilized systems

  18. CONTROL SYSTEM FOR SOLAR HEATING and COOLING

    E-Print Network [OSTI]

    Dols, C.

    2010-01-01

    solar heated, boosted, or heated entirely in the auxiliary heater)for the solar-heated hot water. This heater can be seen insolar heating and cooling system, showing plumbing runs containing solenoid valves, auxiliary heater (

  19. CONTROL SYSTEM FOR SOLAR HEATING and COOLING

    E-Print Network [OSTI]

    Dols, C.

    2010-01-01

    LBL buildings, with the solar collectors on the roof, theCBB 757-5496 Figure 3: Solar Collectors Mounted· on the RoofSolar Heating and Cooling Systems. The components include Collectors (

  20. Rapid starting methanol reactor system

    DOE Patents [OSTI]

    Chludzinski, Paul J. (38 Berkshire St., Swampscott, MA 01907); Dantowitz, Philip (39 Nancy Ave., Peabody, MA 01960); McElroy, James F. (12 Old Cart Rd., Hamilton, MA 01936)

    1984-01-01

    The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

  1. NASA Marshall Space Flight Center Improves Cooling System Performance...

    Office of Environmental Management (EM)

    Marshall Space Flight Center Improves Cooling System Performance NASA Marshall Space Flight Center Improves Cooling System Performance NASA Marshall Space Flight Center Improves...

  2. Energy Efficient HVAC System for Distributed Cooling/Heating...

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

    Efficient HVAC System for Distributed CoolingHeating with Thermoelectric Devices Energy Efficient HVAC System for Distributed CoolingHeating with Thermoelectric Devices 2012 DOE...

  3. The integration of cryogenic cooling systems with superconducting electronic systems

    E-Print Network [OSTI]

    Green, Michael A.

    2011-01-01

    applications for superconductivity have low heat loads in aTransactions on Applied Superconductivity t I, P 2615, (Cooling Systems With Superconducting Electronic Systems M.

  4. Gas hydrate cool storage system

    DOE Patents [OSTI]

    Ternes, Mark P. (Knoxville, TN); Kedl, Robert J. (Oak Ridge, TN)

    1985-01-01

    This invention is a process for formation of a gas hydrate to be used as a cool storage medium using a refrigerant in water. Mixing of the immiscible refrigerant and water is effected by addition of a surfactant and agitation. The difficult problem of subcooling during the process is overcome by using the surfactant and agitation and performance of the process significantly improves and approaches ideal.

  5. Experimental comparison of zone cooling load between radiant and air systems

    E-Print Network [OSTI]

    Feng, Jingjuan Dove; Bauman, Fred; Schiavon, Stefano

    2014-01-01

    be used for radiant system cooling load, which should beKeyword Radiant Cooling System, Cooling load, Air system,guidance on radiant system cooling load prediction and

  6. Reactor control rod timing system

    DOE Patents [OSTI]

    Wu, Peter T. K. (Clifton Park, NY)

    1982-01-01

    A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  7. Reactor control rod timing system

    SciTech Connect (OSTI)

    Wu, P.T.

    1982-02-09

    A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (Above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  8. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    DOE Patents [OSTI]

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  9. Performance of the Lead-Alloy-Cooled Reactor Concept Balanced for Actinide Burning and Electricity Production

    SciTech Connect (OSTI)

    Hejzlar, Pavel [Massachusetts Institute of Technology (United States); Davis, Cliff B. [Idaho National Engineering and Environmental Laboratory (United States)

    2004-09-15

    A lead-bismuth-cooled fast reactor concept targeted for a balanced mission of actinide burning and low-cost electricity production is proposed and its performance analyzed. The design explores the potential benefits of thorium-based fuel in actinide-burning cores, in particular in terms of the reduction of the large reactivity swing and enhancement of the small Doppler coefficient typical of fertile-free actinide burners. Reduced electricity production cost is pursued through a longer cycle length than that used for fertile-free burners and thus a higher capacity factor. It is shown that the concept can achieve a high transuranics destruction rate, which is only 20% lower than that of an accelerator-driven system with fertile-free fuel. The small negative fuel temperature reactivity coefficient, small positive coolant temperature reactivity coefficient, and negative core radial expansion coefficient provide self-regulating characteristics so that the reactor is capable of inherent shutdown during major transients without scram, as in the Integral Fast Reactor. This is confirmed by thermal-hydraulic analysis of several transients without scram, including primary coolant pump trip, station blackout, and reactivity step insertion, which showed that the reactor was able to meet all identified thermal limits. However, the benefits of high actinide consumption and small reactivity swing can be attained only if the uranium from the discharged fuel is separated and not recycled. This additional uranium separation step and thorium reprocessing significantly increase the fuel cycle costs. Because the higher fuel cycle cost has a larger impact on the overall cost of electricity than the savings from the higher capacity factor afforded through use of thorium, this concept appears less promising than the fertile-free actinide burners.

  10. A parametric study of the breeding ratio in sodium cooled fast breeder reactors 

    E-Print Network [OSTI]

    Sobey, Thomas Milburn

    1969-01-01

    of fuel as opposed to the destruction of it on an isotopic level, rather than an elemental level. The objective of this -thesis is to investigate the dependence of the breeding ratio i. n a Sodium Cooled Fast Breeder Reactor on the 5sotopic composii.... 'FPHETRIC STIJDY OF THE BREEDING RATIO SODIlnnJ COOL J'. P FAST BREEJ)ER REACTORS A Tgesis TEO'. ~YS ', ~JILBlJRY SUBEY Sugmitt d to tlso Graduate College of Icxa~ ASH I'niversity in '. artia1 Iuliiliniost of t!ns reguireeents for tge deg ee...

  11. HTGR (High Temperature Gas-Cooled Reactor) ingress analysis using MINET

    SciTech Connect (OSTI)

    Van Tuyle, G.J.; Yang, J.W.; Kroeger, P.G.; Mallen, A.N.; Aronson, A.L.

    1989-04-01

    Modeling of water/steam ingress into the primary (helium) cooling circuit of a High Temperature Gas-Cooled Reactor (HTGR) is described. This modeling was implemented in the MINET Code, which is a program for analyzing transients in intricate fluid flow and heat transfer networks. Results from the simulation of a water ingress event postulated for the Modular HTGR are discussed. 27 refs., 6 figs., 6 tabs.

  12. Cooling system for continuous metal casting machines

    DOE Patents [OSTI]

    Draper, R.; Sumpman, W.C.; Baker, R.J.; Williams, R.S.

    1988-06-07

    A continuous metal caster cooling system is provided in which water is supplied in jets from a large number of small nozzles against the inner surface of rim at a temperature and with sufficient pressure that the velocity of the jets is sufficiently high that the mode of heat transfer is substantially by forced convection, the liquid being returned from the cooling chambers through return pipes distributed interstitially among the nozzles. 9 figs.

  13. Cooling system for continuous metal casting machines

    DOE Patents [OSTI]

    Draper, Robert (Churchill Boro, PA); Sumpman, Wayne C. (North Huntingdon, PA); Baker, Robert J. (Wilkins Township, Allegheny County, PA); Williams, Robert S. (Plum Borough, PA)

    1988-01-01

    A continuous metal caster cooling system is provided in which water is supplied in jets from a large number of small nozzles 19 against the inner surface of rim 13 at a temperature and with sufficient pressure that the velocity of the jets is sufficiently high that the mode of heat transfer is substantially by forced convection, the liquid being returned from the cooling chambers 30 through return pipes 25 distributed interstitially among the nozzles.

  14. Simulation of ultrasonic inspection for sodium cooled reactors using CIVA

    SciTech Connect (OSTI)

    Reverdy, F. [CEA, LIST, Departement Imagerie et Simulation Pour Le Controle, F-91191 Gif-sur-Yvette (France); Baque, F. [CEA, DEN DTN, 13108 Saint Paul lez Durance Cedex (France); Lu, B.; Jezzine, K.; Dorval, V. [CEA, LIST, Departement Imagerie et Simulation Pour Le Controle, F-91191 Gif-sur-Yvette (France); Augem, J. M. [EDF, 12-14 avenue dutrievoz, 69628 Villeurbanne (France)

    2011-07-01

    In-service inspection of sodium fast reactors (SFR) requires the development of non-destructive techniques adapted to the harsh conditions of the environment (opaque and hot) and the complexity of the examination (large and littered reactor block). Ultrasonic techniques are seen as suitable candidates for the inspection of SFRs and two approaches are being followed: inside inspection where transducers are directly immersed in sodium coolant and inspection from outside with transducers positioned along the wall of the main vessel. Probe design and inspection performances can be predicted by using comprehensive models that can take into account the various variables of the problem. These models are explained in this paper. (authors)

  15. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    SciTech Connect (OSTI)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  16. Reactor vessel annealing system

    DOE Patents [OSTI]

    Miller, Phillip E. (Greensburg, PA); Katz, Leonoard R. (Pittsburgh, PA); Nath, Raymond J. (Murrysville, PA); Blaushild, Ronald M. (Export, PA); Tatch, Michael D. (Randolph, NJ); Kordalski, Frank J. (White Oak, PA); Wykstra, Donald T. (Pittsburgh, PA); Kavalkovich, William M. (Monroeville, PA)

    1991-01-01

    A system for annealing a vessel (14) in situ by heating the vessel (14) to a defined temperature, composed of: an electrically operated heater assembly (10) insertable into the vessel (14) for heating the vessel (14) to the defined temperature; temperature monitoring components positioned relative to the heater assembly (10) for monitoring the temperature of the vessel (14); a controllable electric power supply unit (32-60) for supplying electric power required by the heater assembly (10); a control unit (80-86) for controlling the power supplied by the power supply unit (32-60); a first vehicle (2) containing the power supply unit (32-60); a second vehicle (4) containing the control unit (80-86); power conductors (18,22) connectable between the power supply unit (32-60) and the heater unit (10) for delivering the power supplied by the power supply unit (32-60) to the heater assembly (10); signal conductors (20,24) connectable between the temperature monitoring components and the control unit (80-86) for delivering temperature indicating signals from the temperature monitoring components to the control unit (80-86); and control conductors (8) connectable between the control unit (80-86) and the power supply unit (32-60) for delivering to the power supply unit (32-60) control signals for controlling the level of power supplied by the power supply unit (32-60) to the heater assembly (10).

  17. Passive cooling system for a vehicle

    DOE Patents [OSTI]

    Hendricks, Terry Joseph; Thoensen, Thomas

    2005-11-15

    A passive cooling system for a vehicle (114) transfers heat from an overheated internal component, for example, an instrument panel (100), to an external portion (116) of the vehicle (114), for example, a side body panel (126). The passive cooling system includes one or more heat pipes (112) having an evaporator section (118) embedded in the overheated internal component and a condenser section (120) at the external portion (116) of the vehicle (114). The evaporator (118) and condenser (120) sections are in fluid communication. The passive cooling system may also include a thermally conductive film (140) for thermally connecting the evaporator sections (118) of the heat pipes (112) to each other and to the instrument panel (100).

  18. Passive Cooling System for a Vehicle

    DOE Patents [OSTI]

    Hendricks, T. J.; Thoensen, T.

    2005-11-15

    A passive cooling system for a vehicle (114) transfers heat from an overheated internal component, for example, an instrument panel (100), to an external portion (116) of the vehicle (114), for example, a side body panel (126). The passive cooling system includes one or more heat pipes (112) having an evaporator section (118) embedded in the overheated internal component and a condenser section (120) at the external portion (116) of the vehicle (114). The evaporator (118) and condenser (120) sections are in fluid communication. The passive cooling system may also include a thermally conductive film (140) for thermally connecting the evaporator sections (118) of the heat pipes (112) to each other and to the instrument panel (100).

  19. High-temperature gas-cooled reactors: preliminary safety and environmental information document. Volume IV

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    Information is presented concerning medium-enriched uranium/thorium once-through fuel cycle; medium-enrichment uranium-233/thorium recycle fuel; high-enrichment uranium-235/thorium recycle (spiked) fuel cycle; high-enrichment uranium-233/thorium recycle (spiked) fuel cycle; and gas-turbine high-temperature gas-cooled reactor.

  20. Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

    SciTech Connect (OSTI)

    Lee Nelson

    2011-09-01

    This report is a summary of analyses performed by the NGNP project to determine whether it is technically and economically feasible to integrate high temperature gas cooled reactor (HTGR) technology into industrial processes. To avoid an overly optimistic environmental and economic baseline for comparing nuclear integrated and conventional processes, a conservative approach was used for the assumptions and calculations.

  1. Design of a 2400MW liquid-salt cooled flexible conversion ratio reactor

    E-Print Network [OSTI]

    Petroski, Robert C

    2008-01-01

    A 2400MWth liquid-salt cooled flexible conversion ratio reactor was designed, utilizing the ternary chloride salt NaCl-KCl-MgCI2 (30%-20%-50%) as coolant. The reference design uses a wire-wrapped, hex lattice core, and is ...

  2. Process Cooling Pumping Systems Analysis 

    E-Print Network [OSTI]

    Sherman, C.

    2008-01-01

    An analysis of the mill water pumping systems at a North American manufacturing facility was conducted late las year and the following issues were observed: 1. Overpumping – Both systems were overpumped to a significant degree against...

  3. Thermodynamics cycle analysis and numerical modeling of thermoelastic cooling systems

    E-Print Network [OSTI]

    Rubloff, Gary W.

    Thermodynamics cycle analysis and numerical modeling of thermoelastic cooling systems Suxin Qian level. However, a thermoelastic cooling system integrated with heat transfer fluid loops have not been;2012) (a.k.a. elastocaloric cooling). These solid-state cooling systems offer us alternatives to eliminate

  4. Cool and Save: Cooling Aware Dynamic Workload Scheduling in Multi-socket CPU Systems

    E-Print Network [OSTI]

    Simunic, Tajana

    Cool and Save: Cooling Aware Dynamic Workload Scheduling in Multi-socket CPU Systems Raid Ayoub, dissipating the high temper- ature requires a large and energy hungry cooling system which increases the cost and fan control in multi-socket systems have been designed sep- arately leading to less efficient

  5. Cooling load design tool for UFAD systems.

    E-Print Network [OSTI]

    Bauman, Fred; Schiavon, Stefano; Webster, Tom; Lee, Kwang Ho

    2010-01-01

    De- velopment of a Simplified Cooling Load Design Tool forand C. Benedek. 2007. “Cooling airflow design calculationscalculation method for design cooling loads in underfloor

  6. Cooling system for a gas turbine

    DOE Patents [OSTI]

    Wilson, Ian David (Mauldin, SC); Salamah, Samir Armando (Niskayuna, NY); Bylina, Noel Jacob (Niskayuna, NY)

    2003-01-01

    A plurality of arcuate circumferentially spaced supply and return manifold segments are arranged on the rim of a rotor for respectively receiving and distributing cooling steam through exit ports for distribution to first and second-stage buckets and receiving spent cooling steam from the first and second-stage buckets through inlet ports for transmission to axially extending return passages. Each of the supply and return manifold segments has a retention system for precluding substantial axial, radial and circumferential displacement relative to the rotor. The segments also include guide vanes for minimizing pressure losses in the supply and return of the cooling steam. The segments lie substantially equal distances from the centerline of the rotor and crossover tubes extend through each of the segments for communicating steam between the axially adjacent buckets of the first and second stages, respectively.

  7. Method of fabricating a cooled electronic system

    DOE Patents [OSTI]

    Chainer, Timothy J; Gaynes, Michael A; Graybill, David P; Iyengar, Madhusudan K; Kamath, Vinod; Kochuparambil, Bejoy J; Schmidt, Roger R; Schultz, Mark D; Simco, Daniel P; Steinke, Mark E

    2014-02-11

    A method of fabricating a liquid-cooled electronic system is provided which includes an electronic assembly having an electronics card and a socket with a latch at one end. The latch facilitates securing of the card within the socket. The method includes providing a liquid-cooled cold rail at the one end of the socket, and a thermal spreader to couple the electronics card to the cold rail. The thermal spreader includes first and second thermal transfer plates coupled to first and second surfaces on opposite sides of the card, and thermally conductive extensions extending from end edges of the plates, which couple the respective transfer plates to the liquid-cooled cold rail. The extensions are disposed to the sides of the latch, and the card is securable within or removable from the socket using the latch without removing the cold rail or the thermal spreader.

  8. Redesigning Process Cooling Systems in the Plastics Industry 

    E-Print Network [OSTI]

    Anderson, G. R.

    2006-01-01

    systems were designed with one thing in mind – ensuring adequate capacity. Energy consumption was a much lower priority with their process cooling systems, resulting in inefficient chillers, oversized pumps, undersized cooling towers, and poorly... sequenced operations. Lifetime decided to step back and evaluate their entire cooling system for opportunities to reduce energy use after they recognized the potential for “free cooling” from the chiller’s cooling towers during the winter. Lifetime’s...

  9. Preliminary Study of Bypass Flow in Prismatic Core of Very High Temperature Reactor Using Small-Scale Model 

    E-Print Network [OSTI]

    Kanjanakijkasem, Worasit 1975-

    2012-11-29

    energy are: gas-cooled fast reactor (GFR); lead-cooled fast reactor (LFR); molten salt reactor (MSR); sodium-cooled fast reactor (SFR); supercritical water-cooled reactor (SCWR); and very high-temperature gas-cooled reactor (VHTR) [3]. These reactor...-temperature gas-cooled reactor (VHTR) is a graphite-moderated, helium-cooled, thermal neutron spectrum reactor with a once-through uranium fuel cycle [4]. The reactor core can be either a prismatic block or a pebble-bed core. The VHTR system is designed to be a...

  10. Shutdown system for a nuclear reactor

    DOE Patents [OSTI]

    Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

    1984-06-05

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

  11. Shutdown system for a nuclear reactor

    DOE Patents [OSTI]

    Groh, Edward F. (Naperville, IL); Olson, Arne P. (Western Springs, IL); Wade, David C. (Naperville, IL); Robinson, Bryan W. (Oak Lawn, IL)

    1984-01-01

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

  12. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    SciTech Connect (OSTI)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  13. Simulation of cooling systems in gas turbines

    SciTech Connect (OSTI)

    Ebenhoch, G.; Speer, T.M. [Motoren- und Turbinen-Union Muenchen GmbH (Germany)

    1996-04-01

    The design of cooling systems for gas turbine engine blades and vanes calls for efficient simulation programs. The main purpose of the described program is to determine the complete boundary condition at the coolant side to support a temperature calculation for the solid. For the simulation of convection and heat pick up of the coolant flow, pressure loss, and further effects to be found in a rotating frame, the cooling systems are represented by networks of nodes and flow elements. Within each flow element the fluid flow is modeled by a system of ordinary differential equations based on the one-dimensional conservation of mass, momentum, and energy. In this respect, the computer program differs from many other network computation programs. Concerning cooling configurations in rotating systems, the solution for a single flow element or the entire flow system is not guaranteed to be unique. This is due to rotational forces in combination with heat transfer and causes considerable computational difficulties, which can be overcome by a special path following method in which the angular velocity is selected as the parameter of homotopy. Results of the program are compared with measurements for three applications.

  14. Reactor and Nuclear Systems Division (RNSD)

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

    RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation...

  15. A Qualitative Assessment of Thorium-Based Fuels in Supercritical Pressure Water Cooled Reactors

    SciTech Connect (OSTI)

    Weaver, Kevan Dean; Mac Donald, Philip Elsworth

    2002-10-01

    The requirements for the next generation of reactors include better economics and safety, waste minimization (particularly of the long-lived isotopes), and better proliferation resistance (both intrinsic and extrinsic). A supercritical pressure water cooled reactor has been chosen as one of the lead contenders as a Generation IV reactor due to the high thermal efficiency and compact/simplified plant design. In addition, interest in the use of thorium-based fuels for Generation IV reactors has increased based on the abundance of thorium, and the minimization of transuranics in a neutron flux; as plutonium (and thus the minor actinides) is not a by-product in the thorium chain. In order to better understand the possibility of the combination of these concepts to meet the Generation IV goals, the qualitative burnup potential and discharge isotopics of thorium and uranium fuel were studied using pin cell analyses in a supercritical pressure water cooled reactor environment. Each of these fertile materials were used in both nitride and metallic form, with light water reactor grade plutonium and minor actinides added. While the uranium-based fuels achieved burnups that were 1.3 to 2.7 times greater than their thorium-based counterparts, the thorium-based fuels destroyed 2 to 7 times more of the plutonium and minor actinides. The fission-to-capture ratio is much higher in this reactor as compared to PWR’s and BWR’s due to the harder neutron spectrum, thus allowing more efficient destruction of the transuranic elements. However, while the uranium-based fuels do achieve a net depletion of plutonium and minor actinides, the breeding of these isotopes limits this depletion; especially as compared to the thorium-based fuels.

  16. Controlled cooling of an electronic system based on projected conditions

    DOE Patents [OSTI]

    David, Milnes P.; Iyengar, Madhusudan K.; Schmidt, Roger R.

    2015-08-18

    Energy efficient control of a cooling system cooling an electronic system is provided based, in part, on projected conditions. The control includes automatically determining an adjusted control setting(s) for an adjustable cooling component(s) of the cooling system. The automatically determining is based, at least in part, on projected power consumed by the electronic system at a future time and projected temperature at the future time of a heat sink to which heat extracted is rejected. The automatically determining operates to reduce power consumption of the cooling system and/or the electronic system while ensuring that at least one targeted temperature associated with the cooling system or the electronic system is within a desired range. The automatically determining may be based, at least in part, on an experimentally obtained model(s) relating the targeted temperature and power consumption of the adjustable cooling component(s) of the cooling system.

  17. New baseline for the magnet cooling system Yury Ivanyushenkov

    E-Print Network [OSTI]

    McDonald, Kirk

    1 New baseline for the magnet cooling system Yury Ivanyushenkov Engineering and Instrumentation Department, Rutherford Appleton Laboratory #12;2 Liquid nitrogen cooling system: Conceptual points · Magnet the magnet as a stand alone system. #12;3 Liquid nitrogen cooling system: Diagram Drawn by Peter Titus #12

  18. Low pressure cooling seal system for a gas turbine engine

    DOE Patents [OSTI]

    Marra, John J

    2014-04-01

    A low pressure cooling system for a turbine engine for directing cooling fluids at low pressure, such as at ambient pressure, through at least one cooling fluid supply channel and into a cooling fluid mixing chamber positioned immediately downstream from a row of turbine blades extending radially outward from a rotor assembly to prevent ingestion of hot gases into internal aspects of the rotor assembly. The low pressure cooling system may also include at least one bleed channel that may extend through the rotor assembly and exhaust cooling fluids into the cooling fluid mixing chamber to seal a gap between rotational turbine blades and a downstream, stationary turbine component. Use of ambient pressure cooling fluids by the low pressure cooling system results in tremendous efficiencies by eliminating the need for pressurized cooling fluids for sealing this gap.

  19. Summary of space nuclear reactor power systems, 1983--1992

    SciTech Connect (OSTI)

    Buden, D.

    1993-08-11

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  20. Measurement of Flow Phenomena in a Lower Plenum Model of a Prismatic Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Hugh M. McIlroy, Jr.; Donald M. McEligot; Robert J. Pink

    2008-05-01

    Mean-velocity-field and turbulence data are presented that measure turbulent flow phenomena in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. The data were obtained in the Matched-Index-of-Refraction (MIR) facility at Idaho National Laboratory (INL) and are offered for assessing computational fluid dynamics (CFD) software. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. This paper reviews the experimental apparatus and procedures, presents a sample of the data set, and reviews the INL Standard Problem. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum consists of multiple jets injected into a confined cross flow - with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the mineral oil working fluid so that optical techniques may be employed for the measurements. The benefit of the MIR technique is that it permits optical measurements to determine flow characteristics in complex passages in and around objects to be obtained without locating intrusive transducers that will disturb the flow field and without distortion of the optical paths. An advantage of the INL system is its large size, leading to improved spatial and temporal resolution compared to similar facilities at smaller scales. A three-dimensional (3-D) Particle Image Velocimetry (PIV) system was used to collect the data. Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. Uncertainty analysis and a discussion of the standard problem are included. The measurements reveal undeveloped, non-uniform, turbulent flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and presentations that describe the component flows at specific regions in the model. Information on inlet conditions are also presented.

  1. Tandem Mirror Reactor Systems Code (Version I)

    SciTech Connect (OSTI)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

  2. NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions

    SciTech Connect (OSTI)

    Wayne Moe

    2013-05-01

    This document provides key definitions, plant capabilities, and inputs and assumptions related to the Next Generation Nuclear Plant to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor. These definitions, capabilities, and assumptions were extracted from a number of NGNP Project sources such as licensing related white papers, previously issued requirement documents, and preapplication interactions with the Nuclear Regulatory Commission (NRC).

  3. Geothermal Heating and Cooling Systems Featured on NBC Nightly...

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

    Heating and Cooling Systems Featured on NBC Nightly News Geothermal Heating and Cooling Systems Featured on NBC Nightly News April 13, 2009 - 11:24am Addthis NBC Nightly News...

  4. EIS-0121: Alternative Cooling Water Systems, Savannah River Plant, Aiken, South Carolina

    Office of Energy Efficiency and Renewable Energy (EERE)

    The purpose of this Environmental Impact Statement (EIS) is to provide environmental input into the selection and implementation of cooling water systems for thermal discharges from K– and C-Reactors and from a coal-fired powerhouse in the D-Area at the Savannah River Plant (SRP)

  5. Cooling System Basics | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative Fuels DataEnergy Webinar:IAbout Us » Contact PPPOContributingCooling System

  6. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    SciTech Connect (OSTI)

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

    2014-03-01

    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  7. Materials testing and development of functionally graded composite fuel cladding and piping for the Lead-Bismuth cooled nuclear reactor

    E-Print Network [OSTI]

    Fray, Elliott Shepard

    2013-01-01

    This study has extended the development of an exciting technology which promises to enable the Pb-Bi eutectic cooled reactors to operate at temperatures up to 650-700°C. This new technology is a functionally graded composite ...

  8. Conceptual design of a lead-bismuth cooled fast reactor with in-vessel direct-contact steam generation

    E-Print Network [OSTI]

    Buongiorno, Jacopo, 1971-

    2001-01-01

    The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid ...

  9. Conceptual Design of a Lead-Bismuth Cooled Fast Reactor with In-Vessel Direct-Contact Steam Generation

    E-Print Network [OSTI]

    Buongiorno, J.

    The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid ...

  10. Investigation and design of a secure, transportable fluoride-salt-cooled high-temperature reactor (TFHR) for isolated locations

    E-Print Network [OSTI]

    Macdonald, Ruaridh (Ruaridh R.)

    2014-01-01

    In this work we describe a preliminary design for a transportable fluoride salt cooled high temperature reactor (TFHR) intended for use as a variable output heat and electricity source for off-grid locations. The goals of ...

  11. The design of a functionally graded composite for service in high temperature lead and lead-bismuth cooled nuclear reactors

    E-Print Network [OSTI]

    Short, Michael Philip

    2010-01-01

    A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700°C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required ...

  12. ASTRID sodium cooled fast reactor: Program for improving in service inspection and repair

    SciTech Connect (OSTI)

    Jadot, F.; De Dinechin, G.; Augem, J. M.; Sibilo, J.

    2011-07-01

    In the frame of the CEA, EDF, AREVA coordinated research program for the development of Generation IV sodium-cooled fast reactors (SFR), the ASTRID project was launched in 2010. For the future prototype, the improvement of in-service inspection and repair (ISI and R) capabilities was identified as a major issue. Following the pluri-annual SFR research program, the ISI and R main R and D axes remain: i) improvement of the primary system conceptual design, ii) development of measurement and inspection techniques (continuous monitoring instrumentation and periodic inspection tools), iii) accessibility and associated robotics, and iv) development and validation of repair processes. Associated ISI and R needs are being defined through an iterative method between designers and instrumentation specialists: adaptation of the Design to ISI and R requirements, fission chamber development, validation of the ultrasonic and chemical transducers, of ultrasonic non destructive simulation, of acoustic surveillance, of laser repair intervention processes, of connected robotic equipment. Moreover, CEA, as leader of the ASTRID Project, is willing to find new contributors, partners or suppliers, in order to get innovative, diversified, exhaustive and efficient solutions. (authors)

  13. Concept of an inherently-safe high temperature gas-cooled reactor

    SciTech Connect (OSTI)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

    2012-06-06

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

  14. Home Cooling Systems | Department of Energy

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

    Fans In many climates, you can use a whole-house fan to meet all or most of your home cooling needs. Evaporative Cooling For homes in dry climates, evaporative cooling or...

  15. Maintenance Guide for Greenhouse Ventilation, Evaporative Cooling Heating Systems1

    E-Print Network [OSTI]

    Hill, Jeffrey E.

    AE26 Maintenance Guide for Greenhouse Ventilation, Evaporative Cooling Heating Systems1 D. E and preventive maintenance procedures for ventilation, evaporative cooling and heating systems. Ventilation a ventilation system is not operating properly, the results can be pockets of stagnant air, inadequate cooling

  16. Home Cooling Systems | Department of Energy

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

    Home cooling accounts for 6 percent of the average household's energy use. To help you save money by saving energy, our experts are answering your home cooling questions. |...

  17. A CLASSIFICATION SCHEME FOR THE COMMON PASSIVE AND HYBRID HEATING AND COOLING SYSTEMS

    E-Print Network [OSTI]

    Holtz, Michael J.

    2011-01-01

    AND HYBRID HEATING AND COOLING SYSTEMS Michael J. Holtzmost common passive cooling systems and a representativepassive space heating and cooling systems. It is based upon

  18. Impact of Solar Heat Gain on Radiant Floor Cooling System Design

    E-Print Network [OSTI]

    Feng, Jingjuan Dove; Schiavon, Stefano; Bauman, Fred

    2013-01-01

    Radiant Heating and Cooling Systems, in, 2012. [15] F.Gain on Radiant Floor Cooling System Design. Proceedings ofof radiant floor cooling systems and their associated air

  19. Performance Evaluation for Modular, Scalable Liquid-Rack Cooling Systems in Data Centers

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01

    while modular system cooling kW/ton value exhibited aback into the modular cooling system and passed through thethe ventilation and cooling systems being less than optimal.

  20. Simulations of sizing and comfort improvements for residential forced-air heating and cooling systems

    E-Print Network [OSTI]

    Walker, I.S.; Degenetais, G.; Siegel, J.A.

    2002-01-01

    the effect of heating and cooling system inefficiencies onwith inefficient heating and cooling systems in CaliforniaForced-Air Heating and Cooling Systems May 2002 Walker, I. ,

  1. A computer simulation appraisal of non-residential low energy cooling systems in California

    E-Print Network [OSTI]

    Bourassa, Norman; Haves, Philip; Huang, Joe

    2002-01-01

    of Nonresidential Low Energy Cooling Systems in California-of Nonresidential Low Energy Cooling Systems in Californiaof Nonresidential Low Energy Cooling Systems in California

  2. Performance Evaluation for Modular, Scalable Cooling Systems with Hot Aisle Containment in Data Centers

    E-Print Network [OSTI]

    Adams, Barbara J

    2009-01-01

    on the characteristics of cooling systems and servers Theraised, CRAH cooling system and the modular, scalableINFORMATION ON THE CHARACTERISTICS OF COOLING SYSTEMS AND

  3. Performance Evaluation for a Modular, Scalable Passive Cooling System in Data Centers

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01

    by the passive modular cooling system; and P hydraulic isthe ventilation and cooling systems being less than optimal.modular and scalable cooling systems aim at significantly

  4. Cooling load calculations for radiant systems: are they the same traditional methods?

    E-Print Network [OSTI]

    Bauman, Fred; Feng, Jingjuan Dove; Schiavon, Stefano

    2013-01-01

    of Radiant vs. Air System Cooling Loads Recent simulationsystem, the radiant system cooling Advertisement formerly inthe mea- sured radiant system cooling load. The RTS method

  5. Considerations of Alloy N for Fluoride Salt-Cooled High-Temperature Reactor Applications

    SciTech Connect (OSTI)

    Ren, Weiju [ORNL; Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Holcomb, David Eugene [ORNL

    2011-01-01

    Fluoride Salt-Cooled High-Temperature Reactors (FHRs) are a promising new class of thermal-spectrum nuclear reactors. The reactor structural materials must possess high-temperature strength and chemical compatibility with the liquid fluoride salt as well as with a power cycle fluid such as supercritical water while remaining resistant to residual air within the containment. Alloy N was developed for use with liquid fluoride salts and it possesses adequate strength and chemical compatibility up to about 700 C. A distinctive property of FHRs is that their maximum allowable coolant temperature is restricted by their structural alloy maximum service temperature. As the reactor thermal efficiency directly increases with the maximum coolant temperature, higher temperature resistant alloys are strongly desired. This paper reviews the current status of Alloy N and its relevance to FHRs including its design principles, development history, high temperature strength, environmental resistance, metallurgical stability, component manufacturability, ASME codification status, and reactor service requirements. The review will identify issues and provide guidance for improving the alloy properties or implementing engineering solutions.

  6. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Progress Report for Work Through September 2003, 2nd Annual/8th Quarterly Report

    SciTech Connect (OSTI)

    Philip E. MacDonald

    2003-09-01

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation-IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% vs. about 33% efficiency for current Light Water Reactors, LWRs) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus the need for recirculation and jet pumps, a pressurizer, steam generators, steam separators and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies, LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which is also in use around the world.

  7. Cooling System for the MERIT High-Power Target Experiment

    E-Print Network [OSTI]

    McDonald, Kirk

    Cooling System for the MERIT High-Power Target Experiment Haug F., Pereira H., Silva P., Pezzetti M cryogenic cooling system of novel design permits the transfer of nitrogen by the sole means of differential a free mercury jet inside a normal conducting pulsed 15 T capture solenoid magnet cooled with liquid

  8. Closed Brayton cycle power conversion systems for nuclear reactors :

    SciTech Connect (OSTI)

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.; Sanchez, Travis

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at the manufacturers site (Barber-Nichols Inc.) and installed and operated at Sandia. A sufficiently detailed description of the loop is provided in this report along with the design characteristics of the turbo-alternator-compressor set to allow other researchers to compare their results with those measured in the Sandia test-loop. The third task consisted of a validation effort. In this task the test loop was operated and compared with the modeled results to develop a more complete understanding of this electrically heated closed power generation system and to validate the model. The measured and predicted system temperatures and pressures are in good agreement, indicating that the model is a reasonable representation of the test loop. Typical deviations between the model and the hardware results are less than 10%. Additional tests were performed to assess the capability of the Brayton engine to continue to remove decay heat after the reactor/heater is shutdown, to develop safe and effective control strategies, and to access the effectiveness of gas inventory control as an alternative means to provide load following. In one test the heater power was turned off to simulate a rapid reactor shutdown, and the turbomachinery was driven solely by the sensible heat stored in the heater for over 71 minutes without external power input. This is an important safety feature for CBC systems as it means that the closed Brayton loop will keep cooling the reactor without the need for auxiliary power (other than that needed to circulate the waste heat rejection coolant) provided the heat sink is available.

  9. The Thermodynamic and Cost Benefits of Floating Cooling Systems 

    E-Print Network [OSTI]

    Svoboda, K. J.; Klooster, H. J.; Johnnie, D. H., Jr.

    1983-01-01

    Historically, a fixed cooling concept is used in the design of evaporative heat rejection systems for process and power plants. In the fixed cooling mode, a plant is designed for maximum output at the design summer wet bulb temperature...

  10. Potential of Evaporative Cooling Systems for Buildings in India 

    E-Print Network [OSTI]

    Maiya, M. P.; Vijay, S.

    2010-01-01

    psychrometric chart for different cities in India like Ahmadabad, Jodhpur, Nagpur and New Delhi representing different climatic conditions of India. While satisfactorily comfort can be achieved at cool and dry weather conditions by evaporative cooling system...

  11. A Free Cooling Based Chilled Water System at Kingston 

    E-Print Network [OSTI]

    Jansen, P. R.

    1984-01-01

    In efforts to reduce operating costs, the IBM site at Kingston, New York incorporated the energy saving concept of 'free cooling' (direct cooling of chilled water with condenser water) with the expansion of the site chilled water system. Free...

  12. CCHP System with Interconnecting Cooling and Heating Network 

    E-Print Network [OSTI]

    Fu, L.; Geng, K.; Zheng, Z.; Jiang, Y.

    2006-01-01

    The consistency between building heating load, cooling load and power load are analyzed in this paper. The problem of energy waste and low equipment usage in a traditional CCHP (combined cooling, heating and power) system with generated electricity...

  13. Improving the Water Efficiency of Cooling Production System 

    E-Print Network [OSTI]

    Maheshwari, G.; Al-Hadban, Y.; Al-Taqi, H. H.; Alasseri, R.

    2010-01-01

    For most of the time, cooling towers (CTs) of cooling systems operate under partial load conditions and by regulating the air circulation with a variable frequency drive (VFD), significant reduction in the fan power can be achieved. In Kuwait...

  14. NASA's Marshall Space Flight Center Improves Cooling System Performance

    Broader source: Energy.gov [DOE]

    Case study details Marshall Space Flight Center's innovative technologies to improve water efficiency and cooling performance for one of its problematic cooling systems. The program saved the facility more than 800,000 gallons of water in eight months.

  15. Integrated exhaust gas recirculation and charge cooling system

    DOE Patents [OSTI]

    Wu, Ko-Jen

    2013-12-10

    An intake system for an internal combustion engine comprises an exhaust driven turbocharger configured to deliver compressed intake charge, comprising exhaust gas from the exhaust system and ambient air, through an intake charge conduit and to cylinders of the internal combustion engine. An intake charge cooler is in fluid communication with the intake charge conduit. A cooling system, independent of the cooling system for the internal combustion engine, is in fluid communication with the intake charge cooler through a cooling system conduit. A coolant pump delivers a low temperature cooling medium from the cooling system to and through the intake charge cooler for the transfer of heat from the compressed intake charge thereto. A low temperature cooler receives the heated cooling medium through the cooling system conduit for the transfer or heat therefrom.

  16. An integrated performance model for high temperature gas cooled reactor coated particle fuel

    E-Print Network [OSTI]

    Wang, Jing, 1976-

    2004-01-01

    The performance of coated fuel particles is essential for the development and deployment of High Temperature Gas Reactor (HTGR) systems for future power generation. Fuel performance modeling is indispensable for understanding ...

  17. Thermal hydraulic design and analysis of a large lead-cooled reactor with flexible conversion ratio

    E-Print Network [OSTI]

    Nikiforova, Anna S., S.M. Massachusetts Institute of Technology

    2008-01-01

    This thesis contributes to the Flexible Conversion Ratio Fast Reactor Systems Evaluation Project, a part of the Nuclear Cycle Technology and Policy Program funded by the Department of Energy through the Nuclear Energy ...

  18. Polk power station syngas cooling system

    SciTech Connect (OSTI)

    Jenkins, S.D.

    1995-01-01

    Tampa Electric Company (TEC) is in the site development and construction phase of the new Polk Power Station Unit No. 1. This will be the first unit at a new site and will use Integrated Gasification Combined Cycle (IGCC) Technology. The unit will utilize Texaco`s oxygen-blown, entrained-flow coal gasification, along with combined cycle power generation, to produce nominal 260MW. Integral to the gasification process is the syngas cooling system. The design, integration, fabrication, transportation, and erection of this equipment have provided and continue to provide major challenges for this project.

  19. Self-actuating reactor shutdown system

    DOE Patents [OSTI]

    Barrus, Donald M. (San Jose, CA); Brummond, Willian A (Livermore, CA); Peterson, Leslie F. (Danville, CA)

    1988-01-01

    A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

  20. Cedarville School District Retrofit of Heating and Cooling Systems...

    Open Energy Info (EERE)

    - Remove unusable antiquated existing equipment and systems from the campus heating and cooling system, but utilize ductwork, piping, etc. where feasible. The campus is served by...

  1. Economizer refrigeration cycle space heating and cooling system and process

    DOE Patents [OSTI]

    Jardine, Douglas M. (Colorado Springs, CO)

    1983-01-01

    This invention relates to heating and cooling systems and more particularly to an improved system utilizing a Stirling Cycle engine heat pump in a refrigeration cycle.

  2. Economizer refrigeration cycle space heating and cooling system and process

    DOE Patents [OSTI]

    Jardine, D.M.

    1983-03-22

    This invention relates to heating and cooling systems and more particularly to an improved system utilizing a Stirling Cycle engine heat pump in a refrigeration cycle. 18 figs.

  3. Magnitude and reactivity consequences of moisture ingress into the modular High-Temperature Gas-Cooled Reactor core

    SciTech Connect (OSTI)

    Smith, O.L. (Oak Ridge National Lab., TN (United States))

    1992-12-01

    Inadvertent admission of moisture into the primary system of a modular high-temperature gas-cooled reactor has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moisture-ingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. The rate and magnitude of steam buildup are found to be dominated by major system features such as break size compared with safety valve capacity and reliability and less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions.

  4. A CLASSIFICATION SCHEME FOR THE COMMON PASSIVE AND HYBRID HEATING AND COOLING SYSTEMS

    E-Print Network [OSTI]

    Holtz, Michael J.

    2011-01-01

    Common Passive and Hybrid Heating Cooling Systems Michael].THE COMMON PASSIVE AND HYBRID HEATING AND COOLING SYSTEMS

  5. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    SciTech Connect (OSTI)

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  6. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat

    SciTech Connect (OSTI)

    Forsberg, Charles; Hu, Lin-wen; Peterson, Per; Sridharan, Kumar

    2015-01-21

    In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.

  7. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    SciTech Connect (OSTI)

    Mcwilliams, A. J.

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  8. Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.

    SciTech Connect (OSTI)

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal resistance of a gas-filled gap.

  9. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    SciTech Connect (OSTI)

    Larry Demick

    2010-08-01

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

  10. Ice Thermal Storage Systems for Nuclear Power Plant Supplemental Cooling and Peak Power Shifting

    SciTech Connect (OSTI)

    Haihua Zhao; Hongbin Zhang; Phil Sharpe; Blaise Hamanaka; Wei Yan; WoonSeong Jeong

    2013-03-01

    Availability of cooling water has been one of the major issues for the nuclear power plant site selection. Cooling water issues have frequently disrupted the normal operation at some nuclear power plants during heat waves and long draught. One potential solution is to use ice thermal storage (ITS) systems that reduce cooling water requirements and boost the plant’s thermal efficiency in hot hours. ITS uses cheap off-peak electricity to make ice and uses the ice for supplemental cooling during peak demand time. ITS also provides a way to shift a large amount of electricity from off peak time to peak time. For once-through cooling plants near a limited water body, adding ITS can bring significant economic benefits and avoid forced derating and shutdown during extremely hot weather. For the new plants using dry cooling towers, adding the ITS systems can effectively reduce the efficiency loss during hot weather so that new plants could be considered in regions lack of cooling water. This paper will review light water reactor cooling issues and present the feasibility study results.

  11. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    SciTech Connect (OSTI)

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  12. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    A. W. HUNE. “Accelerator Driven Systems for Transmutation:for use in accelerator-driven systems (ADS) as they can beAccelerator-driven B&Bs Terrapower LLC Commercial sodium-cooled SWR designs Figure 2.2: The research history of B&B systems

  13. Phasing of Debuncher Stochastic Cooling Transverse Systems

    SciTech Connect (OSTI)

    Pasquinelli, Ralph; /Fermilab

    2000-03-09

    With the higher frequency of the cooling systems in the Debuncher, a modified method of making transfer functions has been developed for transverse systems. (Measuring of the momentum systems is unchanged.) Speed in making the measurements is critical, as the beam tends to decelerate due to vacuum lifetime. In the 4-8 GHz band, the harmonics in the Debuncher are 6,700 to 13,400 times the revolution frequency. Every Hertz change in revolution frequency is multiplied by this harmonic number and becomes a frequency measurement error, which is an appreciable percent of the momentum width of the beam. It was originally thought that a momentum cooling system would be phased first so that the beam could be kept from drifting in revolution frequency. As it turned out, the momentum cooling was so effective (even with the gain turned down) that the momentum width normalized to fo became less than one Hertz on the Schottky pickup. A beam this narrow requires very precise measurement of tune and revolution frequency. It was difficult to get repeatable results. For initial measuring of the transverse arrays, relative phase and delay is all that is required, so the measurement settings outlined below will suffice. Once all input and output arrays are phased, a more precise measurement of all pickups to all kickers can be done with more points and both upper and lower side bands, as in figure 1. Settings on the network analyzer were adjusted for maximum measurement speed. Data is not analyzed until a complete set of measurements is taken. Start and stop frequencies should be chosen to be just slightly wider than the band being measured. For transverse systems, select betatron USB for the measurement type. This will make the measurement two times faster. Select 101 for the number of points, sweep time of 5 seconds, IF bandwidth 30 Hz, averages = 1. It is important during the phasing to continually measure the revolution frequency and beam width of the beam for transverse systems. Beam width is defined as the 3 dB bandwidth of the momentum Schottky divided by 127 (the harmonic of the Schottky pickup in the Debuncher.) Every three to five minutes, the beam drifts enough to make a significant change in the data. Knowing the revolution frequency and beam width to 0.5 Hz is important. If the beam width exceeds 10 Hz, the quality of the measurement will be impaired. Large beam widths can be caused by excessive forward proton beam current. There are also signs that the front-end amplifiers saturate with beam currents above several hundred microamps. The cooling systems were designed to be very sensitive, (that's why the front end is at liquid helium temperature) so a hundred microamps will go a long way. It should be possible to phase the systems with Pbars as a signal to noise ratio of 30 dB was observed with 100 microamps of beam current.

  14. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    SciTech Connect (OSTI)

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  15. Scanning tunneling microscope assembly, reactor, and system

    DOE Patents [OSTI]

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  16. High strength and heat resistant chromium steels for sodium-cooled fast reactors.

    SciTech Connect (OSTI)

    Kamal, S.; Grandy, C.; Farmer, M.; Brunsvold, A.

    2004-12-22

    This report provides the results of a preliminary phase of a project supporting the Advanced Nuclear Fuel Cycle Technology Initiative at ANL. The project targets the Generation IV nuclear energy systems, particularly the area of reducing the cost of sodium-cooled fast-reactors by utilizing innovative materials. The main goal of the project is to provide the nuclear heat exchanger designers a simplified means to quantify the cost advantages of the recently developed high strength and heat resistant ferritic steels with 9 to 13% chromium content. The emphasis in the preliminary phase is on two steels that show distinctive advantages and have been proposed as candidate materials for heat exchangers and also for reactor vessels and near-core components of Gen IV reactors. These steels are the 12Cr-2W (HCM12A) and 9Cr-1MoVNb (modified 9Cr-1Mo). When these steels are in tube form, they are referred to in ASTM Standards as T122 and T91, respectively. A simple thermal-hydraulics analytical model of a counter-flow, shell-and-tube, once-through type superheated steam generator is developed to determine the required tube length and tube wall temperature profile. The single-tube model calculations are then extended to cover the following design criteria: (i) ratio of the tube stress due to water/steam pressure to the ASME B&PV Code allowable stress, (ii) ratio of the strain due to through-tube-wall temperature differences to the material fatigue limit, (iii) overall differential thermal expansion between the tube and shell, and (iv) total amount of tube material required for the specified heat exchanger thermal power. Calculations were done for a 292 MW steam generator design with 2200 tubes and a steam exit condition of 457 C and 16 MPa. The calculations were performed with the tubes made of the two advanced ferritic steels, 12Cr-2W and 9Cr-1MoVNb, and of the most commonly used steel, 2 1/4Cr-1Mo. Compared to the 2 1/4Cr-1Mo results, the 12Cr-2W tubes required 29% less material and the 9Cr-1MoVNb tubes required 25% less material. Also, with the advanced steels, the thermal strains in the tubes and differential thermal expansion between tubes and shell were significantly better. For steam generators with higher steam exit temperatures, the benefits of the advanced steels become much larger. A thorough search for the thermal and mechanical properties of the two advanced steels was conducted. A summary of the search results is provided. It shows what is presently known about these two advanced steels and what still needs to be determined so that they can be used in nuclear heat exchanger designs. Possible follow up steps are outlined.

  17. A FEASIBILITY AND OPTIMIZATION STUDY TO DETERMINE COOLING TIME AND BURNUP OF ADVANCED TEST REACTOR FUELS USING A NONDESTRUCTIVE TECHNIQUE

    SciTech Connect (OSTI)

    Jorge Navarro

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method for ATR applications the technique was tested using one-isotope, multi-isotope and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr3 detector in an above the water configuration and deconvolution algorithms.

  18. Critical review of water based radiant cooling system design methods

    E-Print Network [OSTI]

    Feng, Jingjuan Dove; Bauman, Fred; Schiavon, Stefano

    2014-01-01

    buildings CRITICAL REVIEW OF WATER BASED RADIANT COOLINGare two primary types of water-based radiant systems: (1)cooling/heating output, water supply temperatures Notes NA

  19. Cedarville School District Retrofit of Heating and Cooling Systems...

    Energy Savers [EERE]

    Cedarville School District Retrofit of Heating and Cooling Systems with Geothermal Heat Pumpsand Ground Source Water Loops Cedarville School District Retrofit of Heating and...

  20. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    SciTech Connect (OSTI)

    Sonat Sen; Gilles Youinou

    2013-02-01

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  1. Designing a 'Near Optimum' Cooling-Water System 

    E-Print Network [OSTI]

    Crozier, R. A., Jr.

    1981-01-01

    developed similar procedures for designing and optimizing a cooling-water once through-exchanger system. This article attempts to fill the void by presenting a design basis that will produce a 'near optimum' system. A cooling-water system consists of four...

  2. Temperature and cooling management in computing systems

    E-Print Network [OSTI]

    Ayoub, Raid

    2011-01-01

    78 5.2 Combined Energy, Thermal and Coolingdo not account for energy and thermal optimizations for theWe propose a combined energy, thermal and cooling management

  3. Cooling load calculations for radiant systems: are they the same traditional methods?

    E-Print Network [OSTI]

    Bauman, Fred; Feng, Jingjuan Dove; Schiavon, Stefano

    2013-01-01

    B. 2008. “Radiant floor cooling systems. ” ASHRAE Journal 4.embedded radiant heating and cooling. Geneva: InternationalM. Deru. 2010. “Radiant slab cooling for retail. ” ASHRAE

  4. Principles of passive and active cooling of mirror-based hybrid systems employing liquid metals

    SciTech Connect (OSTI)

    Anglart, Henryk [Div. of Nuclear Technology, School of Engineering Sciences, Royal Institute of Technology Roslagstullsbacken 21, 106-91 Stockholm (Sweden)

    2012-06-19

    This paper presents principles of passive and active cooling that are suitable to mirrorbased hybrid, nuclear fission/fusion systems. It is shown that liquid metal lead-bismuth cooling of the mirror machine with 25 m height and 1.5 GW thermal power is feasible both in the active mode during the normal operation and in the passive mode after the reactor shutdown. In the active mode the achievable required pumping power can well be below 50 MW, whereas the passive mode provides enough coolant flow to keep the clad temperature below the damage limits.

  5. Cooling Water Systems - Energy Savings/Lower Costs By Reusing Cooling Tower Blowdown 

    E-Print Network [OSTI]

    Puckorius, P. R.

    1981-01-01

    down for reuse into the cooling tower system. Several plants have been built and operated with considerable difficulty regarding effective operation of the softener due to improper chemical selection. However, other plants have utilized the proper...

  6. Optimized Design of a Furnace Cooling System 

    E-Print Network [OSTI]

    Morelli, F.; Bretschneider, R.; Dauzat, J.; Guymon, M.; Studebaker, J.; Rasmussen, B. P.

    2013-01-01

    This paper presents a case study of manufacturing furnace optimized re-design. The bottleneck in the production process is the cooling of heat treatment furnaces. These ovens are on an approximate 24-hour cycle, heating for 12 hours and cooling...

  7. Debris trap in a turbine cooling system

    DOE Patents [OSTI]

    Wilson, Ian David (Clifton Park, NY)

    2002-01-01

    In a turbine having a rotor and a plurality of stages, each stage comprising a row of buckets mounted on the rotor for rotation therewith; and wherein the buckets of at least one of the stages are cooled by steam, the improvement comprising at least one axially extending cooling steam supply conduit communicating with an at least partially annular steam supply manifold; one or more axially extending cooling steam feed tubes connected to the manifold at a location radially outwardly of the cooling steam supply conduit, the feed tubes arranged to supply cooling steam to the buckets of at least one of the plurality of stages; the manifold extending radially beyond the feed tubes to thereby create a debris trap region for collecting debris under centrifugal loading caused by rotation of the rotor.

  8. Control system for a small fission reactor

    DOE Patents [OSTI]

    Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

    1985-02-08

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

  9. Absolute Dynamical Limit to Cooling Weakly-Coupled Quantum Systems

    E-Print Network [OSTI]

    X. Wang; Sai Vinjanampathy; Frederick W. Strauch; Kurt Jacobs

    2012-05-15

    Cooling of a quantum system is limited by the size of the control forces that are available (the "speed" of control). We consider the most general cooling process, albeit restricted to the regime in which the thermodynamics of the system is preserved (weak coupling). Within this regime, we further focus on the most useful control regime, in which a large cooling factor, and good ground-state cooling can be achieved. We present a control protocol for cooling, and give clear structural arguments, as well as strong numerical evidence, that this protocol is globally optimal. From this we obtain simple expressions for the limit to cooling that is imposed by the speed of control.

  10. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    SciTech Connect (OSTI)

    Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik; Aziz, Ferhat; Sekimoto, Hiroshi

    2014-02-12

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  11. NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions

    SciTech Connect (OSTI)

    Phillip Mills

    2012-02-01

    This document is intended to provide a Next Generation Nuclear Plant (NGNP) Project tool in which to collect and identify key definitions, plant capabilities, and inputs and assumptions to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor (HTGR). These definitions, capabilities, and assumptions are extracted from a number of sources, including NGNP Project documents such as licensing related white papers [References 1-11] and previously issued requirement documents [References 13-15]. Also included is information agreed upon by the NGNP Regulatory Affairs group's Licensing Working Group and Configuration Council. The NGNP Project approach to licensing an HTGR plant via a combined license (COL) is defined within the referenced white papers and reference [12], and is not duplicated here.

  12. High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

    SciTech Connect (OSTI)

    Larry Demick

    2011-08-01

    This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.

  13. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  14. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L. (Albuquerque, NM); Williams, Brian E. (Pacoima, CA); Benander, Robert E. (Pacoima, CA)

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  15. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L. (Albuquerque, NM); Williams, Brian E. (Pocoima, CA); Benander, Robert E. (Pacoima, CA)

    2010-02-23

    Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

  16. Safety design approach for external events in Japan sodium-cooled fast reactor

    SciTech Connect (OSTI)

    Yamano, H.; Kubo, S. [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki, 311-1393 (Japan); Tani, A. [Mitsubishi FBR Systems, Inc., 2-34-17, Jingumae, Shibuya-ku, Tokyo, 150-0001 (Japan); Nishino, H.; Sakai, T. [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki, 311-1393 (Japan)

    2012-07-01

    This paper describes a safety design approach for external events in the design study of Japan sodium-cooled fast reactor. An emphasis is introduction of a design extension external condition (DEEC). In addition to seismic design, other external events such as tsunami, strong wind, abnormal temperature, etc. were addressed in this study. From a wide variety of external events consisting of natural hazards and human-induced ones, a screening method was developed in terms of siting, consequence, frequency to select representative events. Design approaches for these events were categorized on the probabilistic, statistical and deterministic basis. External hazard conditions were considered mainly for DEECs. In the probabilistic approach, the DEECs of earthquake, tsunami and strong wind were defined as 1/10 of exceedance probability of the external design bases. The other representative DEECs were also defined based on statistical or deterministic approaches. (authors)

  17. Ventilation Systems for Cooling | Department of Energy

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

    ventilation can help keep your home cool during hot days. To avoid heat buildup in your home, plan ahead by landscaping your lot to shade your house. If you replace your roof,...

  18. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    in particular, in boiling water reactors) there is a strongdirect contact boiling water fast reactor (PBWFR)”. In:

  19. Heat pump system with selective space cooling

    DOE Patents [OSTI]

    Pendergrass, Joseph C. (Gainesville, GA)

    1997-01-01

    A reversible heat pump provides multiple heating and cooling modes and includes a compressor, an evaporator and heat exchanger all interconnected and charged with refrigerant fluid. The heat exchanger includes tanks connected in series to the water supply and a condenser feed line with heat transfer sections connected in counterflow relationship. The heat pump has an accumulator and suction line for the refrigerant fluid upstream of the compressor. Sub-cool transfer tubes associated with the accumulator/suction line reclaim a portion of the heat from the heat exchanger. A reversing valve switches between heating/cooling modes. A first bypass is operative to direct the refrigerant fluid around the sub-cool transfer tubes in the space cooling only mode and during which an expansion valve is utilized upstream of the evaporator/indoor coil. A second bypass is provided around the expansion valve. A programmable microprocessor activates the first bypass in the cooling only mode and deactivates the second bypass, and vice-versa in the multiple heating modes for said heat exchanger. In the heating modes, the evaporator may include an auxiliary outdoor coil for direct supplemental heat dissipation into ambient air. In the multiple heating modes, the condensed refrigerant fluid is regulated by a flow control valve.

  20. Heat pump system with selective space cooling

    DOE Patents [OSTI]

    Pendergrass, J.C.

    1997-05-13

    A reversible heat pump provides multiple heating and cooling modes and includes a compressor, an evaporator and heat exchanger all interconnected and charged with refrigerant fluid. The heat exchanger includes tanks connected in series to the water supply and a condenser feed line with heat transfer sections connected in counterflow relationship. The heat pump has an accumulator and suction line for the refrigerant fluid upstream of the compressor. Sub-cool transfer tubes associated with the accumulator/suction line reclaim a portion of the heat from the heat exchanger. A reversing valve switches between heating/cooling modes. A first bypass is operative to direct the refrigerant fluid around the sub-cool transfer tubes in the space cooling only mode and during which an expansion valve is utilized upstream of the evaporator/indoor coil. A second bypass is provided around the expansion valve. A programmable microprocessor activates the first bypass in the cooling only mode and deactivates the second bypass, and vice-versa in the multiple heating modes for said heat exchanger. In the heating modes, the evaporator may include an auxiliary outdoor coil for direct supplemental heat dissipation into ambient air. In the multiple heating modes, the condensed refrigerant fluid is regulated by a flow control valve. 4 figs.

  1. Understanding and reducing energy and costs in industrial cooling systems 

    E-Print Network [OSTI]

    Muller, M.R.; Muller, M.B.

    2012-01-01

    demonstrates the large amount of increased saving from a critical review of plant chilled water systems with both hardware and operational improvements. After showing several reasons why cooling systems are often ignored during plant energy surveys (their...

  2. Cooling system early-stage design tool for naval applications

    E-Print Network [OSTI]

    Fiedel, Ethan R

    2011-01-01

    This thesis utilizes concepts taken from the NAVSEA Design Practices and Criteria Manualfor Surface Ship Freshwater Systems and other references to create a Cooling System Design Tool (CSDT). With the development of new ...

  3. A Lithium Self-Cooled Blanket for the HAPL Conceptual Inertial Confinement Reactor

    SciTech Connect (OSTI)

    Sviatoslavsky, I.N. [University of Wisconsin-Madison (United States); Raffray, A.R. [University of California, San Diego (United States); Sawan, M.E. [University of Wisconsin-Madison (United States); Wang, X. [University of California, San Diego (United States)

    2005-04-15

    A multi-institutional study HAPL (High Average Power Laser) is investigating a relatively near term conceptual design of a laser driven inertial confinement reactor. A primary focus of the study is the protection of the first wall (FW) from the target emanations. This paper gives a brief analysis of one of several possible blankets that can be integrated with the chosen FW protection scheme. The structural material is conventional ferritic steel (FS) F82H cooled with liquid lithium. The maximum average temperature is constrained to 550 deg. C. The chamber radius is 6.5 m at midplane, tapering to 2.5 m at the ends, and is surrounded by a cylindrical vacuum vessel. The first wall (FW) is 0.35 cm FS, which has a 0.1 cm thick layer of tungsten bonded to it facing the target. The FW is cooled with Li admitted at the bottom of the blanket, flows through a gap between 0.25-0.5 cm to the top, then returns through the center of the blanket channel to the bottom. There are 60 laser beam ports situated around the chamber. The tritium breeding ratio (TBR) is 1.124. A Brayton cycle is envisaged with an efficiency in the range of 42-44%.

  4. Development of an internally cooled annular fuel bundle for pressurized heavy water reactors

    SciTech Connect (OSTI)

    Hamilton, H.; Armstrong, J.; Kittmer, A.; Zhuchkova, A.; Xu, R.; Hyland, B.; King, M.; Nava-Dominguez, A.; Livingstone, S.; Bergeron, A. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

    2013-07-01

    A number of preliminary studies have been conducted at Atomic Energy of Canada Limited to explore the potential of using internally cooled annular fuel (ICAF) in CANDU reactors including finite element thermo-mechanical modelling, reactor physics, thermal hydraulics, fabrication and mechanical design. The most compelling argument for this design compared to the conventional solid-rod design is the significant reduction in maximum fuel temperature for equivalent LERs (linear element ratings). This feature presents the potential for power up-rating or higher burnup and a decreased defect probability due to in-core power increases. The thermal-mechanical evaluation confirmed the significant reduction in maximum fuel temperatures for ICAF fuel compared to solid-rod fuel for equivalent LER. The maximum fuel temperature increase as a function of LER increase is also significantly less for ICAF fuel. As a result, the sheath stress induced by an equivalent power increase is approximately six times less for ICAF fuel than solid-rod fuel. This suggests that the power-increase thresholds to failure (due to stress-corrosion cracking) for ICAF fuel should be well above those for solid-rod fuel, providing improvement in operation flexibility and safety.

  5. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    SciTech Connect (OSTI)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  6. Closed loop air cooling system for combustion turbines

    DOE Patents [OSTI]

    Huber, David John (North Canton, OH); Briesch, Michael Scot (Orlando, FL)

    1998-01-01

    Convective cooling of turbine hot parts using a closed loop system is disclosed. Preferably, the present invention is applied to cooling the hot parts of combustion turbine power plants, and the cooling provided permits an increase in the inlet temperature and the concomitant benefits of increased efficiency and output. In preferred embodiments, methods and apparatus are disclosed wherein air is removed from the combustion turbine compressor and delivered to passages internal to one or more of a combustor and turbine hot parts. The air cools the combustor and turbine hot parts via convection and heat is transferred through the surfaces of the combustor and turbine hot parts.

  7. Compact Solid State Cooling Systems: Compact MEMS Electrocaloric Module

    SciTech Connect (OSTI)

    None

    2010-10-01

    BEETIT Project: UCLA is developing a novel solid-state cooling technology to translate a recent scientific discovery of the so-called giant electrocaloric effect into commercially viable compact cooling systems. Traditional air conditioners use noisy, vapor compression systems that include a polluting liquid refrigerant to circulate within the air conditioner, absorb heat, and pump the heat out into the environment. Electrocaloric materials achieve the same result by heating up when placed within an electric field and cooling down when removed—effectively pumping heat out from a cooler to warmer environment. This electrocaloric-based solid state cooling system is quiet and does not use liquid refrigerants. The innovation includes developing nano-structured materials and reliable interfaces for heat exchange. With these innovations and advances in micro/nano-scale manufacturing technologies pioneered by semiconductor companies, UCLA is aiming to extend the performance/reliability of the cooling module.

  8. A COOLING SYSTEM FOR BUIDINGS USING WIND ENERGY

    E-Print Network [OSTI]

    A COOLING SYSTEM FOR BUIDINGS USING WIND ENERGY Hamid Daiyan Islamic Azad University - Semnan Branch, Iran hamid.daiyan@semnaniau.ac.ir Abstract In Iranian historical architecture wind tower is used for cooling and ventilation. Wind tower is a tall structure that stands on building. Wind tower is used

  9. Hybrid Molten Salt Reactor (HMSR) System Study

    SciTech Connect (OSTI)

    Woolley, Robert D; Miller, Laurence F

    2014-04-01

    Can the hybrid system combination of (1) a critical fission Molten Salt Reactor (MSR) having a thermal spectrum and a high Conversion Ratio (CR) with (2) an external source of high energy neutrons provide an attractive solution to the world's expanding demand for energy? The present study indicates the answer is an emphatic yes.

  10. Microchannel Reactor System for Catalytic Hydrogenation

    SciTech Connect (OSTI)

    2004-07-01

    This factsheet describes a research project whose goal is to design, fabricate, evaluate, and optimize a laboratory-scale microchannel reactor/heat exchanger system with thin-film or particulate catalysts for hydrogenation of o-nitroanisole and other nitro aromatic compounds, under moderate temperature and pressure.

  11. Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts

    SciTech Connect (OSTI)

    Ronald Farris; David Gertman; Jacques Hugo

    2014-03-01

    This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was to develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that would provide sophisticated operational information visualization, coupled with adaptive automation schemes and operator support systems to reduce complexity. These all have to be mapped at some point to human performance requirements. The EBR-II results will be used as a baseline that will be extrapolated in the extended Cognitive Work Analysis phase to the analysis of a selected advanced sodium-cooled SMR design as a way to establish non-conventional operational concepts. The Work Domain Analysis results achieved during this phase have not only established an organizing and analytical framework for describing existing sociotechnical systems, but have also indicated that the method is particularly suited to the analysis of prospective and immature designs. The results of the EBR-II Work Domain Analysis have indicated that the methodology is scientifically sound and generalizable to any operating environment.

  12. Feasibility Study of Supercritical Light Water Cooled Reactors for Electrical Power Production, 5th Quarterly Report, October - December 2002

    SciTech Connect (OSTI)

    Philip MacDonald; Jacopo Buongiorno; Cliff Davis; J. Stephen Herring; Kevan Weaver; Ron Latanision; Bryce Mitton; Gary Was; Luca Oriani; Mario Carelli; Dmitry Paramonov; Lawrence Conway

    2003-01-01

    The overall objective of this project is to evaluate the feasibility of supercritical light water cooled reactors for electric power production. The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies for the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR that can also burn actinides. The project is organized into three tasks:

  13. Rodded shutdown system for a nuclear reactor

    DOE Patents [OSTI]

    Golden, Martin P. (Penn Township, Allegheny County, PA); Govi, Aldo R. (Greensburg, PA)

    1978-01-01

    A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core.

  14. Adequacy of the Heat-Mass Transfer Analogy to Simulate Containment Atmospheric Cooling in the New Generation of Advanced Nuclear Reactors: Experimental Confirmation

    SciTech Connect (OSTI)

    Herranz, Luis E. [CIEMAT (Spain); Campo, Antonio [Idaho State University (United States)

    2002-09-15

    Driving forces of passive cooling systems of advanced reactor containments are substantially weaker than those brought in by active systems of operating power plants. This fact along with the new geometries being used suggest the need either to develop new reliable simulation techniques or to adapt and validate traditional approaches. Suitability of the heat-mass transfer analogy for this purpose is investigated based on previous authors' experience. Major analogy drawbacks are identified and overcome by supplementing it with analytically derived factors. By comparing against experimental data available, the heat-mass transfer analogy is demonstrated to be a sound, configuration-independent, and accurate-enough theoretical approximation.

  15. Reactor control rod timing system. [LMFBR

    DOE Patents [OSTI]

    Wu, P.T.K.

    1980-03-18

    A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  16. Vertical Pretreatment Reactor System (Poster)

    SciTech Connect (OSTI)

    Not Available

    2012-09-01

    IBRF poster developed for the IBRF showcase. Describes the two-vessel system for primary and secondary pretreatment of biomass solids at different temperatures.

  17. Fault-tolerant reactor protection system

    DOE Patents [OSTI]

    Gaubatz, D.C.

    1997-04-15

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service. 16 figs.

  18. Fault-tolerant reactor protection system

    DOE Patents [OSTI]

    Gaubatz, Donald C. (Cupertino, CA)

    1997-01-01

    A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service.

  19. Comparison of Zone Cooling Load for Radiant and All-Air Conditioning Systems

    E-Print Network [OSTI]

    Feng, Jingjuan; Schiavon, Stefano; Bauman, Fred

    2012-01-01

    Radiant Heating and Cooling Systems. Olesen, B. (2012). "surface heating and cooling systems: . Brussels, EuropeanIn the case of the all-air system, cooling load matches the

  20. A CLASSIFICATION SCHEME FOR THE COMMON PASSIVE AND HYBRID HEATING AND COOLING SYSTEMS

    E-Print Network [OSTI]

    Holtz, Michael J.

    2011-01-01

    Passive and Hybrid Heating Cooling Systems Michael]. Holtz,PASSIVE AND HYBRID HEATING AND COOLING SYSTEMS Michael J.of passive and hybrid space heating and cooling systems are

  1. March 1, 2013. Campus Wide District Heating & Cooling System

    E-Print Network [OSTI]

    ____________________________ March 1, 2013. Campus Wide District Heating & Cooling System. Today · Decentralisation of the heating plant · Introduction of an Energy Loop · Geothermal 4. Results 5 · Decentralisation of the heating plant · Introduction of an Energy Loop · Geothermal 4. Results 5. Tomorrow 6

  2. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    increase for a typical sodium fast reactor fuel rod geometryof the new Russian sodium fast reactor BN-800 [111]. Thethe strong focus of sodium fast reactor research to avoid

  3. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    AND-BURN REACTOR PHYSICS wave burnup principle. The CANDLEand physical principle Breed-and-burn reactors (B&B) areBURN REACTOR PHYSICS The FIMA burnup unit - principles and

  4. System and method for pre-cooling of buildings

    DOE Patents [OSTI]

    Springer, David A.; Rainer, Leo I.

    2011-08-09

    A method for nighttime pre-cooling of a building comprising inputting one or more user settings, lowering the indoor temperature reading of the building during nighttime by operating an outside air ventilation system followed, if necessary, by a vapor compression cooling system. The method provides for nighttime pre-cooling of a building that maintains indoor temperatures within a comfort range based on the user input settings, calculated operational settings, and predictions of indoor and outdoor temperature trends for a future period of time such as the next day.

  5. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Peterson, Per

    2012-10-30

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of numerical models were developed in parallel to the experimental work. RELAP5-3D models were developed for the salt-cooled PB-AHTR, and for the simulat fluid CIET natural circulation experimental loop. These models are to be validated by the data collected from CIET. COMSOL finite element models were used to predict the temperature and fluid flow distribution in the annular pebble bed core; they were instrumental for design of SETs, and they can be used for code-to-code comparisons with RELAP5-3D. A number of other small SETs, and numerical models were constructed, as needed, in support of this work. The experiments were designed, constructed and performed to meet CAES quality assurance requirements for test planning, implementation, and documentation; equipment calibration and documentation, procurement document control; training and personnel qualification; analysis/modeling software verification and validation; data acquisition/collection and analysis; and peer review.

  6. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    SciTech Connect (OSTI)

    Memmott, M. J.; Stansbury, C.; Taylor, C.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  7. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    SciTech Connect (OSTI)

    Qualls, A.L.; Cetiner, M.S.; Wilson, T.L., Jr.

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary system relate to flows within the reactor vessel during severe events and the resulting temperature profiles (temperature and duration) for major components. Critical components include the fuel, reactor vessel, primary piping, and the primary-to-intermediate heat exchangers (P-IHXs). The major AHTR power system loops are shown in Fig. 3. The intermediate heat transfer system is a group of three pumped salt loops that transports the energy produced in the primary system to the power conversion system. Two dynamic system models are used to analyze the AHTR. A Matlab/Simulink?-based model initiated in 2011 has been updated to reflect the evolving design parameters related to the heat flows associated with the reactor vessel. The Matlab model utilizes simplified flow assumptions within the vessel and incorporates an empirical representation of the Direct Reactor Auxiliary Cooling System (DRACS). A Dymola/Modelica? model incorporates a more sophisticated representation of primary coolant flow and a physics-based representation of the three-loop DRACS thermal hydraulics. This model is not currently operating in a fully integrated mode. The Matlab model serves as a prototype and provides verification for the Dymola model, and its use will be phased out as the Dymola model nears completion. The heat exchangers in the system are sized using spreadsheet-based, steady-state calculations. The detail features of the heat exchangers are programmed into the dynamic models, and the overall dimensions are used to generate realistic plant designs. For the modeling cases where the emphasis is on understanding responses within the intermediate and primary systems, the power conversion system may be modeled as a simple boundary condition at the intermediate-to-power conversion system heat exchangers.

  8. Integral reactor system and method for fuel cells

    DOE Patents [OSTI]

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  9. Cavity cooling of an ensemble spin system

    E-Print Network [OSTI]

    Christopher J. Wood; Troy W. Borneman; David G. Cory

    2014-02-24

    We describe how sideband cooling techniques may be applied to large spin ensembles in magnetic resonance. Using the Tavis-Cummings model in the presence of a Rabi drive, we solve a Markovian master equation describing the joint spin-cavity dynamics to derive cooling rates as a function of ensemble size. Our calculations indicate that the coupled angular momentum subspaces of a spin ensemble containing roughly $10^{11}$ electron spins may be polarized in a time many orders of magnitude shorter than the typical thermal relaxation time. The described techniques should permit efficient removal of entropy for spin-based quantum information processors and fast polarization of spin samples. The proposed application of a standard technique in quantum optics to magnetic resonance also serves to reinforce the connection between the two fields, which has recently begun to be explored in further detail due to the development of hybrid designs for manufacturing noise-resilient quantum devices.

  10. Cooling system for three hook ring segment

    DOE Patents [OSTI]

    Campbell, Christian X.; Eng, Darryl; Lee, Ching-Pang; Patat, Harry

    2014-08-26

    A triple hook ring segment including forward, midsection and aft mounting hooks for engagement with respective hangers formed on a ring segment carrier for supporting a ring segment panel, and defining a forward high pressure chamber and an aft low pressure chamber on opposing sides of the midsection mounting hook. An isolation plate is provided on the aft side of the midsection mounting hook to form an isolation chamber between the aft low pressure chamber and the ring segment panel. High pressure air is supplied to the forward chamber and flows to the isolation chamber through crossover passages in the midsection hook. The isolation chamber provides convection cooling air to an aft portion of the ring segment panel and enables a reduction of air pressure in the aft low pressure chamber to reduce leakage flow of cooling air from the ring segment.

  11. Double dark state cooling in a three level system

    E-Print Network [OSTI]

    Javier Cerrillo; Alex Retzker; Martin B. Plenio

    2011-12-29

    A detailed study of a robust and fast laser cooling scheme on a three level system is presented. A special laser configuration, applicable to trapped ions, atoms or cantilevers, designs a quantum interference that eliminates the blue sideband in addition to the carrier transition, thus excluding any heating process involving up to one-phonon processes. As a consequence cooling achieves vanishing phonon occupation up to first order in the Lamb-Dicke parameter expansion. Underlying this scheme is a combined action of two cooling schemes which makes the proposal very stable under fluctuations of the physical parameters such as laser intensity or detuning, making it a viable candidate for experimental implementation. Furthermore, it is considerably faster than existing ground state cooling schemes, overcoming one of the limitations of current quantum information processing implementations. Its suitability as a cooling scheme for several ions in a trap or for a cloud of atoms in a dipole trap is shown.

  12. Active noise canceling system for mechanically cooled germanium radiation detectors

    DOE Patents [OSTI]

    Nelson, Karl Einar; Burks, Morgan T

    2014-04-22

    A microphonics noise cancellation system and method for improving the energy resolution for mechanically cooled high-purity Germanium (HPGe) detector systems. A classical adaptive noise canceling digital processing system using an adaptive predictor is used in an MCA to attenuate the microphonics noise source making the system more deployable.

  13. Challenges in the Development of Advanced Reactors

    SciTech Connect (OSTI)

    P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

    2012-08-01

    Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

  14. Microchannel Reactor System for Catalytic Hydrogenation

    SciTech Connect (OSTI)

    Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie

    2010-12-22

    We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstrated on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.

  15. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  16. Plasma generators, reactor systems and related methods

    DOE Patents [OSTI]

    Kong, Peter C. (Idaho Falls, ID); Pink, Robert J. (Pocatello, ID); Lee, James E. (Idaho Falls, ID)

    2007-06-19

    A plasma generator, reactor and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably coupled together to form a chamber. Associated with each section is an electrode set including three electrodes with each electrode being coupled to a single phase of a three-phase alternating current (AC) power supply. The electrodes are disposed about a longitudinal centerline of the chamber and are arranged to provide and extended arc and generate an extended body of plasma. The electrodes are displaceable relative to the longitudinal centerline of the chamber. A control system may be utilized so as to automatically displace the electrodes and define an electrode gap responsive to measure voltage or current levels of the associated power supply.

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

    SciTech Connect (OSTI)

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

    2005-07-01

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

  18. Staged membrane oxidation reactor system

    DOE Patents [OSTI]

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2012-09-11

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  19. Staged membrane oxidation reactor system

    DOE Patents [OSTI]

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2014-05-20

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  20. Staged membrane oxidation reactor system

    DOE Patents [OSTI]

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2013-04-16

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  1. Space reactor electric systems: system integration studies, Phase 1 report

    SciTech Connect (OSTI)

    Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

    1983-03-29

    This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied.

  2. Rotary engine cooling system with flow balancing

    SciTech Connect (OSTI)

    Jones, C.

    1987-05-12

    This patent describes a rotary internal combustion engine having a trochoid rotor housing section and having a group of cooling passages extending through a top-dead-center (TDC) region. The engine is characterized by: at least one passage of the group following a curved path which extends through a first hotter portion of the TDC region, by at least one further passage of the group following a substantially uncurved path through a second cooler portion of the TDC region, and by a fluid restriction for restricting fluid flow through the at least one further passage to balance coolant flow between the passages.

  3. Application of subgroup decomposition in diffusion theory to gas cooled thermal reactor problem

    SciTech Connect (OSTI)

    Yasseri, S.; Rahnema, F. [Nuclear and Radiological Engineering and Medical Physics Program, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States)

    2013-07-01

    In this paper, the accuracy and computational efficiency of the subgroup decomposition (SGD) method in diffusion theory is assessed in a ID benchmark problem characteristic of gas cooled thermal systems. This method can be viewed as a significant improvement in accuracy of standard coarse-group calculations used for VHTR whole core analysis in which core environmental effect and energy angle coupling are pronounced. It is shown that a 2-group SGD calculation reproduces fine-group (47) results with 1.5 to 6 times faster computational speed depending on the stabilizing schemes while it is as efficient as single standard 6-group diffusion calculation. (authors)

  4. Steam cooling system for a gas turbine

    DOE Patents [OSTI]

    Wilson, Ian David (Mauldin, SC); Barb, Kevin Joseph (Halfmoon, NY); Li, Ming Cheng (Cincinnati, OH); Hyde, Susan Marie (Schenectady, NY); Mashey, Thomas Charles (Coxsackie, NY); Wesorick, Ronald Richard (Albany, NY); Glynn, Christopher Charles (Hamilton, OH); Hemsworth, Martin C. (Cincinnati, OH)

    2002-01-01

    The steam cooling circuit for a gas turbine includes a bore tube assembly supplying steam to circumferentially spaced radial tubes coupled to supply elbows for transitioning the radial steam flow in an axial direction along steam supply tubes adjacent the rim of the rotor. The supply tubes supply steam to circumferentially spaced manifold segments located on the aft side of the 1-2 spacer for supplying steam to the buckets of the first and second stages. Spent return steam from these buckets flows to a plurality of circumferentially spaced return manifold segments disposed on the forward face of the 1-2 spacer. Crossover tubes couple the steam supply from the steam supply manifold segments through the 1-2 spacer to the buckets of the first stage. Crossover tubes through the 1-2 spacer also return steam from the buckets of the second stage to the return manifold segments. Axially extending return tubes convey spent cooling steam from the return manifold segments to radial tubes via return elbows.

  5. System and method for cooling a combustion gas charge

    DOE Patents [OSTI]

    Massey, Mary Cecelia; Boberg, Thomas Earl

    2010-05-25

    The present invention relates to a system and method for cooling a combustion gas charge prior. The combustion gas charge may include compressed intake air, exhaust gas, or a mixture thereof. An evaporator is provided that may then receive a relatively high temperature combustion gas charge and discharge at a relatively lower temperature. The evaporator may be configured to operate with refrigeration cycle components and/or to receive a fluid below atmospheric pressure as the phase-change cooling medium.

  6. American Indian Complex to Cool Off Using Ice Storage System

    Broader source: Energy.gov [DOE]

    In Oklahoma City, summer temperatures can get above 100 degrees, making cooling more of a necessity than a luxury. But the designers of the American Indian Cultural Center and Museum (AICCM) wanted to make cooling choices that reflect American Indian cultures' respect for the land. So, rather than using conventional air-conditioning, the museum's main complex will use an ice storage system estimated to save 644,000 kilowatt hours of electricity a year.

  7. Closed-loop air cooling system for a turbine engine

    DOE Patents [OSTI]

    North, William Edward (Winter Springs, FL)

    2000-01-01

    Method and apparatus are disclosed for providing a closed-loop air cooling system for a turbine engine. The method and apparatus provide for bleeding pressurized air from a gas turbine engine compressor for use in cooling the turbine components. The compressed air is cascaded through the various stages of the turbine. At each stage a portion of the compressed air is returned to the compressor where useful work is recovered.

  8. Intermediate Heat Transfer Loop Study for High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    C. H. Oh; C. Davis; S. Sherman

    2008-08-01

    A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic and cycleefficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. This paper also includes a portion of stress analyses performed on pipe configurations.

  9. Modeling of the Reactor Core Isolation Cooling Reposnse to Beyond Design Basis Operations - Interim Report

    SciTech Connect (OSTI)

    Ross, Kyle; Cardoni, Jeffrey N.; Wilson, Chisom Shawn; Morrow, Charles; Osborn, Douglas; Gauntt, Randall O.

    2015-12-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine decreases the developed turbine torque; the RCIC speed then slows, and thus the pump flow rate to the RPV decreases. Subsequently, RPV water level decreases due to continued boiling and the liquid fraction flowing to the RCIC decreases, thereby accelerating the RCIC and refilling the RPV. The feedback cycle then repeats itself and/or reaches a quasi-steady equilibrium condition. In other words, the water carry-over is limited by cyclic RCIC performance degradation, and hence the system becomes self-regulating. The indications achieved to date with the system model are more qualitative than quantitative. The avenues being pursued to increase the fidelity of the model are expected to add quantitative realism. The end product will be generic in the sense that the RCIC model will be incorporable within the larger reactor coolant system model of any nuclear power plant or experimental configuration. ACKNOWLEDGEMENTS This work is funded through the U.S. Department of Energy - Office of Nuclear Energy's Light Water Reactor Sustainability Program. Clinton Smith of Phoenix Analysis & Design Technologies (PADT) is acknowledged for providing assistance on the FLUENT efforts in this report, as well as modifying the geometry model of the Terry turbine to make it more amenable for FLUENT analysis with rotating reference frames.

  10. Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

    SciTech Connect (OSTI)

    Scheele, Randall D.; Casella, Andrew M.

    2010-09-28

    This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

  11. Fuel handling system for a nuclear reactor

    DOE Patents [OSTI]

    Saiveau, James G. (Hickory Hills, IL); Kann, William J. (Park Ridge, IL); Burelbach, James P. (Glen Ellyn, IL)

    1986-01-01

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  12. Systems analysis of the CANDU 3 Reactor

    SciTech Connect (OSTI)

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H.

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  13. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    SciTech Connect (OSTI)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-30

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  14. Development of a system model for advanced small modular reactors.

    SciTech Connect (OSTI)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandia's concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  15. Weld monitor and failure detector for nuclear reactor system

    DOE Patents [OSTI]

    Sutton, Jr., Harry G. (Mt. Lebanon, PA)

    1987-01-01

    Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

  16. Automated Test Coverage Measurement for Reactor Protection System Software

    E-Print Network [OSTI]

    in implementing safety critical systems such as nuclear reactor protection systems. We have defined new structural) are widely used to implement safety- critical systems such as nuclear reactor protection systems, testing implementation language. The Korea Nuclear Instrumentation and Control System R&D Center (KNICS) project, whose

  17. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    SciTech Connect (OSTI)

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R.

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  18. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    SciTech Connect (OSTI)

    Not Available

    1982-06-01

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments.

  19. A Peltier cooling system for SiPM temperature stabilization

    E-Print Network [OSTI]

    Hebbeker, Thomas

    A Peltier cooling system for SiPM temperature stabilization von Simon Nieswand Bachelorarbeit außen thermisch isolierten Kupferblockes einzulassen, an welchen ein Peltier-Element angebracht wird. Um das System zu automatisieren, werden der Temperatursensor und die Stromquelle des Peltier- Elements

  20. Nuclear reactor with makeup water assist from residual heat removal system

    DOE Patents [OSTI]

    Corletti, Michael M. (New Kensington, PA); Schulz, Terry L. (Murrysville, PA)

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

  1. Nuclear reactor with makeup water assist from residual heat removal system

    DOE Patents [OSTI]

    Corletti, M.M.; Schulz, T.L.

    1993-12-07

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

  2. Integrated systems analysis of the PIUS reactor

    SciTech Connect (OSTI)

    Fullwood, F.; Kroeger, P.; Higgins, J.

    1993-11-01

    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  3. Thermal Storage with Conventional Cooling Systems 

    E-Print Network [OSTI]

    McGee, E. E.

    1990-01-01

    "Thermal Storage" is a term that describes a mechanical systems ability to sustain normal HVAC operations through a thermal retention source. This system allows for the curtailment of operating major refrigeration equipment during periods of high kw...

  4. Continuous production of tritium in an isotope-production reactor with a separate circulation system

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

  5. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    SciTech Connect (OSTI)

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  6. Radiation detector system having heat pipe based cooling

    DOE Patents [OSTI]

    Iwanczyk, Jan S.; Saveliev, Valeri D.; Barkan, Shaul

    2006-10-31

    A radiation detector system having a heat pipe based cooling. The radiation detector system includes a radiation detector thermally coupled to a thermo electric cooler (TEC). The TEC cools down the radiation detector, whereby heat is generated by the TEC. A heat removal device dissipates the heat generated by the TEC to surrounding environment. A heat pipe has a first end thermally coupled to the TEC to receive the heat generated by the TEC, and a second end thermally coupled to the heat removal device. The heat pipe transfers the heat generated by the TEC from the first end to the second end to be removed by the heat removal device.

  7. Solar Cooling Using Variable Geometry Ejectors Centre for Sustainable Energy Systems

    E-Print Network [OSTI]

    Solar Cooling Using Variable Geometry Ejectors M. Dennis Centre for Sustainable Energy Systems-mail:Mike.Dennis@anu.edu.au ABSTRACT Solar heat driven cooling systems are an attractive concept. The need for cooling is associated with high ambient temperatures. Ejector based cooling systems have been in existence for some time but have

  8. Performance enhancement of a compressive thermoelastic cooling system using multi-objective

    E-Print Network [OSTI]

    Rubloff, Gary W.

    Performance enhancement of a compressive thermoelastic cooling system using multi152 W cooling ca- pacity enhancement evaluated under 10 K waterewater system temperature lift. In addi cooling in this study refers to a solid-state cooling system using shape memory alloys (SMAs), due

  9. Cooling Flows of Self-Gravitating, Rotating, Viscous Systems

    E-Print Network [OSTI]

    Mohsen Shadmehri; Jamshid Ghanbari

    2002-04-06

    We obtain self-similar solutions that describe the dynamics of a self-gravitating, rotating, viscous system. We use simplifying assumptions; but explicitly include viscosity and the cooling due to the dissipation of energy. By assuming that the turbulent dissipation of energy is as power law of the density and the speed v_{rms} and for a power-law dependence of viscosity on the density, pressure, and rotational velocity, we investigate turbulent cooling flows. It has been shown that for the cylindrically and the spherically cooling flows the similarity indices are the same, and they depend only on the exponents of the dissipation rate and the viscosity model. Depending on the values of the exponents, which the mechanisms of the dissipation and viscosity determine them, we may have solutions with different general physical properties. The conservation of the total mass and the angular momentum of the system strongly depends on the mechanisms of energy dissipation and the viscosity model.

  10. Systems Evaluation at the Cool Energy House

    SciTech Connect (OSTI)

    J. Williamson and S. Puttagunta

    2013-09-01

    Steven Winter Associates, Inc. (SWA) monitored several advanced mechanical systems within a 2012 deep energy retrofitted home in the small Orlando suburb of Windermere, FL. This report provides performance results of one of the home's heat pump water heaters (HPWH) and the whole-house dehumidifier (WHD) over a six month period. In addition to assessing the energy performance of these systems, this study sought to quantify potential comfort improvements over traditional systems. This information is applicable to researchers, designers, plumbers, and HVAC contractors. Though builders and homeowners can find useful information within this report, the corresponding case studies are a likely better reference for this audience.

  11. Summary of evaporative cooling system for the SSC silicon tracker

    SciTech Connect (OSTI)

    Woloshun, K.; Barber, R.L.; Christensen, W.; Hanlon, J.A.; Keddy, M.D.; Miller, W.O.; Reid, R.S.; Ziock, H.J.

    1994-10-01

    An evaporative cooling system has been developed for the Superconducting Supercollider (SSC) Solenoidal Detector Collaboration (SDC) and the Gamma, Electron and Muon Detector (GEM) silicon tracker electronics. The system operated on the principles of the heat pipe; specifically, evaporation at near vapor-liquid equilibrium without the presence of noncondensible gases, and with a capillary media used to distribute the working fluid. The system used butane as a working fluid for operation at O{degrees}C and 1 atm. pressure. This paper summarizes the evolution of the system design, emphasizing key developments that may be useful for further work. Results of the system performance as of the close-of-effort are presented. A brief summary of results of experiments using a pumped single-phase cooling system are also presented.

  12. Wind turbine generators having wind assisted cooling systems and cooling methods

    DOE Patents [OSTI]

    Bagepalli, Bharat (Niskayuna, NY); Barnes, Gary R. (Delanson, NY); Gadre, Aniruddha D. (Rexford, NY); Jansen, Patrick L. (Scotia, NY); Bouchard, Jr., Charles G. (Schenectady, NY); Jarczynski, Emil D. (Scotia, NY); Garg, Jivtesh (Cambridge, MA)

    2008-09-23

    A wind generator includes: a nacelle; a hub carried by the nacelle and including at least a pair of wind turbine blades; and an electricity producing generator including a stator and a rotor carried by the nacelle. The rotor is connected to the hub and rotatable in response to wind acting on the blades to rotate the rotor relative to the stator to generate electricity. A cooling system is carried by the nacelle and includes at least one ambient air inlet port opening through a surface of the nacelle downstream of the hub and blades, and a duct for flowing air from the inlet port in a generally upstream direction toward the hub and in cooling relation to the stator.

  13. INCREMENTAL COOLING LOAD DETERMINATION FOR PASSIVE DIRECT GAIN HEATING SYSTEMS

    E-Print Network [OSTI]

    Sullivan, Paul W.

    2013-01-01

    May 27-30, 1981 INCREMENTAL COOLING LOAD DETERMINATION FOR12048 May 1981 INCREMENTAL COOLING LOAD DETERMINATION FORfor increases in the building cooling load resulting from

  14. BETTER DUCT SYSTEMS FOR HOME HEATING AND COOLING.

    SciTech Connect (OSTI)

    ANDREWS,J.

    2001-01-01

    This is a series of six guides intended to provide a working knowledge of residential heating and cooling duct systems, an understanding of the major issues concerning efficiency, comfort, health, and safety, and practical tips on installation and repair of duct systems. These guides are intended for use by contractors, system designers, advanced technicians, and other HVAC professionals. The first two guides are also intended to be accessible to the general reader.

  15. Plasma generators, reactor systems and related methods - Energy...

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

    and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably...

  16. Modular hybrid plasma reactor and related systems and methods...

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

    Patent Search Success Stories News Events Find More Like This Return to Search Modular hybrid plasma reactor and related systems and methods United States Patent Patent Number:...

  17. Modular hybrid plasma reactor and related systems and methods...

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

    (27) Visual Patent Search Success Stories News Events Return to Search Modular hybrid plasma reactor and related systems and methods United States Patent Application ***...

  18. Plasma generators, reactor systems and related methods - Energy...

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

    Visit the Technology Transfer and Commercialization Office Website Abstract: A plasma generator, reactor and associated systems and methods are provided in accordance with the...

  19. Auxiliary reactor for a hydrocarbon reforming system

    DOE Patents [OSTI]

    Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

    2006-01-17

    An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

  20. Turbine airfoil with an internal cooling system having vortex forming turbulators

    DOE Patents [OSTI]

    Lee, Ching-Pang

    2014-12-30

    A turbine airfoil usable in a turbine engine and having at least one cooling system is disclosed. At least a portion of the cooling system may include one or more cooling channels having a plurality of turbulators protruding from an inner surface and positioned generally nonorthogonal and nonparallel to a longitudinal axis of the airfoil cooling channel. The configuration of turbulators may create a higher internal convective cooling potential for the blade cooling passage, thereby generating a high rate of internal convective heat transfer and attendant improvement in overall cooling performance. This translates into a reduction in cooling fluid demand and better turbine performance.

  1. Air conditioning system with supplemental ice storing and cooling capacity

    DOE Patents [OSTI]

    Weng, Kuo-Lianq (Taichung, TW); Weng, Kuo-Liang (Taichung, TW)

    1998-01-01

    The present air conditioning system with ice storing and cooling capacity can generate and store ice in its pipe assembly or in an ice storage tank particularly equipped for the system, depending on the type of the air conditioning system. The system is characterized in particular in that ice can be produced and stored in the air conditioning system whereby the time of supplying cooled air can be effectively extended with the merit that the operation cycle of the on and off of the compressor can be prolonged, extending the operation lifespan of the compressor in one aspect. In another aspect, ice production and storage in great amount can be performed in an off-peak period of the electrical power consumption and the stored ice can be utilized in the peak period of the power consumption so as to provide supplemental cooling capacity for the compressor of the air conditioning system whereby the shift of peak and off-peak power consumption can be effected with ease. The present air conditioning system can lower the installation expense for an ice-storing air conditioning system and can also be applied to an old conventional air conditioning system.

  2. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    structure. The reactor uses carbide fuel, a com- posite ofvariables: - Fuel type (Oxide, Carbide, Nitride, Metallic) -99% N-15) Carbide Absorption in non-actinide fuel components

  3. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    waste generation, uranium mining needs and proliferationis little need for uranium mining, since these reactors willis no further need for uranium mining. B&B technology thus

  4. Rapid heating and cooling in two-dimensional Yukawa systems

    E-Print Network [OSTI]

    Yan Feng; Bin Liu; J. Goree

    2011-04-19

    Simulations are reported to investigate solid superheating and liquid supercooling of two-dimensional (2D) systems with a Yukawa interparticle potential. Motivated by experiments where a dusty plasma is heated and then cooled suddenly, we track particle motion using a simulation with Langevin dynamics. Hysteresis is observed when the temperature is varied rapidly in a heating and cooling cycle. As in the experiment, transient solid superheating, but not liquid supercooling, is observed. Solid superheating, which is characterized by solid structure above the melting point, is found to be promoted by a higher rate of temperature increase.

  5. Design and Control of Hydronic Radiant Cooling Systems

    E-Print Network [OSTI]

    Feng, Jingjuan Dove

    2014-01-01

    installed to block solar gain, ISO 11855 cooling capacityinstalled to block solar gain, ISO 11855 cooling capacity

  6. Method and system for powering and cooling semiconductor lasers

    DOE Patents [OSTI]

    Telford, Steven J; Ladran, Anthony S

    2014-02-25

    A semiconductor laser system includes a diode laser tile. The diode laser tile includes a mounting fixture having a first side and a second side opposing the first side and an array of semiconductor laser pumps coupled to the first side of the mounting fixture. The semiconductor laser system also includes an electrical pulse generator thermally coupled to the diode bar and a cooling member thermally coupled to the diode bar and the electrical pulse generator.

  7. Simulation and analysis of district-heating and -cooling systems

    SciTech Connect (OSTI)

    Bloomster, C.H.; Fassbender, L.L.

    1983-03-01

    A computer simulation model, GEOCITY, was developed to study the design and economics of district heating and cooling systems. GEOCITY calculates the cost of district heating based on climate, population, energy source, and financing conditions. The principal input variables are minimum temperature, heating degree-days, population size and density, energy supply temperature and distance from load center, and the interest rate. For district cooling, maximum temperature and cooling degree-hours are required. From this input data the model designs the fluid transport and district heating systems. From this design, GEOCITY calculates the capital and operating costs for the entire system. GEOCITY was originally developed to simulate geothermal district heating systems and thus, in addition to the fluid transport and distribution models, it includes a reservoir model to simulate the production of geothermal energy from geothermal reservoirs. The reservoir model can be adapted to simulate the supply of hot water from any other energy source. GEOCITY has been used extensively and has been validated against other design and cost studies. GEOCITY designs the fluid transport and distribution facilities and then calculates the capital and operating costs for the entire system. GEOCITY can simulate nearly any financial and tax structure through varying the rates of return on equity and debt, the debt-equity ratios, and tax rates. Both private and municipal utility systems can be simulated.

  8. Fluid sampling system for a nuclear reactor

    DOE Patents [OSTI]

    Lau, L.K.; Alper, N.I.

    1994-11-22

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump. 1 fig.

  9. Fluid sampling system for a nuclear reactor

    DOE Patents [OSTI]

    Lau, Louis K. (Monroeville, PA); Alper, Naum I. (Monroeville, PA)

    1994-01-01

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump.

  10. Exhaust system with emissions storage device and plasma reactor

    DOE Patents [OSTI]

    Hoard, John W. (Livonia, MI)

    1998-01-01

    An exhaust system for a combustion system, comprising a storage device for collecting NO.sub.x, hydrocarbon, or particulate emissions, or mixture of these emissions, and a plasma reactor for destroying the collected emissions is described. After the emission is collected in by the storage device for a period of time, the emission is then destroyed in a non-thermal plasma generated by the plasma reactor. With respect to the direction of flow of the exhaust stream, the storage device must be located before the terminus of the plasma reactor, and it may be located wholly before, overlap with, or be contained within the plasma reactor.

  11. Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

    SciTech Connect (OSTI)

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This report presents FY13 activities for the analysis of D2O cooled tight-pitch High-Conversion PWRs (HCPWRs) with U-Pu and Th-U fueled cores aiming at break-even or near breeder conditions while retaining the negative void reactivity. The analyses are carried out from several aspects which could not be covered in FY12 activities. SCALE 6.1 code system is utilized, and a series of simple 3D fuel pin-cell models are developed in order to perform Monte Carlo based criticality and burnup calculations. The performance of U-Pu fueled cores with axial and internal blankets is analyzed in terms of their impact on the relative fissile Pu mass balance, initial Pu enrichment, and void coefficient. In FY12, Pu conversion performances of D2O-cooled HCPWRs fueled with MOX were evaluated with small sized axial/internal DU blankets (approximately 4cm of axial length) in order to ensure the negative void reactivity, which evidently limits the conversion performance of HCPWRs. In this fiscal year report, the axial sizes of DU blankets are extended up to 30 cm in order to evaluate the amount of DU necessary to reach break-even and/or breeding conditions. Several attempts are made in order to attain the milestone of the HCPWR designs (i.e., break-even condition and negative void reactivity) by modeling of HCPWRs under different conditions such as boiling of D2O coolant, MOX with different 235U enrichment, and different target burnups. A similar set of analyses are performed for Th-U fueled cores. Several promising characteristics of 233U over other fissile like 239Pu and 235U, most notably its higher fission neutrons per absorption in thermal and epithermal ranges combined with lower ___ in the fast range than 239Pu allows Th-U cores to be taller than MOX ones. Such an advantage results in 4% higher relative fissile mass balance than that of U-Pu fueled cores while retaining the negative void reactivity until the target burnup of 51 GWd/t. Several other distinctions between U-Pu and Th-U fueled cores are identified by evaluating the sensitivity coefficients of keff, mass balance, and void coefficient. The effect of advanced iron alloy cladding (i.e., FeCrAl) on the performance of Pu conversion in MOX fueled cores is studied instead of using standard stainless-steel cladding. Variations in clad thickness and coolant-to-fuel volume ratio are also exercised. The use of FeCrAl instead of SS as a cladding alloy reduces the required Pu enrichment and improves the Pu conversion rate primarily due to the absence of nickel in the cladding alloy that results in the reduction of the neutron absorption. Also the difference in void coefficients between SS and FeCrAl alloys is nearly 500 pcm over the entire burnup range. The report also shows sensitivity and uncertainty analyses in order to characterize D2O cooled HCPWRs from different aspects. The uncertainties of integral parameters (keff and void coefficient) for selected reactor cores are evaluated at different burnup points in order to find similarities and trends respect to D2O-HCPWR.

  12. Floating Loop System For Cooling Integrated Motors And Inverters Using Hot Liquid Refrigerant

    DOE Patents [OSTI]

    Hsu, John S [Oak Ridge, TN; Ayers, Curtis W [Kingston, TN; Coomer, Chester [Knoxville, TN; Marlino, Laura D [Oak Ridge, TN

    2006-02-07

    A floating loop vehicle component cooling and air-conditioning system having at least one compressor for compressing cool vapor refrigerant into hot vapor refrigerant; at least one condenser for condensing the hot vapor refrigerant into hot liquid refrigerant by exchanging heat with outdoor air; at least one floating loop component cooling device for evaporating the hot liquid refrigerant into hot vapor refrigerant; at least one expansion device for expanding the hot liquid refrigerant into cool liquid refrigerant; at least one air conditioning evaporator for evaporating the cool liquid refrigerant into cool vapor refrigerant by exchanging heat with indoor air; and piping for interconnecting components of the cooling and air conditioning system.

  13. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    SciTech Connect (OSTI)

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  14. Cooling atom-cavity systems into entangled states

    E-Print Network [OSTI]

    J. Busch; S. De; S. S. Ivanov; B. T. Torosov; T. P. Spiller; A. Beige

    2011-05-25

    Generating entanglement by simply cooling a system into a stationary state which is highly entangled has many advantages. Schemes based on this idea are robust against parameter fluctuations, tolerate relatively large spontaneous decay rates, and achieve high fidelities independent of their initial state. A possible implementation of this idea in atom-cavity systems has recently been proposed by Kastoryano et al. [Phys. Rev. Lett. 106, 090502 (2011)]. Here we propose an improved entanglement cooling scheme for two atoms inside an optical cavity which achieves higher fidelities for comparable single-atom cooperativity parameters C. For example, we predict fidelities above 90% even for C as low as 20 without requiring individual laser addressing and without having to detect photons.

  15. Quantum Friction: Cooling Quantum Systems with Unitary Time Evolution

    E-Print Network [OSTI]

    Aurel Bulgac; Michael McNeil Forbes; Kenneth J. Roche; Gabriel Wlaz?owski

    2013-05-29

    We introduce a type of quantum dissipation -- local quantum friction -- by adding to the Hamiltonian a local potential that breaks time-reversal invariance so as to cool the system. Unlike the Kossakowski-Lindblad master equation, local quantum friction directly effects unitary evolution of the wavefunctions rather than the density matrix: it may thus be used to cool fermionic many-body systems with thousands of wavefunctions that must remain orthogonal. In addition to providing an efficient way to simulate quantum dissipation and non-equilibrium dynamics, local quantum friction coupled with adiabatic state preparation significantly speeds up many-body simulations, making the solution of the time-dependent Schr\\"odinger equation significantly simpler than the solution of its stationary counterpart.

  16. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel

    SciTech Connect (OSTI)

    Forsberg, C. W.; Terrani, Kurt A; Snead, Lance Lewis; Katoh, Yutai

    2012-01-01

    The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of {sup 7}LiF and BeF{sub 2} (FLiBe) possessing a boiling point above 1300 C and the figure of merit {rho}C{sub p} (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

  17. Steam Generator Component Model in a Combined Cycle of Power Conversion Unit for Very High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Oh, Chang H; Han, James; Barner, Robert; Sherman, Steven R

    2007-06-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP), Very High Temperature Gas-Cooled Reactor (VHTR) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. A combined cycle is considered as one of the power conversion units to be coupled to the very high-temperature gas-cooled reactor (VHTR). The combined cycle configuration consists of a Brayton top cycle coupled to a Rankine bottoming cycle by means of a steam generator. A detailed sizing and pressure drop model of a steam generator is not available in the HYSYS processes code. Therefore a four region model was developed for implementation into HYSYS. The focus of this study was the validation of a HYSYS steam generator model of two phase flow correlations. The correlations calculated the size and heat exchange of the steam generator. To assess the model, those calculations were input into a RELAP5 model and its results were compared with HYSYS results. The comparison showed many differences in parameters such as the heat transfer coefficients and revealed the different methods used by the codes. Despite differences in approach, the overall results of heat transfer were in good agreement.

  18. Beyond-Design-Basis-Accidents Passive Containment-Cooling Spray System

    SciTech Connect (OSTI)

    Karameldin, Aly; Temraz, Hassan M. Elsawy; Ibrahim, Nady Attia [Atomic Energy Authority (Egypt)

    2001-10-15

    The proposed safety feature considered in this study aims to increase the safety margins of nuclear power plants by proposed water tanks located inside or outside the upper zone of the containment to be utilized for (a) residual heat removal of the reactor in case of station blackout or in case of normal reactor shutdown and (b) beyond-design-basis accidents, in which core melt and debris-concrete interaction take place, associated with accumulative containment pressure increase and partial loss of the active systems. The proposed passive containment system can be implemented by a special mechanism, which can allow the pressurization of the water in the tanks and therefore can enable an additional spray system to start in case of increasing the containment pressure over a certain value just below the design pressure. A conservative case study is that of a Westinghouse 3411-MW(thermal) power station, where the proposed passive containment cooling spray system (PCCSS) will start at a pressure of 6 bars and terminate at a pressure of 3 bars. A one-dimensional lumped model is postulated to describe the thermal and hydraulic process behavior inside the containment after a beyond-design-basis accident. The considered parameters are the spray mass flow rate, the initial droplet diameters, fuel-cooling time, and the ultimate containment pressure. The overall heat and mass balance inside the containment are carried out, during both the containment depressurization (by the spraying system) and pressurization (by the residual energies). The results show that the design of the PCCSS is viable and has a capability to maintain the containment below the design pressure passively for the required grace period of 72 h. Design curves of the proposed PCCSS indicate the effect of the spray flow rate and cooling time on the total sprayed volume during the grace period of 72 h. From these curves it can be concluded that for the grace period of 72 h, the required tank volumes are 3800 and 4700 m{sup 3}, corresponding to fuel-cooling times (time after shutdown) of two weeks and one week, respectively. This large quantity of water serves as an ultimate heat sink available for the residual heat removal in the case of station blackout. The optimal spraying droplet diameter, travel, and mass flow rate are 3 mm, 30 m, and 100 to 125 kg/s, respectively.

  19. NASA's Marshall Space Flight Center Improves Cooling System Performance

    SciTech Connect (OSTI)

    2011-02-22

    National Aeronautics and Space Administration’s (NASA) Marshall Space Flight Center (MSFC) has a longstanding sustainability program that revolves around energy and water efficiency as well as environmental protection. MSFC identified a problematic cooling loop with six separate compressor heat exchangers and a history of poor efficiency. The facility engineering team at MSFC partnered with Flozone Services, Incorporated to implement a comprehensive water treatment platform to improve the overall efficiency of the system.

  20. Heating and Cooling System Support Equipment Basics | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative Fuelsof Energy ServicesContracting OversightEMS Policyand Cooling System Support

  1. 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.

  2. Natural convection heat transport in a small, HLMC reactor system

    SciTech Connect (OSTI)

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

    1999-09-01

    Concepts are being developed and evaluated at Argonne National Laboratory for a small nuclear steam supply system (NSSS) with proliferation-resistant features targeted for export to developing countries. Here the authors are specifically investigating how simple and compact such a system can be. A heavy-liquid-metal coolant (HLMC) is being considered owing to its excellent heat transport characteristics and its relative inertness with the reference thermodynamic working fluid (water/steam). The purpose of the present work is to explore the possibility to take advantage of these HLMC characteristics by eliminating the intermediate loop needed in sodium-cooled systems and additionally eliminating the primary system coolant pumps. The criteria imposed on the system include the following: (1) low power, i.e., 300 MW(thermal); (2) small size for factory fabrication and overland transportation; (3) elimination of fuel access at the site (no refueling, fuel shuffling, nor storage at site); integral fueled module replacement at 15-yr goal interval; and (4) completion of all research and development needed for detailed prototype design within 5 yr. To accomplish the latter requirement, the authors are addressing whether existing coolant and materials technology is capable of supporting the sought-after simplifications. In this regard, they are at present considering technology developed in Russia for Pb-Bi eutectic as a reactor coolant and ferritic-martensitic stainless steel with oxide-layer corrosion protection as cladding. The figure of merit in the investigation is the peak cladding temperature insofar as the cladding technology is considered proven to {approximately}600 C.

  3. Idaho National Laboratory Experimental Program to Measure the Flow Phenomena in a Scaled Model of a Prismatic Gas-Cooled Reactor Lower Plenum for Validation of CFD Codes

    SciTech Connect (OSTI)

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink

    2008-09-01

    The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a prismatic gas-cooled reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A description of the scaling analysis, experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that will be presented include mean-velocity-field and turbulence data in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The flow in the lower plenum consists of multiple jets injected into a confined cross flow - with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the mineral oil working fluid. The benefit of the MIR technique is that it permits high-quality measurements to be obtained without locating intrusive transducers that disturb the flow field and without distortion of the optical paths. An advantage of the INL MIR system is its large size which allows improved spatial and temporal resolution compared to similar facilities at smaller scales. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements reveal developing, non-uniform flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and charts that describe the component flows at specific regions in the model. Information on inlet velocity profiles is also presented.

  4. Developing Fully Coupled Dynamical Reactor Core Isolation System Models in RELAP-7 for Extended Station Black-Out Analysis

    SciTech Connect (OSTI)

    Haihua Zhao; Ling Zou; Hongbin Zhang; David Andrs; Richard Martineau

    2014-04-01

    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup water to the reactor vessel for core cooling when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. It was one of the very few safety systems still available during the Fukushima Daiichi accidents after the tsunamis hit the plants and the system successfully delayed the core meltdown for a few days for unit 2 & 3. Therefore, detailed models for RCIC system components are indispensable to understand extended station black-out accidents (SBO) for BWRs. As part of the effort to develop the new generation reactor system safety analysis code RELAP-7, major components to simulate the RCIC system have been developed. This paper describes the models for those components such as turbine, pump, and wet well. Selected individual component test simulations and a simplified SBO simulation up to but before core damage is presented. The successful implementation of the simplified RCIC and wet well models paves the way to further improve the models for safety analysis by including more detailed physical processes in the near future.

  5. Improving the Efficiency of Your Process Cooling System 

    E-Print Network [OSTI]

    Baker, R.

    2005-01-01

    Many industries require process cooling to achieve desired outcomes of specific processes. This cooling may come from cooling towers, once-through water, mechanical refrigeration, or cryogenic sources such as liquid nitrogen or dry ice. This paper...

  6. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  7. Optica Applicata, Vol. XXXVI, No. 4, 2006 A system for magnetooptical cooling

    E-Print Network [OSTI]

    Optica Applicata, Vol. XXXVI, No. 4, 2006 A system for magnetooptical cooling and trapping of Rb@ifpan.edu.pl A system for magnetooptical cooling and trapping of Rb atoms built in our laboratory (at Institute recent monographs [6, 7]). The backbone system, widely used for cooling and trapping neutral atoms

  8. User-friendly and intuitive graphical approach to the design of thermoelectric cooling systems

    E-Print Network [OSTI]

    User-friendly and intuitive graphical approach to the design of thermoelectric cooling systems)-based active cooling system, including the heatsink role. The method is simple and intuitive and provides com- prehensive information about the cooling system such as its feasibility, required heatsink, the TEC current

  9. OPTIMAL CONTROL OF THE COOLING SYSTEM IN HEAVY VEHICLES1 Niklas Petterssona,b

    E-Print Network [OSTI]

    Johansson, Karl Henrik

    OPTIMAL CONTROL OF THE COOLING SYSTEM IN HEAVY VEHICLES1 Niklas Petterssona,b , Karl Henrik is supported by Scania CV AB and PFF. This paper presents a study on cooling system control in heavy vehicles physical layouts. The outline of the paper is as follows: In section 2 the cooling system is described

  10. Adaptive Thermal Management for Portable System Batteries by Forced Convection Cooling

    E-Print Network [OSTI]

    Pedram, Massoud

    Adaptive Thermal Management for Portable System Batteries by Forced Convection Cooling Qing Xie a portable system. Since the cooling fan is also powered by the same battery, it is critical to develop thermal management problem for batteries (ATMB) in the portable systems with forced convection cooling

  11. Impact of material system thermomechanics and thermofluid performance on He-cooled ceramic

    E-Print Network [OSTI]

    Abdou, Mohamed

    Impact of material system thermomechanics and thermofluid performance on He-cooled ceramic breeder program for high temperature gas-cooled blanket systems using SiCf /SiC as a structural material. Current as with helium-cooled ceramic breeder blanket systems. Thus, both the design and issue relevant R&D emphasis

  12. Review and Projections of Integrated Cooling Systems for Three-Dimensional

    E-Print Network [OSTI]

    Kandlikar, Satish

    Review and Projections of Integrated Cooling Systems for Three-Dimensional Integrated Circuits and integrated cooling systems. For heat fluxes of 50­100 W/cm2 on each side of a chip in a 3D IC package outstanding issues in the cooling system design were outlined. Before reviewing available literature

  13. A Simple and Intuitive Graphical Approach to the Design of Thermoelectric Cooling Systems

    E-Print Network [OSTI]

    A Simple and Intuitive Graphical Approach to the Design of Thermoelectric Cooling Systems Simon, thermoelectric active cooling systems can help maintain electronic devices at a desired temperature condition for calculating the steady-state operational point of a TEC based active cooling system, including the heatsink

  14. A Conceptual Multi-Megawatt System Based on a Tungsten CERMET Reactor

    SciTech Connect (OSTI)

    Jonathan A. Webb; Brian Gross

    2011-02-01

    Abstract. A conceptual reactor system to support Multi-Megawatt Nuclear Electric Propulsion is investigated within this paper. The reactor system consists of a helium cooled Tungsten-UN fission core, surrounded by a beryllium neutron reflector and 13 B4C control drums coupled to a high temperature Brayton power conversion system. Excess heat is rejected via carbon reinforced heat pipe radiators and the gamma and neutron flux is attenuated via segmented shielding consisting of lithium hydride and tungsten layers. Turbine inlet temperatures ranging from 1300 K to 1500 K are investigated for their effects on specific powers and net electrical outputs ranging from 1 MW to 100 MW. The reactor system is estimated to have a mass, which ranges from 15 Mt at 1 MWe and a turbine inlet temperature of 1500 K to 1200 Mt at 100 MWe and a turbine temperature of 1300 K. The reactor systems specific mass ranges from 32 kg/kWe at a turbine inlet temperature of 1300 K and a power of 1 MWe to 9.5 kg/kW at a turbine temperature of 1500 K and a power of 100 MWe.

  15. POOL WATER TREATMENT AND COOLING SYSTEM DESCRIPTION DOCUMENT

    SciTech Connect (OSTI)

    V. King

    2000-06-19

    The Pool Water Treatment and Cooling System is located in the Waste Handling Building (WHB), and is comprised of various process subsystems designed to support waste handling operations. This system maintains the pool water temperature within an acceptable range, maintains water quality standards that support remote underwater operations and prevent corrosion, detects leakage from the pool liner, provides the capability to remove debris from the pool, controls the pool water level, and helps limit radiological exposure to personnel. The pool structure and liner, pool lighting, and the fuel staging racks in the pool are not within the scope of the Pool Water Treatment and Cooling System. Pool water temperature control is accomplished by circulating the pool water through heat exchangers. Adequate circulation and mixing of the pool water is provided to prevent localized thermal hotspots in the pool. Treatment of the pool water is accomplished by a water treatment system that circulates the pool water through filters, and ion exchange units. These water treatment units remove radioactive and non-radioactive particulate and dissolved solids from the water, thereby providing the water clarity needed to conduct waste handling operations. The system also controls pool water chemistry to prevent advanced corrosion of the pool liner, pool components, and fuel assemblies. Removal of radioactivity from the pool water contributes to the project ALARA (as low as is reasonably achievable) goals. A leak detection system is provided to detect and alarm leaks through the pool liner. The pool level control system monitors the water level to ensure that the minimum water level required for adequate radiological shielding is maintained. Through interface with a demineralized water system, adequate makeup is provided to compensate for loss of water inventory through evaporation and waste handling operations. Interface with the Site Radiological Monitoring System provides continuous radiological monitoring of the pool water. The Pool Water Treatment and Cooling System interfaces with the Waste Handling Building System, Site-Generated Radiological Waste Handling System, Site Radiological Monitoring System, Waste Handling Building Electrical System, Site Water System, and the Monitored Geologic Repository Operations Monitoring and Control System.

  16. Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

    SciTech Connect (OSTI)

    Samuel Bays; Pavel Medvedev; Michael Pope; Rodolfo Ferrer; Benoit Forget; Mehdi Asgari

    2009-04-01

    This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

  17. Testing of Passive Safety System Performance for Higher Power Advanced Reactors

    SciTech Connect (OSTI)

    brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

    2004-12-31

    This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

  18. Active cooling for downhole instrumentation: Preliminary analysis and system selection

    SciTech Connect (OSTI)

    Bennett, G.A.

    1988-03-01

    A feasibility study and a series of preliminary designs and analyses were done to identify candidate processes or cycles for use in active cooling systems for downhole electronic instruments. A matrix of energy types and their possible combinations was developed and the energy conversion process for each pari was identified. The feasibility study revealed conventional as well as unconventional processes and possible refrigerants and identified parameters needing further clarifications. A conceptual design or series od oesigns for each system was formulated and a preliminary analysis of each design was completed. The resulting coefficient of performance for each system was compared with the Carnot COP and all systems were ranked by decreasing COP. The system showing the best combination of COP, exchangeability to other operating conditions, failure mode, and system serviceability is chosen for use as a downhole refrigerator. 85 refs., 48 figs., 33 tabs.

  19. Passive-solar directional-radiating cooling system

    DOE Patents [OSTI]

    Hull, John R. (Hinsdale, IL); Schertz, William W. (Batavia, IL)

    1986-01-01

    A radiative cooling system for use with an ice-making system having a radiating surface aimed at the sky for radiating energy at one or more wavelength bands for which the atmosphere is transparent and a cover thermally isolated from the radiating surface and transparent at least to the selected wavelength or wavelengths, the thermal isolation reducing the formation of condensation on the radiating surface and/or cover and permitting the radiation to continue when the radiating surface is below the dewpoint of the atmosphere, and a housing supporting the radiating surface, cover and heat transfer means to an ice storage reservoir.

  20. System and method for cooling a super-conducting device

    DOE Patents [OSTI]

    Bray, James William (Niskayuna, NY); Steinbach, Albert Eugene (Schenectady, NY); Dawson, Richard Nils (Voorheesville, NY); Laskaris, Evangelos Trifon (Schenectady, NY); Huang, Xianrul (Clifton Park, NY)

    2008-01-08

    A system and method for cooling a superconductive rotor coil. The system comprises a rotatable shaft coupled to the superconductive rotor coil. The rotatable shaft may comprise an axial passageway extending through the rotatable shaft and a first passageway extending through a wall of the rotatable shaft to the axial passageway. The axial passageway and the first passageway are operable to convey a cryogenic fluid to the superconductive rotor coil through the wall of the rotatable shaft. A cryogenic transfer coupling may be provided to supply cryogenic fluid to the first passageway.

  1. Passive-solar directional-radiating cooling system

    DOE Patents [OSTI]

    Hull, J.R.; Schertz, W.W.

    1985-06-27

    A radiative cooling system for use with an ice-making system having a radiating surface aimed at the sky for radiating energy at one or more wavelength bands for which the atmosphere is transparent and a cover thermally isolated from the radiating surface and transparent at least to the selected wavelength or wavelengths, the thermal isolation reducing the formation of condensation on the radiating surface and/or cover and permitting the radiation to continue when the radiating surface is below the dewpoint of the atmosphere, and a housing supporting the radiating surface, cover and heat transfer means to an ice storage reservoir.

  2. Cooling and control of a cavity opto-electromechanical system

    E-Print Network [OSTI]

    Kwan H. Lee; Terry G. McRae; Glen I. Harris; Joachim Knittel; Warwick P. Bowen

    2010-02-11

    We implement a cavity opto-electromechanical system integrating electrical actuation capabilities of nanoelectromechanical devices with ultrasensitive mechanical transduction achieved via intra-cavity optomechanical coupling. Electrical gradient forces as large as 0.40 microN are realized, with simultaneous mechanical transduction sensitivity of 1.5 X 10^-18 m/rtHz representing a three orders of magnitude improvement over any nanoelectromechanical system to date. Opto-electromechanical feedback cooling is demonstrated, exhibiting strong squashing of the in-loop transduction signal. Out-of-loop transduction provides accurate temperature calibration even in the critical paradigm where measurement backaction induces opto-mechanical correlations.

  3. SP-100 Program: space reactor system and subsystem investigations

    SciTech Connect (OSTI)

    Harty, R.B.

    1983-09-30

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

  4. Investigation of plant control strategies for the supercritical C0{sub 2}Brayton cycle for a sodium-cooled fast reactor using the plant dynamics code.

    SciTech Connect (OSTI)

    Moisseytsev, A.; Sienicki, J.

    2011-04-12

    The development of a control strategy for the supercritical CO{sub 2} (S-CO{sub 2}) Brayton cycle has been extended to the investigation of alternate control strategies for a Sodium-Cooled Fast Reactor (SFR) nuclear power plant incorporating a S-CO{sub 2} Brayton cycle power converter. The SFR assumed is the 400 MWe (1000 MWt) ABR-1000 preconceptual design incorporating metallic fuel. Three alternative idealized schemes for controlling the reactor side of the plant in combination with the existing automatic control strategy for the S-CO{sub 2} Brayton cycle are explored using the ANL Plant Dynamics Code together with the SAS4A/SASSYS-1 Liquid Metal Reactor (LMR) Analysis Code System coupled together using the iterative coupling formulation previously developed and implemented into the Plant Dynamics Code. The first option assumes that the reactor side can be ideally controlled through movement of control rods and changing the speeds of both the primary and intermediate coolant system sodium pumps such that the intermediate sodium flow rate and inlet temperature to the sodium-to-CO{sub 2} heat exchanger (RHX) remain unvarying while the intermediate sodium outlet temperature changes as the load demand from the electric grid changes and the S-CO{sub 2} cycle conditions adjust according to the S-CO{sub 2} cycle control strategy. For this option, the reactor plant follows an assumed change in load demand from 100 to 0 % nominal at 5 % reduction per minute in a suitable fashion. The second option allows the reactor core power and primary and intermediate coolant system sodium pump flow rates to change autonomously in response to the strong reactivity feedbacks of the metallic fueled core and assumed constant pump torques representing unchanging output from the pump electric motors. The plant behavior to the assumed load demand reduction is surprising close to that calculated for the first option. The only negative result observed is a slight increase in the intermediate inlet sodium temperatures by about 10 C. This temperature rise could presumably be precluded or significantly reduced through fine adjustment of the control rods and pump motors. The third option assumes that the reactor core power and primary and intermediate system flow rates are ideally reduced linearly in a programmed fashion that instantaneously matches the prescribed load demand. The calculated behavior of this idealized case reveals a number of difficulties because the control strategy for the S-CO{sub 2} cycle overcools the reactor potentially resulting in the calculation of sodium bulk freezing and the onset of sodium boiling. The results show that autonomous SFR operation may be viable for the particular assumed load change transient and deserves further investigation for other transients and postulated accidents.

  5. The Helium Cooling System and Cold Mass Support System for theMICE Coupling Solenoid

    SciTech Connect (OSTI)

    Wang, L.; Wu, H.; Li, L.K.; Green, M.A.; Liu, C.S.; Li, L.Y.; Jia, L.X.; Virostek, S.P.

    2007-08-27

    The MICE cooling channel consists of alternating threeabsorber focus coil module (AFC) and two RF coupling coil module (RFCC)where the process of muon cooling and reacceleration occurs. The RFCCmodule comprises a superconducting coupling solenoid mounted around fourconventional conducting 201.25 MHz closed RF cavities and producing up to2.2T magnetic field on the centerline. The coupling coil magnetic fieldis to produce a low muon beam beta function in order to keep the beamwithin the RF cavities. The magnet is to be built using commercialniobium titanium MRI conductors and cooled by pulse tube coolers thatproduce 1.5 W of cooling capacity at 4.2 K each. A self-centering supportsystem is applied for the coupling magnet cold mass support, which isdesigned to carry a longitudinal force up to 500 kN. This report willdescribe the updated design for the MICE coupling magnet. The cold masssupport system and helium cooling system are discussed indetail.

  6. Photothermal Self-Oscillation and Laser Cooling of Graphene Optomechanical Systems

    E-Print Network [OSTI]

    McEuen, Paul L.

    Photothermal Self-Oscillation and Laser Cooling of Graphene Optomechanical Systems Robert A. Barton systems (NEMS), photothermal force, laser cooling, self-oscillation Optomechanics,1,2 which uses optical that a continuous wave laser can be used to cool a graphene vibrational mode or to power a graphene-based tunable

  7. Optimal Scheduling for Biocide and Heat Exchangers Maintenance Towards Environmentally Friendly Seawater Cooling Systems 

    E-Print Network [OSTI]

    Binmahfouz, Abdullah

    2012-10-19

    FOR SEAWATER-COOLED POWER AND DESALINATION PLANTS....................................................... 127 5.1 Overview .............................................................................................. 127 5.2 Introduction... 5.2 Representation of a Once-Thorough Cooling System................................ 141 5.3 An Overall Representation of the Power/Desalination Plant ..................... 152 5.4 The Cooling System for the Case Study...

  8. Impact of Different Glazing Systems on Cooling Load of a Detached Residential Building at Bhubaneswar, India 

    E-Print Network [OSTI]

    Sahoo, P. K.; Sahoo, R.

    2010-01-01

    on solar heat gain and cooling load. The simulation results show reductions in solar heat gain, cooling load and better thermal comfort can be achieved using proper glazing systems for a specific orientation of the building. The significance...

  9. Impact of Solar Heat Gain on Radiant Floor Cooling System Design

    E-Print Network [OSTI]

    Feng, Jingjuan Dove; Schiavon, Stefano; Bauman, Fred

    2013-01-01

    7] B. Borresen, Floor heating and cooling of an atrium, in:thermal performance of floor heating systems, Solar Energy,discussed this issue for floor heating, but not cooling.

  10. Hydraulic Modeling of Large District Cooling Systems for Master Planning Purposes 

    E-Print Network [OSTI]

    Xu, C.; Chen, Q.; Claridge, D. E.; Turner, W. D.; Deng, S.

    2006-01-01

    District Cooling Systems for Master Planning Purposes Chen Xu Qiang Chen David E. Claridge Dan Turner Song Deng Energy Systems Laboratory Texas A&M University College Station, TX 77843-3581 KEYWORD Pipe Network, District Cooling System, Central... Chilled Water System, Master Planning, Hydraulic Simulation ABSTRACT District Cooling Systems (DCS) have been widely applied in large institutions such as universities, government facilities, commercial districts, airports etc. The hydraulic system of a...

  11. Annular core for Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    SciTech Connect (OSTI)

    Turner, R.F.; Baxter, A.M.; Stansfield, O.M.; Vollman, R.E.

    1987-08-01

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40% greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93-m tall. Fuel elements contain TRISO-coated microspheres of 19.8% enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above.

  12. Hybrid sulfur cycle operation for high-temperature gas-cooled reactors

    DOE Patents [OSTI]

    Gorensek, Maximilian B

    2015-02-17

    A hybrid sulfur (HyS) cycle process for the production of hydrogen is provided. The process uses a proton exchange membrane (PEM) SO.sub.2-depolarized electrolyzer (SDE) for the low-temperature, electrochemical reaction step and a bayonet reactor for the high-temperature decomposition step The process can be operated at lower temperature and pressure ranges while still providing an overall energy efficient cycle process.

  13. Numerical simulation of flow distribution for pebble bed high temperature gas cooled reactors 

    E-Print Network [OSTI]

    Yesilyurt, Gokhan

    2004-09-30

    to be investigated. No detailed complete calculations for this kind of reactor to address these local phenomena are available. This work is an attempt to bridge this gap by evaluating this effect. I.2 TURBULENCE MODEL SELECTION The simulation of these local... number of numerical studies on flows around spherical bodies, none of them use the necessary turbulence models that are required to simulate flow where strong separation exists. With the development of high performance computers built for applications...

  14. System and method for cooling a superconducting rotary machine

    DOE Patents [OSTI]

    Ackermann, Robert Adolf (Schenectady, NY); Laskaris, Evangelos Trifon (Schenectady, NY); Huang, Xianrui (Clifton Park, NY); Bray, James William (Niskayuna, NY)

    2011-08-09

    A system for cooling a superconducting rotary machine includes a plurality of sealed siphon tubes disposed in balanced locations around a rotor adjacent to a superconducting coil. Each of the sealed siphon tubes includes a tubular body and a heat transfer medium disposed in the tubular body that undergoes a phase change during operation of the machine to extract heat from the superconducting coil. A siphon heat exchanger is thermally coupled to the siphon tubes for extracting heat from the siphon tubes during operation of the machine.

  15. Systems Issues in Nuclear Reactor Safety

    E-Print Network [OSTI]

    de Weck, Olivier L.

    postulated Loss-of-Coolant Accident (LOCA): 9 (LOCA): a double-ended break of the largest reactor coolant line, the concurrent loss of offsite power, and a single failure of an active ECCS component Loss Of Offsite Power Initiating Event 51,940 Steam Generator Tube Rupture Initiating Event 41,200 12

  16. Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

    SciTech Connect (OSTI)

    Katoh, Yutai [ORNL; Wilson, Dane F [ORNL; Forsberg, Charles W [ORNL

    2007-09-01

    The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

  17. Passive containment cooling water distribution device

    DOE Patents [OSTI]

    Conway, Lawrence E. (Hookstown, PA); Fanto, Susan V. (Plum Borough, PA)

    1994-01-01

    A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using a series of radial guide elements and cascading weir boxes to collect and then distribute the cooling water into a series of distribution areas through a plurality of cascading weirs. The cooling water is then uniformly distributed over the curved surface by a plurality of weir notches in the face plate of the weir box.

  18. Control of reactor coolant flow path during reactor decay heat removal

    DOE Patents [OSTI]

    Hunsbedt, Anstein N. (Los Gatos, CA)

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  19. East Bank District Heating-to-Cooling Conversion Plan Check the date your building's cooling system is scheduled to be on.

    E-Print Network [OSTI]

    Webb, Peter

    East Bank District Heating-to-Cooling Conversion Plan Check the date your building's cooling system Coal Storage Building 39 NA Cooke Hall 56 Donhowe Building 044 East Gateway District Steam Distr. 199

  20. UNIVERSITY OF CALIFORNIA, SAN DIEGO Temperature and Cooling Management in Computing Systems

    E-Print Network [OSTI]

    Simunic, Tajana

    UNIVERSITY OF CALIFORNIA, SAN DIEGO Temperature and Cooling Management in Computing Systems . . . . . . . . . . . 3 1.2 Related work on cooling management . . . . . . . . . . . 5 1.3 Thesis contributions . . . . . . . . . . . . . . . . . . . . . . 39 Chapter 4 Thermal and Cooling Management in Multisocket CPU Servers 41 4.1 Multi-tier thermal

  1. Cooling a quantum circuit via coupling to a multiqubit system

    E-Print Network [OSTI]

    Mihai A. Macovei

    2010-04-19

    The cooling effects of a quantum LC circuit coupled inductively with an ensemble of artificial qubits are investigated. The particles may decay independently or collectively through their interaction with the environmental vacuum electromagnetic field reservoir. For appropriate bath temperatures and the resonator's quality factors, we demonstrate an effective cooling well below the thermal background. In particular, we found that for larger samples the cooling efficiency is better for independent qubits. However, the cooling process can be faster for collectively interacting particles.

  2. Dual annular rotating "windowed" nuclear reflector reactor control system

    DOE Patents [OSTI]

    Jacox, Michael G. (Idaho Falls, ID); Drexler, Robert L. (Idaho Falls, ID); Hunt, Robert N. M. (Idaho Falls, ID); Lake, James A. (Idaho Falls, ID)

    1994-01-01

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

  3. Analysis of the Reactor Cavity Cooling System for Very High Temperature Gas-cooled Reactors Using Computational Fluid Dynamics Tools 

    E-Print Network [OSTI]

    Frisani, Angelo

    2011-08-08

    of the RCCS. Two models were developed to analyze heat exchange in the RCCS. Both models incorporate a 180 degree section resembling the VHTR RCCS bench table test facility performed at Texas A&M University. All the key features of the experimental facility...

  4. Thermoelectric generator cooling system and method of control

    DOE Patents [OSTI]

    Prior, Gregory P; Meisner, Gregory P; Glassford, Daniel B

    2012-10-16

    An apparatus is provided that includes a thermoelectric generator and an exhaust gas system operatively connected to the thermoelectric generator to heat a portion of the thermoelectric generator with exhaust gas flow through the thermoelectric generator. A coolant system is operatively connected to the thermoelectric generator to cool another portion of the thermoelectric generator with coolant flow through the thermoelectric generator. At least one valve is controllable to cause the coolant flow through the thermoelectric generator in a direction that opposes a direction of the exhaust gas flow under a first set of operating conditions and to cause the coolant flow through the thermoelectric generator in the direction of exhaust gas flow under a second set of operating conditions.

  5. Decommissioning the Romanian Water-Cooled Water-Moderated Research Reactor: New Environmental Perspective on the Management of Radioactive Waste

    SciTech Connect (OSTI)

    Barariu, G.; Giumanca, R. [Romanian Authority for Nuclear Activity (RAAN), Subsidiary of Technology and Engineering for Nuclear Objectives (SITON), 111 Atomistilor St., Bucuresti-Magurele, Ilfov (Romania)

    2006-07-01

    Pre-feasibility and feasibility studies were performed for decommissioning of the water-cooled water-moderated research reactor (WWER) located in Bucharest - Magurele, Romania. Using these studies as a starting point, the preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as for the rehabilitation of the existing Radioactive Waste Treatment Plant and for the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and ecological reconstruction of the grounds need to be provided for, in accordance with national and international regulations. In accordance with IAEA recommendations at the time, the pre-feasibility study proposed three stages of decommissioning. However, since then new ideas have surfaced with regard to decommissioning. Thus, taking into account the current IAEA ideology, the feasibility study proposes that decommissioning of the WWER be done in one stage to an unrestricted clearance level of the reactor building in an Immediate Dismantling option. Different options and the corresponding derived preferred option for waste management are discussed taking into account safety measures, but also considering technical, logistical and economic factors. For this purpose, possible types of waste created during each decommissioning stage are reviewed. An approximate inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the recommended international basic safety standards identified in the previous phase of the project. The existing Radioactive Waste Treatment Plant (RWTP) from the Horia Hulubei Institute for Nuclear Physics and Engineering (IFIN-HH), which has been in service with no significant upgrade since 1974, will need refurbishing due to deterioration, as well as upgrading in order to ensure the plant complies with current safety standards. This plant will also need to be adapted to treat wastes generated by WWER dismantling. The Baita-Bihor National Radioactive Waste Disposal Facility consists of two galleries in an abandoned uranium mine located in the central-western part of the Bihor Mountains in Transylvania. The galleries lie at a depth of 840 m. The facility requires a considerable overhaul. Several steps recommended for the upgrade of the facility are explored. Environmental concerns have lately become a crucial part of the radioactive waste management strategy. As such, all decisions must be made with great regard for land utilization around nuclear objectives. (authors)

  6. Reference reactor module for NASA's lunar surface fission power system

    SciTech Connect (OSTI)

    Poston, David I; Kapernick, Richard J; Dixon, David D; Werner, James; Qualls, Louis; Radel, Ross

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  7. Integrated intelligent systems in advanced reactor control rooms

    SciTech Connect (OSTI)

    Beckmeyer, R.R.

    1989-01-01

    An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs., 5 figs.

  8. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect (OSTI)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  9. System aspects of a Space Nuclear Reactor Power System

    SciTech Connect (OSTI)

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Kia, T.; Nesmith, B.

    1988-01-01

    Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, altitude control, reliability, and survivability. Among the findings are that the stowed length of the power system is important to mission design and that orbital storage for months to years may be needed for missions involving orbital assembly. The power system design evolved during the study and has continued to evolve; the current design differs somewhat from that examined in this paper.

  10. Pressurized reactor system and a method of operating the same

    DOE Patents [OSTI]

    Isaksson, J.M.

    1996-06-18

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Super-atmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gasification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor. 2 figs.

  11. Pressurized reactor system and a method of operating the same

    DOE Patents [OSTI]

    Isaksson, Juhani M. (Karhula, FI)

    1996-01-01

    A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

  12. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    criteria for FBR/B&B safety systems/designs . . . . . . . . .Safety systems/designs violations of evaluation critera ARC-LL expansion liquid criteria . . . . . . . . . . . .criteria for systems and design-approaches to improve the inherent safety

  13. Interfacing systems loss-of-coolant accident in Oconee-1 pressurized water reactor

    SciTech Connect (OSTI)

    Nassersharif, B.

    1984-01-01

    The primary system of a pressurized water reactor (PWR) operates at a relatively high pressure (15.5 MPa, 2250 psia) and consists of piping and components designed to withstand these pressures. The low-pressure-injection system (LPIS) connects to the primary system but possesses low-pressure piping passing outside the containment. Therefore, a potential exists for a loss-of-coolant accident (LOCA) outside the containment and concurrent damage to systems needed to cope with this problem. The emergency core-cooling system (ECCS) is assumed to be available for this event. A set of calculations were performed using the TRAC-PF1 code and a model of the Oconee-1 PWR to investigate the consequences of, and possible operator actions for, such an accident scenario.

  14. Regulatory Concerns on the In-Containment Water Storage System of the Korean Next Generation Reactor

    SciTech Connect (OSTI)

    Ahn, Hyung-Joon; Lee, Jae-Hun; Bang, Young-Seok; Kim, Hho-Jung

    2002-07-15

    The in-containment water storage system (IWSS) is a newly adopted system in the design of the Korean Next Generation Reactor (KNGR). It consists of the in-containment refueling water storage tank, holdup volume tank, and cavity flooding system (CFS). The IWSS has the function of steam condensation and heat sink for the steam release from the pressurizer and provides cooling water to the safety injection system and containment spray system in an accident condition and to the CFS in a severe accident condition. With the progress of the KNGR design, the Korea Institute of Nuclear Safety has been developing Safety and Regulatory Requirements and Guidances for safety review of the KNGR. In this paper, regarding the IWSS of the KNGR, the major contents of the General Safety Criteria, Specific Safety Requirements, Safety Regulatory Guides, and Safety Review Procedures were introduced, and the safety review items that have to be reviewed in-depth from the regulatory viewpoint were also identified.

  15. Evaluation of cooling performance of thermally activated building system with evaporative cooling source for typical United States climates

    E-Print Network [OSTI]

    Feng, Jingjuan; Bauman, Fred

    2013-01-01

    and high temperature cooling_REHVA Guidebook, Federation ofEvaluation of cooling performance of thermally activatedsystem with evaporative cooling source for typical United

  16. Nuclear reactor heat transport system component low friction support system

    DOE Patents [OSTI]

    Wade, Elman E. (Ruffs Dale, PA)

    1980-01-01

    A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

  17. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    SciTech Connect (OSTI)

    Mit Basol; John F. Kielb; John F. MuHooly; Kobus Smit

    2007-05-02

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate materials such as Type 321 and Type 347 austenitic stainless steels, Modified 9Cr-1Mo steel for core support structure construction, and Alloy 718 for Threaded Structural Fasteners were among the recommended materials for inclusion in the Code Case. This Task 4 Report identifies the need to address design life beyond 3 x 105 hours, especially in consideration of 60-year design life. A proposed update to the latest Code Case N-201 revision (i.e., Code Case N-201-5) including the items resolved in this report is included as Appendix A.

  18. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is inadequate to permit steady-state operation at reasonable conditions. 4. To enable the HTTF to operate at a more representative steady-state conditions, DOE recently allocated funding via a DOE subcontract to HTTF to permit an OSU infrastructure upgrade such that 2.2 MW will become available for HTTF experiments. 5. Analyses have been performed to study the relationship between HTTF and MHTGR via the hierarchical two-tiered scaling methodology which has been used successfully in the past, e.g., APEX facility scaling to the Westinghouse AP600 plant. These analyses have focused on the relationship between key variables that will be measured in the HTTF to the counterpart variables in the MHTGR with a focus on natural circulation, using nitrogen as a working fluid, and core heat transfer. 6. Both RELAP5-3D and computational fluid dynamics (CD-Adapco’s STAR-CCM+) numerical models of the MHTGR and the HTTF have been constructed and analyses are underway to study the relationship between the reference reactor and the HTTF. The HTTF is presently being designed. It has ¼-scaling relationship to the MHTGR in both the height and the diameter. Decisions have been made to design the reactor cavity cooling system (RCCS) simulation as a boundary condition for the HTTF to ensure that (a) the boundary condition is well defined and (b) the boundary condition can be modified easily to achieve the desired heat transfer sink for HTTF experimental operations.

  19. Westinghouse Reactor Protection System Unavailability, 1984--1995

    SciTech Connect (OSTI)

    Eide, Steven Arvid; Calley, Michael Brennan; Gentillon, Cynthia Ann; Wierman, Thomas Edward; Rasmuson, D.; Marksberry, D.

    1999-08-01

    An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U. S. Westinghouse commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.

  20. Westinghouse Reactor Protection System Unavailability, 1984-1995

    SciTech Connect (OSTI)

    C. D. Gentillon; D. Marksberry (USNRC); D. Rasmuson; M. B. Calley; S. A. Eide; T. Wierman (INEEL)

    1999-08-01

    An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U.S. Westinghouse commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.

  1. Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

    SciTech Connect (OSTI)

    Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

    2011-04-06

    Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

  2. Sealed Battery Block Provided With A Cooling System

    DOE Patents [OSTI]

    Verhoog, Roelof (Bordeaux, FR); Barbotin, Jean-Loup (Pompignac, FR)

    1999-11-16

    The present invention relates to a sealed battery block operating at a pressure of at least 1 bar relative, the battery including a container made of a plastics material and made up of a lid and of a case subdivided into wells by at least one partition, said battery being provided with a cooling system including two cheek plates made of a plastics material and co-operating with the outside faces of respective ones of two opposite walls of said case, each cheek plate co-operating with the corresponding wall to define a compartment provided with a plurality of ribs forming baffles for fluid flow purposes, and with an inlet orifice and an outlet orifice for the fluid, said battery being characterized in that each of said ribs extends in a direction that forms an angle relative to the plane of said partition lying in the range 60.degree. to 90.degree..

  3. Thermochemically recuperated and steam cooled gas turbine system

    DOE Patents [OSTI]

    Viscovich, P.W.; Bannister, R.L.

    1995-07-11

    A gas turbine system is described in which the expanded gas from the turbine section is used to generate the steam in a heat recovery steam generator and to heat a mixture of gaseous hydrocarbon fuel and the steam in a reformer. The reformer converts the hydrocarbon gas to hydrogen and carbon monoxide for combustion in a combustor. A portion of the steam from the heat recovery steam generator is used to cool components, such as the stationary vanes, in the turbine section, thereby superheating the steam. The superheated steam is mixed into the hydrocarbon gas upstream of the reformer, thereby eliminating the need to raise the temperature of the expanded gas discharged from the turbine section in order to achieve effective conversion of the hydrocarbon gas. 4 figs.

  4. Thermochemically recuperated and steam cooled gas turbine system

    DOE Patents [OSTI]

    Viscovich, Paul W. (Longwood, FL); Bannister, Ronald L. (Winter Springs, FL)

    1995-01-01

    A gas turbine system in which the expanded gas from the turbine section is used to generate the steam in a heat recovery steam generator and to heat a mixture of gaseous hydrocarbon fuel and the steam in a reformer. The reformer converts the hydrocarbon gas to hydrogen and carbon monoxide for combustion in a combustor. A portion of the steam from the heat recovery steam generator is used to cool components, such as the stationary vanes, in the turbine section, thereby superheating the steam. The superheated steam is mixed into the hydrocarbon gas upstream of the reformer, thereby eliminating the need to raise the temperature of the expanded gas discharged from the turbine section in order to achieve effective conversion of the hydrocarbon gas.

  5. Comparative Study Between Air-Cooled and Water-Cooled Condensers of the Air-Conditioning Systems 

    E-Print Network [OSTI]

    Maheshwari, G. P.; Mulla Ali, A. A.

    2004-01-01

    The weather in Kuwait is very dry where the dry-bulb temperature exceeds the wet-bulb temperature more than 20oC in most of the summer months. Thus, the air-conditioning (A/C) system with the water-cooled (WC) condensers is expected to perform more...

  6. Cooling system having reduced mass pin fins for components in a gas turbine engine

    DOE Patents [OSTI]

    Lee, Ching-Pang; Jiang, Nan; Marra, John J

    2014-03-11

    A cooling system having one or more pin fins with reduced mass for a gas turbine engine is disclosed. The cooling system may include one or more first surfaces defining at least a portion of the cooling system. The pin fin may extend from the surface defining the cooling system and may have a noncircular cross-section taken generally parallel to the surface and at least part of an outer surface of the cross-section forms at least a quartercircle. A downstream side of the pin fin may have a cavity to reduce mass, thereby creating a more efficient turbine airfoil.

  7. Performance Evaluation for a Modular, Scalable Passive Cooling System in Data Centers

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01

    3. Modius OpenData ® Data Center infrastructure Manager,Passive Cooling System in Data Centers Final Report To TheDATA CENTERS

  8. Performance Evaluation for Modular, Scalable Liquid-Rack Cooling Systems in Data Centers

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01

    3. Modius OpenData ® Data Center infrastructure Manager,Rack Cooling Systems in Data Centers Final Report To TheDATA CENTERS ..

  9. Performance Evaluation for Modular, Scalable Overhead Cooling Systems In Data Centers

    E-Print Network [OSTI]

    Xu, TengFang T.

    2009-01-01

    3. Modius OpenData ® Data Center infrastructure Manager,Overhead Cooling Systems In Data Centers Final Report To TheDATA CENTERS ..

  10. A CLASSIFICATION SCHEME FOR THE COMMON PASSIVE AND HYBRID HEATING AND COOLING SYSTEMS

    E-Print Network [OSTI]

    Holtz, Michael J.

    2011-01-01

    INTRODUCTION A passive building design attempts, within eco-and features of the building, the passive system componentsenergy. Passive cooling also uses elements of the building

  11. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  12. Reactor protection system with automatic self-testing and diagnostic

    DOE Patents [OSTI]

    Gaubatz, Donald C. (Cupertino, CA)

    1996-01-01

    A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically "identical" values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic.

  13. Reactor protection system with automatic self-testing and diagnostic

    DOE Patents [OSTI]

    Gaubatz, D.C.

    1996-12-17

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ``identical`` values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs.

  14. Cooling system for a bearing of a turbine rotor

    DOE Patents [OSTI]

    Schmidt, Mark Christopher (Niskayuna, NY)

    2002-01-01

    In a gas turbine, a bore tube assembly radially inwardly of an aft bearing conveys cooling steam to the buckets of the turbine and returns the cooling steam to a return. To cool the bearing and thermally insulate the bearing from the cooling steam paths, a radiation shield is spaced from the bore tube assembly by a dead air gap. Additionally, an air passageway is provided between the radiation shield and the inner surface of an aft shaft forming part of the rotor. Air is supplied from an inlet for flow along the passage and radially outwardly through bores in the aft shaft disk to cool the bearing and insulate it from transfer of heat from the cooling steam.

  15. Modeling and Dynamic Management of 3D Multicore Systems with Liquid Cooling

    E-Print Network [OSTI]

    Simunic, Tajana

    Modeling and Dynamic Management of 3D Multicore Systems with Liquid Cooling Ayse K. Coskun , Jos liquid cooling. Furthermore, for systems capable of varying the coolant flow rate at runtime, our University of Madrid, Spain. Embedded Systems Laboratory (ESL), Ecole Polytechnique F´ed´erale de Lausanne

  16. Improving Cooling performance of the mechanical resonator with the two-level-system defects

    E-Print Network [OSTI]

    Tian Chen; Xiang-Bin Wang

    2014-06-03

    We study cooling performance of a realistic mechanical resonator containing defects. The normal cooling method through an optomechanical system does not work efficiently due to those defects. We show by employing periodical $\\sigma_z$ pulses, we can eliminate the interaction between defects and their surrounded heat baths up to the first order of time. Compared with the cooling performance of no $\\sigma_z$ pulses case, much better cooling results are obtained. Moreover, this pulse sequence has an ability to improve the cooling performance of the resonator with different defects energy gaps and different defects damping rates.

  17. Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Digby Macdonald; Mirna Urquidi-Macdonald; Yingzi Chen; Jiahe Ai; Pilyeon Park; Han-Sang Kim

    2006-12-12

    Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding and other components in the heat transport circuits in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the accumulation of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order to accurately describe the environment that is contact with the Zircaloy cladding. In the current report, we have derived expressions for the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing Zircaloy oxidation. The models are “deterministic” because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faraday’s law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is evidence that they control the rate of the overall passive film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film. Controlled hydrodynamic methods have been used to measure important parameters in the modified Point Defect Model (PDM), which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing in BWRs and PWRs coolant environments. The modified PDMs recognize the existence of a thick oxide outer layer over a thin barrier layer. From thermodynamic analysis, it is postulated that a hydride barrier layer forms under PWR coolant conditions whereas an oxide barrier layer forms under BWR primary coolant conditions. Thus, the introduction of hydrogen into the solution lowers the corrosion potential of zirconium to the extent that the formation of ZrH2 is predicted to be spontaneous rather than the ZrO2. Mott-Schottky analysis shows that the passive film formed on zirconium is n-type, which is consistent with the PDM, corresponding to a preponderance of oxygen/hydrogen vacancies and/or zirconium interstitials in the barrier layer. The model parameter values were extracted from electrochemical impedance spectroscopic data for zirconium in high temperature, de-aerated and hydrogenated environments by optimization. The results indicate that the corrosion resistance of zirconium is dominated by the porosity and thickness of the outer layer for both cases. The impedance model based on the PDM provides a good account of the growth of the bi-layer passive films described above, and the extracted model parameter values might be used, for example, for predicting the accumulation of general corrosion damage to Zircaloy fuel sheath in BWR and PWR operating environments. Transients in current density and film thickness for passive film formation on zirconium in dearated and hydrogenated coolant conditions have confirmed that the rate law afforded by the Point Defect Model (PDM) adequately describes the growth and thinning of the passive film. The experimental results demonstrate that the kinetics of oxygen or hydrogen vacancy generation at the metal/film interface control the rate of film growth, when the potential is displaced in the positive direction, whereas the kinetics of dissolution of the barrier layer at the barrier layer/solution interface control the rate of passive film thinning when the potential is stepped in the negative direction. In addition, the effects of second phase particles (SPPs) on the electrochemistry of passive zirconium in the

  18. Federspiel Controls’ Data Center Energy Efficient Cooling Control System

    SciTech Connect (OSTI)

    2011-05-31

    Fact sheet about combining artificial intelligence with variable flow control, direct temperature measurement, and best practices that can reduce cooling energy use by up to 50%.

  19. Cool Roof Systems; What is the Condensation Risk?

    SciTech Connect (OSTI)

    Kehrer, Manfred [ORNL; Pallin, Simon B [ORNL

    2014-01-01

    A white roof, or cool roof, is constructed to decrease thermal loads from solar radiation, therefore saving energy by decreasing the cooling demands. Unfortunately, cool roofs with a mechanically attached membrane have shown a higher risk of intermediate condensation in the materials below the membrane in certain climates (Ennis & Kehrer, 2011) and in comparison with similar constructions with a darker exterior surface (Bludau, Zirkelbach, & Kuenzel, 2009). As a consequence, questions have been raised regarding the sustainability and reliability of using cool roof membranes in northern U.S. climate zones.

  20. Air-cooled Condensers in Next-generation Conversion Systems

    Office of Energy Efficiency and Renewable Energy (EERE)

    DOE Geothermal Program Peer Review 2010 - Presentation. Project objective: to reduce the costs associated with the generation of electrical power from air-cooled binary plants.

  1. Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

    SciTech Connect (OSTI)

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  2. Fast cooling for a system of stochastic oscillators

    E-Print Network [OSTI]

    Yongxin Chen; Tryphon Georgiou; Michele Pavon

    2015-06-30

    We study feedback control of coupled nonlinear stochastic oscillators in a force field. We first consider the problem of asymptotically driving the system to a desired {\\em steady state} corresponding to reduced thermal noise. Among the feedback controls achieving the desired asymptotic transfer, we find that the most efficient one {from an energy point of view} is characterized by {\\em time-reversibility}. We also extend the theory of Schr\\"{o}dinger bridges to this model, thereby steering the system in {\\em finite time} and with minimum effort to a target steady-state distribution. The system can then be maintained in this state through the optimal steady-state feedback control. The solution, in the finite-horizon case, involves a space-time harmonic function $\\varphi$, and $-\\log\\varphi$ plays the role of an artificial, time-varying potential in which the desired evolution occurs. This framework appears extremely general and flexible and can be viewed as a considerable generalization of existing active control strategies such as macromolecular cooling. In the case of a quadratic potential, the results assume a form particularly attractive from the algorithmic viewpoint as the optimal control can be computed via deterministic matricial differential equations. An example involving inertial particles illustrates both transient and steady state optimal feedback control.

  3. Three-dimensional neutronics optimization of helium-cooled blanket for multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)

    SciTech Connect (OSTI)

    Jiang, J.; Yuan, B.; Jin, M.; Wang, M.; Long, P.; Hu, L.

    2012-07-01

    Three-dimensional neutronics optimization calculations were performed to analyse the parameters of Tritium Breeding Ratio (TBR) and maximum average Power Density (PDmax) in a helium-cooled multi-functional experimental fusion-fission hybrid reactor named FDS (Fusion-Driven hybrid System)-MFX (Multi-Functional experimental) blanket. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this contribution, the most significant and main goal of the FDS-MFX blanket is to achieve the PDmax of about 100 MW/m3 with self-sustaining tritium (TBR {>=} 1.05) based on the second-stage test with uranium-fueled blanket to check and validate the demonstrator reactor blanket relevant technologies based on the viable fusion and fission technologies. Four different enriched uranium materials were taken into account to evaluate PDmax in subcritical blanket: (i) natural uranium, (ii) 3.2% enriched uranium, (iii) 19.75% enriched uranium, and (iv) 64.4% enriched uranium carbide. These calculations and analyses were performed using a home-developed code VisualBUS and Hybrid Evaluated Nuclear Data Library (HENDL). The results showed that the performance of the blanket loaded with 64.4% enriched uranium was the most attractive and it could be promising to effectively obtain tritium self-sufficiency (TBR-1.05) and a high maximum average power density ({approx}100 MW/m{sup 3}) when the blanket was loaded with the mass of {sup 235}U about 1 ton. (authors)

  4. Ground state cooling of quantum systems via a one-shot measurement

    E-Print Network [OSTI]

    P. V. Pyshkin; Da-Wei Luo; J. Q. You; Lian-Ao Wu

    2015-03-13

    We prove that there exists a family of quantum systems that can be cooled to their ground states by a one-shot projective measurement on the ancillas coupled to these systems. Consequently, this proof gives rise to the conditions for achieving the one-shot measurement ground-state cooling (OSMGSC). We also propose a general procedure for finding unitary propagators and corresponding Hamiltonians to realize such cooling by means of inverse engineering technique.

  5. System and method for regulating EGR cooling using a rankine cycle

    DOE Patents [OSTI]

    Ernst, Timothy C.; Morris, Dave

    2015-12-22

    This disclosure relates to a waste heat recovery (WHR) system and method for regulating exhaust gas recirculation (EGR) cooling, and more particularly, to a Rankine cycle WHR system and method, including a recuperator bypass arrangement to regulate EGR exhaust gas cooling for engine efficiency improvement and thermal management. This disclosure describes other unique bypass arrangements for increased flexibility in the ability to regulate EGR exhaust gas cooling.

  6. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    SciTech Connect (OSTI)

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  7. Reactor control system upgrade for the McClellan Nuclear Radiation Center

    E-Print Network [OSTI]

    Power, Michael A.

    1999-01-01

    a new reactor control system for the McClellan NuclearI REACTOR CONTROL SYSTEM UPGRADE FOR THE McCLELLAN NUCLEARReactor Control System Upgrade for the McClellan Nuclear

  8. Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

    SciTech Connect (OSTI)

    DeVault, G.P.; Bell, C.R.

    1985-01-01

    Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed.

  9. Design, analysis and optimization of the power conversion system for the Modular Pebble Bed Reactor System

    E-Print Network [OSTI]

    Wang, Chunyun, 1968-

    2003-01-01

    The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a GenIV nuclear system. The availability of controllable ...

  10. FULLY INTEGRATED ONE PHASE LIQUID COOLING SYSTEM FOR ORGANIC BOARDS

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    cooling circuit consists of a cool plate for the power components, a heat exchanger to reject the heat and fabrication of micro-channels have been Power Components on µ-Channel - Structure Reservoir Heat Exchanger. It is highlighted that for such a concept a special type of membrane pump with adequate valve technology

  11. NEUTRON ACTIVATION COOL-DOWN OF THE TOKAMAK FUSION TEST REACTOR

    E-Print Network [OSTI]

    maintenance, reliability, and performance requirements. Fig. 1 shows a partial schematic plan view of the TFTR system pump ducts, and the activation measurements. The characteristics of the Test Cell shielding

  12. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01

    of conventional LWR systems (PWR & BWRs), partly due to thethe margin to boiling in a PWR is ?15 ? C, while the coolantprimary heat exhangers of a PWR, in which borated water is

  13. Air cooled turbine component having an internal filtration system

    DOE Patents [OSTI]

    Beeck, Alexander R. (Orlando, FL)

    2012-05-15

    A centrifugal particle separator is provided for removing particles such as microscopic dirt or dust particles from the compressed cooling air prior to reaching and cooling the turbine blades or turbine vanes of a turbine engine. The centrifugal particle separator structure has a substantially cylindrical body with an inlet arranged on a periphery of the substantially cylindrical body. Cooling air enters centrifugal particle separator through the separator inlet port having a linear velocity. When the cooling air impinges the substantially cylindrical body, the linear velocity is transformed into a rotational velocity, separating microscopic particles from the cooling air. Microscopic dust particles exit the centrifugal particle separator through a conical outlet and returned to a working medium.

  14. Cooling laser system for quantum computing with barium-137 ions Tom Chartrand

    E-Print Network [OSTI]

    Blinov, Boris

    Cooling laser system for quantum computing with barium-137 ions Tom Chartrand Department of Physics of a powerful 493 nm laser source for cooling, by the resonant frequency doubling of a 986 nm external cavity diode lasers (ECDL), in a bowtie cavity with stabilization system. I. INTRODUCTION Trapped ions

  15. CFD Simulation and Analysis of the Combined Evaporative Cooling and Radiant Ceiling Air-conditioning System 

    E-Print Network [OSTI]

    Xiang, H.; Yinming, L.; Junmei, W.

    2006-01-01

    Due to such disadvantages as large air duct and high energy consumption of the current all- outdoor air evaporative cooling systems used in the dry region of Northwest China, as well as the superiority of the ceiling cooling system in improving...

  16. A Protocol For Cooling and Controlling Composite Systems by Local Interactions

    E-Print Network [OSTI]

    Daniel Burgarth; Vittorio Giovannetti

    2008-11-14

    We discuss an explicit protocol which allows one to externally cool and control a composite system by operating on a small subset of it. The scheme permits to transfer arbitrary and unknown quantum states from a memory on the network ("upload access") as well as the inverse ("download access"). In particular it yields a method for cooling the system.

  17. ENGINEERING DESIGN OF A LIQUID METAL COOLED SELF-PUMPED LIMITER FOR A TOKAMAK REACTOR*

    E-Print Network [OSTI]

    Harilal, S. S.

    cladding on the front face and leading edges for sputtering control. Up to ~ 3cm of vanadium trapping of a calcium oxide electrical insulator which coats the limiter coolant channels to reduce MHD pressure drops The Tokamak Power Systems Studies (TPSS) goal was to explore and develop ideas that would lead to improvements

  18. Space-reactor electric systems: subsystem technology assessment

    SciTech Connect (OSTI)

    Anderson, R.V.; Bost, D.; Determan, W.R.

    1983-03-29

    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

  19. Electromagnetially-induced-transparency-like ground-state cooling in a double-cavity optomechanical system

    E-Print Network [OSTI]

    Yujie Guo; Kai Li; Wenjie Nie; Yong Li

    2014-07-19

    We propose to cool a mechanical resonator close to its ground state via an electromagnetically-induced-transparency- (EIT-) like cooling mechanism in a double-cavity optomechanical system, where an additional cavity couples to the original one in the standard optomechanical system. By choosing optimal parameters such that the cooling process of the mechanical resonator corresponds to the maximum value of the optical fluctuation spectrum and the heating process to the minimum one, the mechanical resonator can be cooled with the final mean phonon number less than that at the absence of the additional cavity. And we show the mechanical resonator may be cooled close to its ground state via such an EIT-like cooling mechanism even when the original resolved sideband condition is not fulfilled at the absence of the additional cavity.

  20. Westinghouse Small Modular Reactor passive safety system response to postulated events

    SciTech Connect (OSTI)

    Smith, M. C.; Wright, R. F. [Westinghouse Electric Company, 600 Cranberry Woods Drive (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to reduce the reactor pressure in a controlled manner to facilitate the passive injection. Long-term decay heat removal is accomplished using the passive heat removal systems augmented by heat transfer through the containment vessel to the environment. The passive injection systems are designed so that the fuel remains covered and effectively cooled throughout the event. Like during the frequent faults, the passive systems provide effective cooling without the need for ac power for seven days following the accident. Connections are available to add additional water to indefinitely cool the plant. The response of the safety systems of the Westinghouse SMR to various initiating faults has been examined. Among them, two accidents; an extended station blackout event, and a LOCA event have been evaluated to demonstrate how the plant will remain safe in the unlikely event that either should occur. (authors)