Sample records for reactors 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)

    1991-01-01T23:59:59.000Z

    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, Franklin E. (San Jose, CA)

    1992-01-01T23:59:59.000Z

    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.

  3. Reactor core isolation cooling system

    DOE Patents [OSTI]

    Cooke, F.E.

    1992-12-08T23:59:59.000Z

    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.

  4. Liquid metal cooled nuclear reactor plant system

    DOE Patents [OSTI]

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

    1993-01-01T23:59:59.000Z

    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)

    1990-01-01T23:59:59.000Z

    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)

    1991-01-01T23:59:59.000Z

    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. 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)

    1991-01-01T23:59:59.000Z

    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.

  8. 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)

    1993-01-01T23:59:59.000Z

    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.

  9. 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)

    1992-01-01T23:59:59.000Z

    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.

  10. 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)

    1985-01-01T23:59:59.000Z

    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.

  11. Passive cooling system for nuclear reactor containment structure

    DOE Patents [OSTI]

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

    1989-01-01T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

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

    1990-01-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    systems for the Gas Cooled Fast Reactor (GCFR) includes theThey are 1) gas cooled fast reactors (GFR), 2) very high

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

    SciTech Connect (OSTI)

    T. E. Wierman

    2013-10-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2013-01-16T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    DeVolpi, Alexander (Bolingbrook, IL)

    1987-01-01T23:59:59.000Z

    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.

  17. 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.

    1980-06-06T23:59:59.000Z

    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.

  18. 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.

    2014-07-14T23:59:59.000Z

    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 stead-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 stead-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.

  19. Challenges and Innovative Technologies On Fuel Handling Systems for Future Sodium-Cooled Fast Reactors

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    , AREVA, and EDF have an extensive experience and significant expertise in sodium-cooled fast reactorsChallenges and Innovative Technologies On Fuel Handling Systems for Future Sodium-Cooled Fast Reactors Mathieu CHASSIGNET1;Ã , Sebastien DUMAS1 , Christophe PENIGOT1 , Ge´rard PRELE2 , Alain CAPITAINE2

  20. Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System

    SciTech Connect (OSTI)

    Angelo Frisani; Yassin A. Hassan; Victor M. Ugaz

    2010-11-02T23:59:59.000Z

    The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations. The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls' temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the realizable k-epsilon model with two-layer all y+ wall treatment performs better than the other k-epsilon and k-omega turbulence models when compared to the experimental results and the Reynolds stress transport turbulence model results. A scaling analysis was developed to address the distortions introduced by the CFD model in simulating the physical phenomena inside the RCCS system with respect to the full plant configuration. The scaling analysis demonstrated that both the experimental facility and the CFD model achieve a satisfactory resemblance of the main flow characteristics inside the RCCS cavity region, and convection and radiation heat exchange phenomena are properly scaled from the actual plant.

  1. A Computer Program Predicting Steady-State Performance of a Nuclear Research Reactor's Cooling System

    SciTech Connect (OSTI)

    Kamel Sidi Ali [Nuclear Research Center of Birine (Algeria)

    2002-07-01T23:59:59.000Z

    The performances of a nuclear reactor are directly affected by its cooling system, especially when it uses wet towers to evacuate the heat generated in the nuclear reactor core. Failure of the cooling system can yield very serious damages to most of the components of the nuclear reactor core. In this work, a computer program simulating the thermal behavior of a nuclear research reactor's cooling system is presented. Starting from the proposed start-up data of the reactor, the program predicts the cooling capacity of the nuclear reactor while taking into account the current climate conditions and also monitors the behavior of the thermal equipment involved in this process and this for different levels of power. The proposed simulation is based on a set of heat transfer equations representing all the equipment making up the cooling system up to the nuclear reactor core. Owing to the proposed inter-connected set of equations used to predict the thermal behaviour of the system, this program allows the user to modify at will a specified parameter and study the induced resulting effects on the rest of the system. The computer program developed has been experimentally validated on an operational system generating 6.8 MW and the obtained results are in good agreement with experiment. The results produced by the program concern the capacity of the cooling system to evacuate all the heat generated in the nuclear reactor core while taking into account the current climate conditions, the determination of the optimal number of thermal equipment that need to be engaged, the monitoring of the reactor core's entry end exit temperatures as well as the temperatures of all the components of the cooling system. Moreover, the program gives all the characteristics of air at the exit of the cooling towers and the loss of water due to the cooling process. (authors)

  2. CFD analyses of natural circulation in the air-cooled reactor cavity cooling system

    SciTech Connect (OSTI)

    Hu, R. [Nuclear Engineering Division, Argonne National Laboratory, Argonne IL (United States); Pointer, W. D. [Reactor and Nuclear Systems Division, Oak Ridge National Laboratory, Oak Ridge TN (United States)

    2013-07-01T23:59:59.000Z

    The Natural Convection Shutdown Heat Removal Test Facility (NSTF) is currently being built at Argonne National Laboratory, to evaluate the feasibility of the passive Reactor Cavity Cooling System (RCCS) for Next Generation Nuclear Plant (NGNP). CFD simulations have been applied to evaluate the NSTF and NGNP RCCS designs. However, previous simulations found that convergence was very difficult to achieve in simulating the complex natural circulation. To resolve the convergence issue and increase the confidence of the CFD simulation results, additional CFD simulations were conducted using a more detailed mesh and a different solution scheme. It is found that, with the use of coupled flow and coupled energy models, the convergence can be greatly improved. Furthermore, the effects of convection in the cavity and the effects of the uncertainty in solid surface emissivity are also investigated. (authors)

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

    DOE Patents [OSTI]

    Hunsbedt, A.

    1996-03-12T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA)

    1996-01-01T23:59:59.000Z

    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.

  5. 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

    2012-06-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    1] B. Farrar et. al. , Fast reactor decay heat removal:CA [2] B. Farrar et. al. , Fast reactor decay heat removal:They are 1) gas cooled fast reactors (GFR), 2) very high

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    passive safety cooling systems. To develop an understandingthe passive safety cooling system and recommend an approachof Passive Safety Cooling Systems for Advanced Nuclear

  8. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air

    SciTech Connect (OSTI)

    Corradin, Michael; Hassan, Yassin; Tokuhiro, Akira

    2014-10-15T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1991-01-01T23:59:59.000Z

    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.

  10. 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)

    1993-01-01T23:59:59.000Z

    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.

  11. 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.

    1993-12-14T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Whitman, Joshua (Joshua J.)

    2007-01-01T23:59:59.000Z

    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 ...

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

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Hill, P.R.

    1994-12-27T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Hill, Paul R. (Tucson, AZ)

    1994-01-01T23:59:59.000Z

    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.

  16. An investigation of RVACS (reactor vessel auxiliary cooling system) design improvements

    SciTech Connect (OSTI)

    Tzanos, C.P.; Tessier, J.H.; Pedersen, D.R. (Argonne National Laboratory, IL (USA))

    1989-11-01T23:59:59.000Z

    One of the main safety features of the current liquid-metal reactor (LMR) designs is the utilization of decay heat removal systems that remove heat by natural convection. In the reactor vessel auxiliary cooling system (RVACS), decay heat is removed by naturally circulating air in the gap between the guard vessel and a baffle wall surrounding the guard vessel. The objective of this work was to determine the impact of a number of design parameters on the performance of the RVACS of a pool LMR. These parameters were (a) the stack height, (b) the size of the airflow gap, (c) the system pressure loss, (d) fins on the guard vessel or the baffle wall, and (e) roughness (in the form of repeated ribs) on the airflow channel walls. Reactor designs ranging from 400 to 3,500 MW(thermal) were considered. From the RVACS design parameters considered in this analysis, an optimized ribbed configuration gave the best improvement in RVACS performance. For a 3,500-MW(thermal) LMR, the peak sodium and cladding temperatures were reduced by 52 K.

  17. 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

    2013-08-08T23:59:59.000Z

    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...

  18. UPDATE ON SMALL MODULAR REACTORS DYNAMIC SYSTEM MODELING TOOL Molten Salt Cooled Architecture

    SciTech Connect (OSTI)

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

    2014-08-01T23:59:59.000Z

    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.

  19. Passive containment cooling system

    DOE Patents [OSTI]

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

    1991-01-01T23:59:59.000Z

    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.

  20. Cooling water distribution system

    DOE Patents [OSTI]

    Orr, Richard (Pittsburgh, PA)

    1994-01-01T23:59:59.000Z

    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.

  1. 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

    2011-08-08T23:59:59.000Z

    the VHTR performance and safety analysis, one-dimensional (1-D) system type codes, like RELAP5 or MELCOR, and multi-dimensional CFD codes can be used. The choice of 1-D over multi-dimensional codes first involves identifying the main phenomena, and from...

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

    DOE Patents [OSTI]

    DeVolpi, A.

    1984-07-20T23:59:59.000Z

    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.

  3. 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

    2003-09-01T23:59:59.000Z

    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.

  4. 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

    2009-03-01T23:59:59.000Z

    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.

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

    Broader source: Energy.gov [DOE]

    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...

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

    E-Print Network [OSTI]

    Gibbs, Jonathan Paul

    2008-01-01T23:59:59.000Z

    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. Reactor water cleanup system

    DOE Patents [OSTI]

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

    1994-12-20T23:59:59.000Z

    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.

  8. MHTGR (modular high-temperature gas-cooled reactor) control: A non-safety related system

    SciTech Connect (OSTI)

    Rodriguez, C.; Swart, F.

    1988-06-01T23:59:59.000Z

    The modular high-temperature gas-cooled reactor (MHTGR) design meets stringent top-level safety regulatory criteria and user requirements that call for high plant availability and no disruption of the public's day to day activities during normal and off-normal operation of the plant. These requirements lead to a plant design that relies mainly on physical properties and passive design features to ensure plant safety regardless of operator actions, plus simplicity and automation to ensure high plant availability and lower cost of operations. The plant does not require safety-related operator actions, and it does not require the control room to be safety related.

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

    SciTech Connect (OSTI)

    Not Available

    1980-05-01T23:59:59.000Z

    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;

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    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,

  11. Gas-cooled nuclear reactor

    DOE Patents [OSTI]

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

    1985-01-01T23:59:59.000Z

    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.

  12. Emergency core cooling system

    DOE Patents [OSTI]

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

    1983-01-01T23:59:59.000Z

    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.

  13. Stability analysis of supercritical water cooled reactors

    E-Print Network [OSTI]

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

    2005-01-01T23:59:59.000Z

    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 ...

  14. 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)

    2006-01-20T23:59:59.000Z

    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.

  15. RAMI Analysis for Designing and Optimizing Tokamak Cooling Water System (TCWS) for the ITER's Fusion Reactor

    SciTech Connect (OSTI)

    Ferrada, Juan J [ORNL] [ORNL; Reiersen, Wayne T [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

    U.S.-ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment. TCWS transfers heat generated in the Tokamak during nominal pulsed operation - 850 MW at up to 150 C and 4.2 MPa water pressure. Impurities are diffused from in-vessel components and the vacuum vessel by water baking at 200-240 C at up to 4.4 MPa. TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER's success and interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation, each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements. RAMI analysis will indicate appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. This analysis helps demonstrate that using proven, commercially available equipment is better than using custom-designed equipment with no field experience and lowers specific costs while providing higher reliability. This paper presents a brief description of the TCWS conceptual design and the application of RAMI tools to optimize the design at different stages during the project.

  16. Coupled Reactor Kinetics and Heat Transfer Model for Heat Pipe Cooled Reactors

    SciTech Connect (OSTI)

    WRIGHT,STEVEN A.; HOUTS,MICHAEL

    2000-11-22T23:59:59.000Z

    Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). The paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities.

  17. 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-

    2004-01-01T23:59:59.000Z

    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, ...

  18. 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

    2014-01-01T23:59:59.000Z

    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...

  19. A resting bottom sodium cooled fast reactor

    SciTech Connect (OSTI)

    Costes, D. [Consultant (France)

    2012-07-01T23:59:59.000Z

    This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

  20. SRS reactor control rod cooling without normal forced convection cooling

    SciTech Connect (OSTI)

    Smith, D.C. (SAIC, Albuquerque, NM (United States)); Easterling, T.C. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1993-01-01T23:59:59.000Z

    This paper describes an analytical study of the coolability of the control rods in the Savannah River site (SRS) K production reactor under conditions of loss of normal forced convection cooling. The study was performed as part of the overall safety analysis of the reactor supporting its restart. The analysis addresses the buoyancy-driven boiling flow over the control rods that occurs when forced cooling is lost. The objective of the study was to demonstrate that the control rods will remain cooled (i.e., no melting) at powers representative of those anticipated for restart of the reactor.

  1. Passive containment cooling system

    DOE Patents [OSTI]

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

    1994-01-25T23:59:59.000Z

    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.

  2. 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)

    1994-01-01T23:59:59.000Z

    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.

  3. Space power reactor ground test in the Experimental Gas Cooled Reactor (EGCR) at Oak Ridge

    SciTech Connect (OSTI)

    Fontana, M.H.; Holcomb, R.S.; Cooper, R.H.

    1992-08-01T23:59:59.000Z

    The Experimental Gas Cooled Reactor (EGCR) facility and the supporting technical infrastructure at the Oak Ridge National Laboratory have the capabilities of performing ground tests of space nuclear power reactor systems. A candidate test would be a 10 MWt lithium cooled reactor, generating potassium vapor that would drive a power turbine. The facility is a large containment vessel originally intended to test the EGCR. Large, contained, and shielded spaces are available for testing, assembly, disassembly, and post-test examination.

  4. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, R.M.

    1983-11-08T23:59:59.000Z

    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.

  5. 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)

    2012-07-01T23:59:59.000Z

    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)

  6. 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

    2012-07-16T23:59:59.000Z

    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...

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    H. G. MacPherson The molten salt adventure Nuclear Scienceand P.F. Peterson, Molten-Salt-Cooled Advanced High-Clarno Assessment of candidate molten salt coolants for the

  8. Experimental Study of the Effect of Graphite Dispersion on the Heat Transfer Phenomena in a Reactor Cavity Cooling System

    SciTech Connect (OSTI)

    Rodolfo Vaghetto; Luigi Capone; Yassin A. Hassan

    2011-05-31T23:59:59.000Z

    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 (16.5 x 16.5 x 30.4 cm) used for this activity represents half of the reactor cavity with an electrically heated vessel. Water flowing through five vertical pipes removes the heat produced in the vessel and releases it into the environment by mixing with cold water in a large tank. The particle image velocimetry technique was used to study the velocity field of the air inside the cavity. A set of 52 thermocouples was installed in the facility to monitor the temperature profiles of the vessel, pipe walls, and air. Ten grams of a fine graphite powder (average particle size 2 m) was injected into the cavity through a spraying nozzle placed at the bottom of the vessel. The temperatures and air velocity field were recorded and compared with the measurements obtained before the graphite dispersion, showing a decrease of the temperature surfaces that was related to an increase in their emissivity. The results contribute to the understanding of RCCS capability in an accident scenario.

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

    E-Print Network [OSTI]

    Pope, Michael A. (Michael Alexander)

    2004-01-01T23:59:59.000Z

    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 ...

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    dummy>0.95) dummy=pn1/g6; ppp=ge(:,1); elseif aa==2 ge2(k,1)=ge3(:,1); subplot(3,8,1);plot(ppp,mf(:,1),'o');title('Airmflow') subplot(3,8,9);plot(ppp,mf(:,2),'o');title('cooling

  11. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    SciTech Connect (OSTI)

    Not Available

    1980-09-01T23:59:59.000Z

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases.

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    Advisory Committee and Generation IV International Forum.Nuclear Energy Agency The Generation IV International Forum.Technology Roadmap for Generation IV Nuclear Energy Systems.

  13. Cooling Water System Optimization

    E-Print Network [OSTI]

    Aegerter, R.

    2005-01-01T23:59:59.000Z

    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...

  14. 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)

    2008-01-01T23:59:59.000Z

    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 ...

  15. Medium-size high-temperature gas-cooled reactor

    SciTech Connect (OSTI)

    Peinado, C.O.; Koutz, S.L.

    1980-08-01T23:59:59.000Z

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760/sup 0/C (1400/sup 0/F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics (a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant) and engineered safety features (core auxiliary cooling, relief valve, and steam generator dump systems).

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

    E-Print Network [OSTI]

    Moore, Eugene James Thomas

    2006-08-16T23:59:59.000Z

    to material selection and reactor safety. Understanding heat transfer and fluid flow phenomena during normal and transient operation of HTGRs is essential to ensure the adequacy of safety features, such as the reactor cavity cooling system (RCCS). Modeling...

  17. Investigation of vessel exterior air cooling for a HLMC reactor

    SciTech Connect (OSTI)

    Sienicki, J. J.; Spencer, B. W.

    2000-01-13T23:59:59.000Z

    The Secure Transportable Autonomous Reactor (STAR) concept under development at Argonne National Laboratory provides a small (300 MWt) reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100%+ natural circulation heat removal from the low power density/low pressure drop ultra-long lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the Reactor Exterior Cooling System (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the Reactor Vessel Auxiliary Cooling System (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink.

  18. Investigation of vessel exterior air cooling for an HLMC reactor

    SciTech Connect (OSTI)

    Sienicki, J.J.; Spencer, B.W.

    2000-07-01T23:59:59.000Z

    The secure transportable autonomous reactor (STAR) concept under development at Argonne National Laboratory provides a small [300-MW(thermal)] reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100% + natural-circulation heat removal from the low-power-density/low-pressure-drop ultralong lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the reactor exterior cooling system (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the reactor vessel auxiliary cooling system (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink.

  19. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    SciTech Connect (OSTI)

    David E. Shropshire

    2004-04-01T23:59:59.000Z

    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.

  20. Gas turbine cooling system

    DOE Patents [OSTI]

    Bancalari, Eduardo E. (Orlando, FL)

    2001-01-01T23:59:59.000Z

    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).

  1. Candidate Materials Evaluation for Supercritical Water-Cooled Reactor

    SciTech Connect (OSTI)

    T. R. Allen and G. S. Was

    2008-12-12T23:59:59.000Z

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

  2. Hydronic rooftop cooling systems

    DOE Patents [OSTI]

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

    2008-01-29T23:59:59.000Z

    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.

  3. Superconductor rotor cooling system

    DOE Patents [OSTI]

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

    2004-11-02T23:59:59.000Z

    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.

  4. 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)

    2002-01-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

    uranium (LEU) cores. Unlike light water reactors (LWRs), the ultimate heat sink for decay heat removal

  6. Electrochemistry of Water-Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

    geothermal reservoir. Greif provides a review of the use of applications of natural circulation to the cooling

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

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

    Reactor. in Proceedings of ICAPP’10 (American NuclearFHR. in Proceedings of ICAPP’12 (American Nuclear Society,

  9. Cooling System Basics | Department of Energy

    Energy Savers [EERE]

    Homes & Buildings Space Heating & Cooling Cooling System Basics Cooling System Basics August 16, 2013 - 1:08pm Addthis Cooling technologies used in homes and buildings...

  10. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect (OSTI)

    Neil Todreas; Pavel Hejzlar

    2008-06-30T23:59:59.000Z

    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.

  11. Nuclear reactor sealing system

    DOE Patents [OSTI]

    McEdwards, James A. (Calabasas, CA)

    1983-01-01T23:59:59.000Z

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

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

    SciTech Connect (OSTI)

    Zitek, Pavel, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz; Valenta, Vaclav, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz; Klimko, Marek, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz [University of West Bohemia in Pilsen, Univerzitní 8, 306 14 Pilsen (Czech Republic)

    2014-08-06T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

    K. T. Assessment of Candidate Molten Salt Coolants for theK. T. Assessment of Candidate Molten Salt Coolants for thebeginning efforts for a molten salt reactor (MSR) program.

  14. Atmospheric Pressure Reactor System | EMSL

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

    Atmospheric Pressure Reactor System Atmospheric Pressure Reactor System The atmospheric pressure reactor system is designed for testing the efficiency of various catalysts for the...

  15. Diesel lubrication and cooling systems

    SciTech Connect (OSTI)

    NONE

    1994-12-31T23:59:59.000Z

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

  16. Diesel lubrication and cooling systems

    SciTech Connect (OSTI)

    Not Available

    1994-01-01T23:59:59.000Z

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

  17. Sodium-cooled Fast Reactor - Past and Future | Argonne National...

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

    Sodium-cooled Fast Reactor - Past and Future June 16, 2015 10:00AM to 11:00AM Presenter Taek K. Kim (NE), Principal Nuclear Engineer and Department Manager Location Building 205,...

  18. CONTROL SYSTEM FOR SOLAR HEATING and COOLING

    E-Print Network [OSTI]

    Dols, C.

    2010-01-01T23:59:59.000Z

    solar heating and cooling systems covering a wide range ofpractical heating and cooling system configurations andexperimental heating and cooling system, the main purpose of

  19. Development and Evaluation of a Safeguards System Concept for a Pebble-Fueled High Temperature Gas-cooled Reactor 

    E-Print Network [OSTI]

    Gitau, Ernest Travis Ngure

    2012-10-19T23:59:59.000Z

    the world, adequate methods for safeguarding the reactor must be developed. Current safeguards methods for the pebble-fueled HTGR focus on extensive, redundant containment and surveillance (C/S) measures or a combination of item-type and bulk-type material...

  20. Overall plant design specification Modular High Temperature Gas-cooled Reactor. Revision 9

    SciTech Connect (OSTI)

    NONE

    1990-05-01T23:59:59.000Z

    Revision 9 of the ``Overall Plant Design Specification Modular High Temperature Gas-Cooled Reactor,`` DOE-HTGR-86004 (OPDS) has been completed and is hereby distributed for use by the HTGR Program team members. This document, Revision 9 of the ``Overall Plant Design Specification`` (OPDS) reflects those changes in the MHTGR design requirements and configuration resulting form approved Design Change Proposals DCP BNI-003 and DCP BNI-004, involving the Nuclear Island Cooling and Spent Fuel Cooling Systems respectively.

  1. Lamination cooling system

    DOE Patents [OSTI]

    Rippel, Wally E.; Kobayashi, Daryl M.

    2005-10-11T23:59:59.000Z

    An electric motor, transformer or inductor having a lamination cooling system including a stack of laminations, each defining a plurality of apertures at least partially coincident with apertures of adjacent laminations. The apertures define a plurality of cooling-fluid passageways through the lamination stack, and gaps between the adjacent laminations are sealed to prevent a liquid cooling fluid in the passageways from escaping between the laminations. The gaps are sealed by injecting a heat-cured sealant into the passageways, expelling excess sealant, and heat-curing the lamination stack. The apertures of each lamination can be coincident with the same-sized apertures of adjacent laminations to form straight passageways, or they can vary in size, shape and/or position to form non-axial passageways, angled passageways, bidirectional passageways, and manifold sections of passageways that connect a plurality of different passageway sections. 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. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    fission gas plenum212 Conventional fast reactor core designGUPTA. “A Compact Gas-Cooled Fast Reactor with an Ultra-Longbreed and burn gas-cooled fast reactor”. Ph.D. Thesis. MIT,

  3. Hybrid Radiator Cooling System | Argonne National Laboratory

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

    Radiator Cooling System Technology available for licensing: Hybrid radiator cooling system uses conventional finned air cooling under most driving conditions that would be...

  4. Improved vortex reactor system

    DOE Patents [OSTI]

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

    1995-01-01T23:59:59.000Z

    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.

  5. 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 ...

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

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

    system areas, Nuclear Heat Supply and Heat Transport, areand Subsystems AREAS Nuclear Heat Supply Heat Transport Main

  8. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    SciTech Connect (OSTI)

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-09-01T23:59:59.000Z

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850ºC at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05).

  9. Deployment Scenario of Heavy Water Cooled Thorium Breeder Reactor

    SciTech Connect (OSTI)

    Mardiansah, Deby; Takaki, Naoyuki [Course of Applied Science, School of Engineering, Tokai University (Japan)

    2010-06-22T23:59:59.000Z

    Deployment scenario of heavy water cooled thorium breeder reactor has been studied. We have assumed to use plutonium and thorium oxide fuel in water cooled reactor to produce {sup 233}U which will be used in thorium breeder reactor. The objective is to analysis the potential of water cooled Th-Pu reactor for replacing all of current LWRs especially in Japan. In this paper, the standard Pressurize Water Reactor (PWR) has been designed to produce 3423 MWt; (i) Th-Pu PWR, (ii) Th-Pu HWR (MFR = 1.0) and (iii) Th-Pu HWR (MFR 1.2). The properties and performance of the core were investigated by using cell and core calculation code. Th-Pu PWR or HWR produces {sup 233}U to introduce thorium breeder reactor. The result showed that to replace all (60 GWe) LWR by thorium breeder reactor within a period of one century, Th-Pu oxide fueled PWR has insufficient capability to produce necessary amount of {sup 233}U and Th-Pu oxide fueled HWR has almost enough potential to produce {sup 233}U but shows positive void reactivity coefficient.

  10. 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

    2004-09-01T23:59:59.000Z

    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.

  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

    2012-01-01T23:59:59.000Z

    the field of chemical engineering, design of scaled systemsDesign Approach to Safety and Reliability Yields Great Benefits. Chemical Engineering

  12. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    SciTech Connect (OSTI)

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01T23:59:59.000Z

    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. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom and Switzerland), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above for this fiscal year. In addition, this report fulfills the Level 2 milestones, ''Complete annual status report on GFR reactor design'', and ''Complete annual status report on pre-conceptual GFR reactor designs'' in work package GI0401K01. GFR funding for FY05 included FY04 carryover funds, and was comprised of multiple tasks. These tasks involved a consortium of national laboratories and universities, including the Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Oak Ridge National Laboratory (ORNL), Auburn University (AU), Idaho State University (ISU), and the University of Wisconsin-Madison (UW-M). The total funding for FY05 was $1000K, with FY04 carryover of $174K. The cost breakdown can be seen in Table 1.

  13. Reactor vessel support system

    DOE Patents [OSTI]

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

    1982-01-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01T23:59:59.000Z

    2009, UCBTH09-001. ). The RELAP5-3D (c) Code DevelopmentTeam. RELAP5-3D © Code Manual Volume IV: Models andare being investigated: RELAP5-3D and Flownex. These system

  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

    2011-05-01T23:59:59.000Z

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

    DOE Patents [OSTI]

    Hunsbedt, A.; Boardman, C.E.

    1995-04-11T23:59:59.000Z

    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.

  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)

    1995-01-01T23:59:59.000Z

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

    SciTech Connect (OSTI)

    Marshall, A.C.

    1989-08-01T23:59:59.000Z

    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.

  19. A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

    SciTech Connect (OSTI)

    Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Williamson, Joshua [Advanced Nuclear Concepts Department, Sandia National Laboratories, P.O Box 5800, Albuquerque, NM 87185 (United States); Peters, Curtis D.; Brown, Nicholas [Advanced Nuclear Concepts Department, Sandia National Laboratories, P.O Box 5800, Albuquerque, NM 87185 (United States); Department of Chemical and Nuclear Engineering, University of New Mexico, Albuquerque, NM 87108 (United States); Jablonski, Jennifer [Advanced Nuclear Concepts Department, Sandia National Laboratories, P.O Box 5800, Albuquerque, NM 87185 (United States); Department of Education, University of New Mexico, Albuquerque, NM 87108 (United States)

    2005-02-06T23:59:59.000Z

    A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

  20. A gas-cooled-reactor closed-Brayton-cycle demonstration with nuclear heating.

    SciTech Connect (OSTI)

    Jablonski, Jennifer A.; Williamson, Joshua J.; Wright, Steven Alan; Dorsey, Daniel John; Brown, Nicholas; Peters, Curtis D.; Lipinski, Ronald J.

    2004-09-01T23:59:59.000Z

    A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

  1. Liquid metal reactor air cooling baffle

    DOE Patents [OSTI]

    Hunsbedt, A.

    1994-08-16T23:59:59.000Z

    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.

  2. Liquid metal reactor air cooling baffle

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA)

    1994-01-01T23:59:59.000Z

    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. Emergency cooling system and method

    DOE Patents [OSTI]

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

    1994-01-04T23:59:59.000Z

    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.

    1995-01-01T23:59:59.000Z

    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. SSTAR: The U.S. Lead-Cooled Fast Reactor (LFR)

    SciTech Connect (OSTI)

    Smith, C F; Halsey, W G; Brown, N W; Sienicki, J J; Moisseytsev, A; Wade, D C

    2007-09-25T23:59:59.000Z

    It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the Global Nuclear Energy Partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the Small Secure Transportable Autonomous Reactor (SSTAR) reactor has been under ongoing development under the U.S. Generation IV Nuclear Energy Systems Initiative. It a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation aims, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the U.S. Generation IV Lead-cooled Fast Reactor system.

  6. Improved vortex reactor system

    DOE Patents [OSTI]

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

    1995-05-09T23:59:59.000Z

    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.

  7. advanced-gas-cooled-nuclear-reactor materials evaluation: Topics...

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

    advanced-gas-cooled-nuclear-reactor materials evaluation First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index...

  8. Temperature initiated passive cooling system

    DOE Patents [OSTI]

    Forsberg, C.W.

    1994-11-01T23:59:59.000Z

    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.

  9. Temperature initiated passive cooling system

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1994-01-01T23:59:59.000Z

    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.

  10. air cooled reactors: Topics by E-print Network

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

    absorptionAssessment of adsorber bed designs in waste-heat driven adsorption cooling systems for vehicle air conditioning and refrigeration Amir Sharafian, Majid Bahrami n...

  11. Supercritical CO2Brayton Cycle Control Strategy for Autonomous Liquid Metal-Cooled Reactors

    SciTech Connect (OSTI)

    Moisseytsev, A.; Sienicki, J.J.

    2004-10-06T23:59:59.000Z

    This presentation discusses a supercritical carbon dioxide brayton cycle control strategy for autonomous liquid metal-cooled reactors.

  12. A Small Secure Transportable Autonomous Lead-Cooled Fast Reactor for Deployment at Remote Sites

    SciTech Connect (OSTI)

    Sienicki, J .J.; Smith, M.A.; Mosseytsev, A.V.; Yang, W.S.; Wade, D.C.

    2004-10-06T23:59:59.000Z

    This presentation discusses a small secure transportable autonomous lead-cooled fast reactor for deployment at remote sites.

  13. Reactor & Nuclear Systems Publications | ORNL

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

    Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications The...

  14. CONTROL SYSTEM FOR SOLAR HEATING and COOLING

    E-Print Network [OSTI]

    Dols, C.

    2010-01-01T23:59:59.000Z

    Research Applied to National Needs. EXPERIMENTAL SYSTEM A generalized system for solar heating and cooling

  15. Combustor liner cooling system

    DOE Patents [OSTI]

    Lacy, Benjamin Paul; Berkman, Mert Enis

    2013-08-06T23:59:59.000Z

    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.

  16. Evaluation of in-vessel corium retention through external reactor vessel cooling for integral reactor

    SciTech Connect (OSTI)

    Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y. [Korea Atomic Energy Research Inst., 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2012-07-01T23:59:59.000Z

    In-vessel corium retention through external reactor vessel cooling (IVR-ERVC) for a small integral reactor has been evaluated to determine the thermal margin for the prevention of a reactor vessel failure. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel has been performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall has been determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin has been evaluated by comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated at approximately 0.25 MW/m{sup 2} in the metallic layer because of the focusing effect. The CHF of the outer reactor vessel is approximately 1.1 MW/m{sup 2} because of a two phase natural circulation mass flow. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident of a small integral reactor. (authors)

  17. High-temperature gas-cooled reactor (HTGR): long term program plan

    SciTech Connect (OSTI)

    Not Available

    1980-10-09T23:59:59.000Z

    The FY 1980 effort was to investigate four technology options identified by program participants as potentially viable candidates for near-term demonstration: the Gas Turbine system (HTGR-GT), reflecting its perceived compatibility with the dry-cooling market, two systems addressing the process heat market, the Reforming (HTGR-R) and Steam Cycle (HTGR-SC) systems, and a more developmental reactor system, The Nuclear Heat Source Demonstration Reactor (NHSDR), which was to serve as a basis for both the HTGR-GT and HTGR-R systems as well as the further potential for developing advanced applications such as steam-coal gasification and water splitting.

  18. Non-intrusive cooling system

    DOE Patents [OSTI]

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

    2001-05-22T23:59:59.000Z

    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.

  19. A 50-100 kWe gas-cooled reactor for use on Mars.

    SciTech Connect (OSTI)

    Peters, Curtis D. (.)

    2006-04-01T23:59:59.000Z

    In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered.

  20. Compressor bleed cooling fluid feed system

    DOE Patents [OSTI]

    Donahoo, Eric E; Ross, Christopher W

    2014-11-25T23:59:59.000Z

    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.

  1. Solar-powered cooling system

    DOE Patents [OSTI]

    Farmer, Joseph C

    2013-12-24T23:59:59.000Z

    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.

  2. Preliminary requirements for a Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR)

    SciTech Connect (OSTI)

    Massie, M.; Forsberg, C.; Forget, B. [Dept. of Nuclear Science and Engineering, Massachusetts Inst. of Technology, Cambridge, MA 02139 (United States); Hu, L. W. [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

    2012-07-01T23:59:59.000Z

    A Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR) design is being developed at MIT to provide the first demonstration and test of a salt-cooled reactor using high-temperature fuel. The first step is to define the requirements. The top level requirements are (1) provide the confidence that a larger demonstration reactor is warranted and (2) develop the necessary data for a larger-scale reactor. Because requirements will drive the design of the FHTR, a significant effort is being undertaken to define requirements and understand the tradeoffs that will be required for a practical design. The preliminary requirements include specifications for design parameters and necessary tests of major reactor systems. Testing requirements include demonstration of components, systems, and procedures for refueling, instrumentation, salt temperature control to avoid coolant freezing, salt chemistry and volume control, tritium monitoring and control, and in-service inspection. Safety tests include thermal hydraulics, neutronics - including intrinsic core shutdown mechanisms such as Doppler feedback - and decay heat removal systems. Materials and coolant testing includes fuels (including mechanical wear and fatigue) and system corrosion behavior. Preliminary analysis indicates a thermal power output below 30 MW, an initial core using pebble-bed or prismatic-block fuel, peak outlet temperatures of at least 700 deg. C, and use of FLi{sup 7}Be ({sup 7}LiF-BeF{sub 2}) coolant. The option to change-out the reactor core, fuel type, and major components is being investigated. While the FHTR will be used for materials testing, its primary mission is as a reactor system performance test to enable the design and licensing of a FHR demonstration power reactor. (authors)

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

    E-Print Network [OSTI]

    Yarsky, Peter

    2005-01-01T23:59:59.000Z

    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 ...

  4. A better cooling water system

    SciTech Connect (OSTI)

    Gale, T.E.; Beecher, J.

    1987-12-01T23:59:59.000Z

    To prepare their newly constructed reduced crude conversion (RCC) open recirculating cooling water system for the implementation of a corrosion and deposit control water treatment program, Ashland Petroleum, Catlettsburg, Ky., made plans for and carried out precleaning and passivation procedures. Here, the authors share the results, and some potential guidelines for one's own operations. Inspection of equipment after precleaning showed that the precleaning procedures was very effective in removing undesirable matter. After precleaning and passivation of the system, the recommended cooling water treatment program was started. Corrosion rates for mild steel specimens ranged from 0.5 to 1.5 mils per year (mpy), with an average of 1.0 mpy. The corrosion rates for admiralty specimens ranged from 0.1 to 0.2 mpy. Benefits of the precleaning and passivating programs greatly outweigh the costs, since the normal cooling water treatment program for corrosion and deposit control can then operate more effectively.

  5. Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles

    SciTech Connect (OSTI)

    R. M. Ferrer; S. Bays; M. Pope

    2008-04-01T23:59:59.000Z

    The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.

  6. 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.

    2009-01-01T23:59:59.000Z

    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

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

    E-Print Network [OSTI]

    Patalano, Giovanbattista

    2007-01-01T23:59:59.000Z

    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 ...

  8. advanced reactor systems: Topics by E-print Network

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

    20 21 22 23 24 25 Next Page Last Page Topic Index 1 Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors University of California...

  9. Lamination cooling system formation method

    DOE Patents [OSTI]

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

    2012-06-19T23:59:59.000Z

    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.

  10. Lamination cooling system formation method

    DOE Patents [OSTI]

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

    2009-05-12T23:59:59.000Z

    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.

  11. Gas hydrate cool storage system

    DOE Patents [OSTI]

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

    1984-09-12T23:59:59.000Z

    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)

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

    E-Print Network [OSTI]

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

    2007-01-01T23:59:59.000Z

    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 ...

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

    E-Print Network [OSTI]

    Whitman, Joshua (Joshua J.)

    2009-01-01T23:59:59.000Z

    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 ...

  14. Cooling load differences between radiant and air systems

    E-Print Network [OSTI]

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

    2013-01-01T23:59:59.000Z

    radiant heating and cooling systems, in: Proceedings ofof radiant heating and cooling systems versus air systems,Gain on Radiant Floor Cooling System Design, in: Proceedings

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

    SciTech Connect (OSTI)

    Lee O. Nelson

    2011-04-01T23:59:59.000Z

    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.

  16. Cryo Utilities Room Cooling System

    SciTech Connect (OSTI)

    Ball, G.S.; /Fermilab

    1989-01-26T23:59:59.000Z

    Many of the mechanical equipment failures at the Laboratory are due to the loss of cooling water. In order to insure the proper operating temperatures and to increase the reliability of the mechanical equipment in the D0 Cryo Utilities Room it is necessary to provide an independent liquid cooling system. To this end, an enclosed glycoVwater cooling system which transfers heat from two vane-type vacuum pumps and an air compressor to the outside air has been installed in the Cryo Utilities Room. From the appended list it can be seen that only the Thermal Precision PFC-121-D and Ingersoll-Rand WAC 16 deserve closer investigation based on price. The disadvantages of the WAC 16 are that: it runs a little warmer, it requires more valving to properly install a backup pump, inlet and outlet piping are not included, and temperature and pressure indicators are not included. Its only advantage is that it is $818 cheaper than the PFC-121-D. The advantages of the PFC-121-D are that: it has automatic pump switching during shutdown, it has a temperature regulator on one fan control, it has a switch which indicates proper operation, has a sight glass on the expansion tank, and comes with an ASME approved expansion tank and relief valve. For these reasons the Thermal Precision PFC-121-D was chosen. In the past, we have always found the pond water to be muddy and to sometimes contain rocks of greater than 1/2 inch diameter. Thus a system completely dependent on the pond water from the accelerator was deemed unacceptable. A closed system was selected based on its ability to greatly improve reliability, while remaining economical. It is charged with a 50/50 glycol/water mixture capable of withstanding outside temperatures down to -33 F. The fluid will be circulated by a totally enclosed air cooled Thermal Precision PFC-121-D pump. The system will be on emergency power and an automatically controlled backup pump, identical to the primary, is available should the main pump fail. The fan unit is used as a primary cooler and the trim cooler cools the fluid further on extremely hot days. The trim cooler has also been sized to cool the system in the event of a total shutdown provided that the pond water supply has adequate pressure. Due to a broken filter, we found it necessary to install a strainer in the pond water supply line. The expansion tank separates air bubbles, ensures a net positive suction head, protects against surges and over pressurization of the system, and allows for the filling of the system without shutting it off. All piping has been installed, flushed, charged with the glycol/water mix, and hydrostatically tested to 55 psi. The condition of all pumps and flow conditions will be recorded at the PLC. It has been decided not to include the regulator valve in the pond water return line. This valve was designated by the manufacturer to reduce the amount of water flowing through the trim cooler. This is not necessary in our application. There is some concern that the cooling fluid may cool the mechanical eqUipment too much when they are not operating or during very cold days. This issue will be addressed and the conclusion appended to this engineering note.

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

    SciTech Connect (OSTI)

    Yoder Jr, Graydon L [ORNL] [ORNL; Aaron, Adam M [ORNL] [ORNL; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)] [University of Tennessee, Knoxville (UTK); Fugate, David L [ORNL] [ORNL; Holcomb, David Eugene [ORNL] [ORNL; Kisner, Roger A [ORNL] [ORNL; Peretz, Fred J [ORNL] [ORNL; Robb, Kevin R [ORNL] [ORNL; Wilgen, John B [ORNL] [ORNL; Wilson, Dane F [ORNL] [ORNL

    2014-01-01T23:59:59.000Z

    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.

  18. Information technology equipment cooling system

    SciTech Connect (OSTI)

    Schultz, Mark D.

    2014-06-10T23:59:59.000Z

    According to one embodiment, a system for removing heat from a rack of information technology equipment may include a sidecar indoor air to liquid heat exchanger that cools warm air generated by the rack of information technology equipment. The system may also include a liquid to liquid heat exchanger and an outdoor heat exchanger. The system may further include configurable pathways to connect and control fluid flow through the sidecar heat exchanger, the liquid to liquid heat exchanger, the rack of information technology equipment, and the outdoor heat exchanger based upon ambient temperature and/or ambient humidity to remove heat from the rack of information technology equipment.

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

    SciTech Connect (OSTI)

    Fan-Bill Cheung; Joy L. Rempe

    2004-06-01T23:59:59.000Z

    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.

  20. 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

    2008-04-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

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

    2014-01-01T23:59:59.000Z

    Embedded Radiant Heating and Cooling Systems, InternationalWATER BASED RADIANT COOLING SYSTEM DESIGN METHODS Jingjuan (Keywords: Radiant Cooling System, Design Approach,

  2. Design and Control of Hydronic Radiant Cooling Systems

    E-Print Network [OSTI]

    Feng, Jingjuan Dove

    2014-01-01T23:59:59.000Z

    embedded heating and cooling systems. Brussels, Belgium,of radiant heating/cooling systems for non-residentalSimulations of floor cooling system capacity." Applied

  3. Model Predictive Control for the Operation of Building Cooling Systems

    E-Print Network [OSTI]

    Ma, Yudong

    2010-01-01T23:59:59.000Z

    storage in building cooling systems. Technical report,storage in building cooling systems. Decision and Control,for the Operation of Building Cooling Systems Yudong Ma ? ,

  4. Cooling system for superconducting magnet

    DOE Patents [OSTI]

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

    1998-12-15T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2013-11-01T23:59:59.000Z

    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.

  6. Reactor protection system design alternatives for sodium fast reactors

    E-Print Network [OSTI]

    DeWitte, Jacob D. (Jacob Dominic)

    2011-01-01T23:59:59.000Z

    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 ...

  7. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    SciTech Connect (OSTI)

    Permana, Sidik [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Sekimoto, Hiroshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Waris, Abdul; Subhki, Muhamad Nurul [Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Ismail, [BAPETEN (Indonesia)

    2010-12-23T23:59:59.000Z

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

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

    SciTech Connect (OSTI)

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

    2010-11-01T23:59:59.000Z

    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.

  9. Evaporative cooling enhanced cold storage system

    DOE Patents [OSTI]

    Carr, P.

    1991-10-15T23:59:59.000Z

    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.

  10. Evaporative cooling enhanced cold storage system

    DOE Patents [OSTI]

    Carr, Peter (Cary, NC)

    1991-01-01T23:59:59.000Z

    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.

  11. High temperature cooling system and method

    DOE Patents [OSTI]

    Loewen, Eric P.

    2006-12-12T23:59:59.000Z

    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. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-01-01T23:59:59.000Z

    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.

  13. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-09-28T23:59:59.000Z

    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.

  14. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

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

    1980-05-09T23:59:59.000Z

    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.

  15. Gaseous reactor control system

    SciTech Connect (OSTI)

    Abdel-Khalik, S.

    1991-09-03T23:59:59.000Z

    This paper describes a nuclear reactor control system for controlling the reactivity of the core of a nuclear reactor. It includes a control gas having a high neutron cross-section; a first tank containing a first supply of the control gas; a first conduit providing a first fluid passage extending into the core, the first conduit being operatively connected to communicate with the first tank; a first valve operatively connected to regulate the flow of the control gas between the first tank and the first conduit; a second conduit concentrically disposed around the first conduit such that a second fluid passage is defined between the outer surface of the first conduit and the inner surface of the second conduit; a second tank containing a second supply of the control gas, the second tank being operatively connected to communicate with the second fluid passage; a second supply valve operatively connected to regulate the flow of the control gas between the second tank and the second fluid passage.

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

    SciTech Connect (OSTI)

    Cheng,L.Y.; Ludewig, H.

    2007-08-05T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Rodriguez, Judy N

    2013-01-01T23:59:59.000Z

    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 ...

  18. Cooling load differences between radiant and air systems

    E-Print Network [OSTI]

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

    2013-01-01T23:59:59.000Z

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

  19. High temperature gas-cooled reactor: gas turbine application study

    SciTech Connect (OSTI)

    Not Available

    1980-12-01T23:59:59.000Z

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  20. 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

    2013-08-08T23:59:59.000Z

    with nine pipes in the cavity, return and supply manifolds connecting standing pipes with water tank and a cylindrical water tank situated at top of the cavity (as shown in Figure 5). In the facility, the cylindrical reactor vessel is approximately... Simulation ......................................................................... 14 2.3.1 Water Tank as Single Volume Without Secondary Loop ............................. 14 2.3.2 Water Tank as Pipe with Secondary Loop...

  1. Containment system for supercritical water oxidation reactor

    DOE Patents [OSTI]

    Chastagner, P.

    1994-07-05T23:59:59.000Z

    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.

  2. Containment system for supercritical water oxidation reactor

    DOE Patents [OSTI]

    Chastagner, Philippe (3134 Natalie Cir., Augusta, GA 30909-2748)

    1994-01-01T23:59:59.000Z

    A system 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.

  3. Thermal Response of the Hybrid Loop-Pool Design for Sodium Cooled Faster Reactors

    SciTech Connect (OSTI)

    Zhang, Hongbin; Zhao, Haihua; Davis, Cliff

    2008-09-01T23:59:59.000Z

    An innovative hybrid loop-pool design for the sodium cooled fast reactor (SFR) has been recently proposed with the primary objective of achieving cost reduction and safety enhancement. With the hybrid loop-pool design, closed primary loops are immersed in a secondary buffer tank. This design takes advantage of features from conventional both pool and loop designs to further improve economics and safety. This paper will briefly introduce the hybrid loop-pool design concept and present the calculated thermal responses for unproctected (without reactor scram) loss of forced circulation (ULOF) transients using RELAP5-3D. The analyses examine both the inherent reactivity shutdown capability and decay heat removal performance by passive safety systems.

  4. Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

    SciTech Connect (OSTI)

    Cahalan, J.; Wigeland, R. (Argonne National Lab., IL (USA)); Friedel, G. (Internationale Atomreaktorbau GmbH (INTERATOM), Bergisch Gladbach (Germany, F.R.)); Kussmaul, G.; Royl, P. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.)); Moreau, J. (CEA Centre d'Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France)); Perks, M. (UKAEA Risley Nuclear Power Development Establishment (UK)

    1990-01-01T23:59:59.000Z

    In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs.

  5. Evaluation of an Integrated Gas-Cooled Reactor Simulator and Brayton Turbine-Generator

    SciTech Connect (OSTI)

    Hissam, D. Andy; Stewart, Eric [National Aeronautics and Space Administration, Marshall Space Flight Center, ER34, Huntsville, AL 35812 (United States)

    2006-07-01T23:59:59.000Z

    A closed-loop Brayton cycle, powered by a fission reactor, offers an attractive option for generating both planetary and in-space electric power. Non-nuclear testing of this type of system provides the opportunity to safely work out integration and system control challenges for a modest investment. Recognizing this potential, a team at Marshall Space Flight Center has evaluated the viability of integrating and testing an existing gas-cooled reactor simulator and a modified, commercially available, Brayton turbine-generator. Since these two systems were developed independently of one another, this evaluation sought to determine if they could be operated together at acceptable power levels, temperatures, and pressures. Thermal, fluid, and structural analyses show that this combined system can operate at acceptable power levels and temperatures. In addition, pressure drops across the reactor simulator, although higher than desired, are also viewed as acceptable. Three potential working fluids for the system were evaluated: N{sub 2}, He/Ar, and He/Xe. Other technical issues, such as electrical breakdown in the generator and the operation of the Brayton foil bearings using various gas mixtures, were also investigated. (authors)

  6. Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

    SciTech Connect (OSTI)

    Godfroy, Thomas J.; Bragg-Sitton, Shannon M. [NASA Marshall Space Flight Center, TD40, Huntsville, Alabama, 35812 (United States); University of Michgan, Dept. of Nuclear Engineering and Radiological Sciences, Ann Arbor MI 48109 (United States); Kapernick, Richard J. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2004-02-04T23:59:59.000Z

    One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

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

    E-Print Network [OSTI]

    Nitta, Christopher C

    2010-01-01T23:59:59.000Z

    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 ...

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

    E-Print Network [OSTI]

    McKee, Stephanie A

    2008-01-01T23:59:59.000Z

    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 ...

  9. 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 ...

  10. Optimization of actinide transmutation in innovative lead-cooled fast reactors

    E-Print Network [OSTI]

    Romano, Antonino, 1972-

    2003-01-01T23:59:59.000Z

    The thesis investigates the potential of fertile free fast lead-cooled modular reactors as efficient incinerators of plutonium and minor actinides (MAs) for application to dedicated fuel cycles for transmutation. A methodology ...

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

    E-Print Network [OSTI]

    Johnson, Brian C. (Brian Carl)

    2010-01-01T23:59:59.000Z

    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 ...

  12. ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT

    SciTech Connect (OSTI)

    M. G. McKellar; E. A. Harvego; A. M. Gandrik

    2010-11-01T23:59:59.000Z

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  13. High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

    SciTech Connect (OSTI)

    Pudelek, R. E.; Gilbert, W. C.

    2002-02-26T23:59:59.000Z

    This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste, except for the asbestos, was volume reduced via a private contract mechanism established by BJC. After volume reduction, the waste was packaged for rail shipment. This large waste management project successfully met cost and schedule goals.

  14. Hot gas path component cooling system

    DOE Patents [OSTI]

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

    2014-02-18T23:59:59.000Z

    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.

  15. HIGH TEMPERATURE GAS-COOLED REACTOR KNOWLEDGE MANAGEMENT

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

    be scrammed without regard to the onset of a core-conduction cool down without active cooling. However, even if all control and shutdown rods are scrammed, the operator must...

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

    SciTech Connect (OSTI)

    Not Available

    1980-02-01T23:59:59.000Z

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

  17. Utility/user requirements for the Modular High Temperature Gas-Cooled Reactor Plant

    SciTech Connect (OSTI)

    Swart, F.E.

    1987-06-01T23:59:59.000Z

    The purpose of this document is to set forth the top level Utilty/User requirements for a Modular High Temperature Gas-Cooled Reactor electric generating plant that incorporates 4 reactors and 2 turbine-generators to produce a nominal electrical output of 550 MW net.

  18. 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

    2013-01-01T23:59:59.000Z

    was simulated. The radiant cooling system was an exposedcooling + radiant cooling system alone may not be able toembedded surface radiant cooling systems. Table 3 summarizes

  19. Direct conversion nuclear reactor space power systems

    SciTech Connect (OSTI)

    Britt, E.J.; Fitzpatrick, G.O.

    1982-08-01T23:59:59.000Z

    This paper presents the results of a study of space nuclear reactor power systems using either thermoelectric or thermionic energy converters. An in-core reactor design and two heat pipe cooled out-of-core reactor designs were considered. One of the out-of-core cases utilized, long heat pipes (LHP) directly coupled to the energy converter. The second utilized a larger number of smaller heat pipes (mini-pipe) radiatively coupled to the energy converter. In all cases the entire system, including power conditioning, was constrained to be launched in a single shuttle flight. Assuming presently available performance, both the LHP thermoelectric system and minipipe thermionic system, designed to produce 100 kWe for seven years, would have a specific mass near 22kg/kWe. The specific mass of the thermionic minipipe system designed for a one year mission is 165 kg/kWe due to less fuel swelling. Shuttle imposed growth limits are near 300 kWe and 1.2 MWe for the thermoelectric and thermionic systems, respectively. Converter performance improvements could double this potential, and over 10 MWe may be possible for very short missions.

  20. Experimental and CFD Analysis of Advanced Convective Cooling Systems

    SciTech Connect (OSTI)

    Hassan, Yassin A; Ugaz, Victor M

    2012-06-27T23:59:59.000Z

    The objective of this project is to study the fundamental physical phenomena in the reactor cavity cooling system (RCCS) of very high-temperature reactors (VHTRs). One of the primary design objectives is to assure that RCCS acts as an ultimate heat sink capable of maintaining thermal integrity of the fuel, vessel, and equipment within the reactor cavity for the entire spectrum of postulated accident scenarios. Since construction of full-scale experimental test facilities to study these phenomena is impractical, it is logical to expect that computational fluid dynamics (CFD) simulations will play a key role in the RCCS design process. An important question then arises: To what extent are conventional CFD codes able to accurately capture the most important flow phenomena, and how can they be modified to improve their quantitative predictions? Researchers are working to tackle this problem in two ways. First, in the experimental phase, the research team plans to design and construct an innovative platform that will provide a standard test setting for validating CFD codes proposed for the RCCS design. This capability will significantly advance the state of knowledge in both liquid-cooled and gas-cooled (e.g., sodium fast reactor) reactor technology. This work will also extend flow measurements to micro-scale levels not obtainable in large-scale test facilities, thereby revealing previously undetectable phenomena that will complement the existing infrastructure. Second, in the computational phase of this work, numerical simulation of the flow and temperature profiles will be performed using advanced turbulence models to simulate the complex conditions of flows in critical zones of the cavity. These models will be validated and verified so that they can be implemented into commercially available CFD codes. Ultimately, the results of these validation studies can then be used to enable a more accurate design and safety evaluation of systems in actual nuclear power applications (both during normal operation and accident scenarios).

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

    SciTech Connect (OSTI)

    Ian Mckirdy

    2010-12-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Perko, Z. [Section Physics of Nuclear Reactors, Department of Radiation, Radionuclides and Reactors, TU Delft, Mekelweg 15, 2629 JB, Delft (Netherlands); Gilli, L.; Lathouwers, D.; Kloosterman, J. L. [Section Physics of Nuclear Reactors, Department of Radiation, Radionuclides and Reactors, Delft University of Technology, Mekelweg 15, 2629 JB, Delft (Netherlands)

    2013-07-01T23:59:59.000Z

    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)

  3. 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

  4. CONTROL SYSTEM FOR SOLAR HEATING and COOLING

    E-Print Network [OSTI]

    Dols, C.

    2010-01-01T23:59:59.000Z

    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 (

  5. Advanced Open-Cycle Desiccant Cooling System

    E-Print Network [OSTI]

    Ko, Y. J.; Charoensupaya, D.; Lavan, Z.

    1989-01-01T23:59:59.000Z

    The concept of staged regeneration as means of improving the desiccant cooling system performance is the subject of investigation in this study. In the staged regeneration, the regeneration section of desiccant dehumidifier is divided into two parts...

  6. CONTROL SYSTEM FOR SOLAR HEATING and COOLING

    E-Print Network [OSTI]

    Dols, C.

    2010-01-01T23:59:59.000Z

    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 (

  7. 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...

  8. Work Breakdown Structure and Plant/Equipment Designation System Numbering Scheme for the High Temperature Gas- Cooled Reactor (HTGR) Component Test Capability (CTC)

    SciTech Connect (OSTI)

    Jeffrey D Bryan

    2009-09-01T23:59:59.000Z

    This white paper investigates the potential integration of the CTC work breakdown structure numbering scheme with a plant/equipment numbering system (PNS), or alternatively referred to in industry as a reference designation system (RDS). Ideally, the goal of such integration would be a single, common referencing system for the life cycle of the CTC that supports all the various processes (e.g., information, execution, and control) that necessitate plant and equipment numbers be assigned. This white paper focuses on discovering the full scope of Idaho National Laboratory (INL) processes to which this goal might be applied as well as the factors likely to affect decisions about implementation. Later, a procedure for assigning these numbers will be developed using this white paper as a starting point and that reflects the resolved scope and outcome of associated decisions.

  9. 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

    2012-07-16T23:59:59.000Z

    by the forced convection of water is released in the atmosphere by an active secondary heat removal system. In case of accident the heat is removed by natural circulation of water and released in the atmosphere by evaporation. 4 Figure 2. Water... to the decay heat level, the core is expected to heat up due to the decrease in the heat removal. This is mainly caused by the loss of forced convection of the coolant and the system depressurization. In particular, during blowdown, the graphite dust...

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

    E-Print Network [OSTI]

    Pope, Michael A. (Michael Alexander)

    2006-01-01T23:59:59.000Z

    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 ...

  11. A helium-cooled blanket design of the low aspect ratio reactor

    SciTech Connect (OSTI)

    Wong, C.P.; Baxi, C.B.; Reis, E.E. [General Atomics, San Diego, CA (United States); Cerbone, R.; Cheng, E.T. [TSI Research, Solana Beach, CA (United States)

    1998-03-01T23:59:59.000Z

    An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh.

  12. Gas hydrate cool storage system

    DOE Patents [OSTI]

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

    1985-01-01T23:59:59.000Z

    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.

  13. 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)

    2013-01-01T23:59:59.000Z

    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 ...

  14. 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.

    2007-08-13T23:59:59.000Z

    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.

  15. Feasibility of Water Cooled Thorium Breeder Reactor Based on LWR Technology

    SciTech Connect (OSTI)

    Takaki, Naoyuki; Permana, Sidik; Sekimoto, Hiroshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

    2007-07-01T23:59:59.000Z

    The feasibility of Th-{sup 233}U fueled, homogenous breeder reactor based on matured conventional LWR technology was studied. The famous demonstration at Shipping-port showed that the Th-{sup 233}U fueled, heterogeneous PWR with four different lattice fuels was possible to breed fissile but its low averaged burn-up including blanket fuel and the complicated core configuration were not suitable for economically competitive reactor. The authors investigated the wide design range in terms of fuel cell design, power density, averaged discharge burn-up, etc. to determine the potential of water-cooled Th reactor as a competitive breeder. It is found that a low moderated (MFR=0.3) H{sub 2}O-cooled reactor with comparable burn-up with current LWR is feasible to breed fissile fuel but the core size is too large to be economical because of the low pellet power density. On the other hand, D{sub 2}O-cooled reactor shows relatively wider feasible design window, therefore it is possible to design a core having better neutronic and economic performance than H{sub 2}O-cooled. Both coolant-type cores show negative void reactivity coefficient while achieving breeding capability which is a distinguished characteristics of thorium based fuel breeder reactor. (authors)

  16. Reactor refueling containment system

    DOE Patents [OSTI]

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

    1995-05-02T23:59:59.000Z

    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.

  17. Reactor refueling containment system

    DOE Patents [OSTI]

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

    1995-01-01T23:59:59.000Z

    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.

  18. Role of small lead-cooled fast reactors for international deployment in worldwide sustainable nuclear energy supply.

    SciTech Connect (OSTI)

    Sienicki, J. J.; Wade, D. C.; Moisseytsev, A.; Nuclear Engineering Division

    2008-01-01T23:59:59.000Z

    Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. Meeting future worldwide projected energy demands during this century (e.g., 1000 to 2000 GWe by 2050) in a sustainable manner while maintaining CO2 emissions at or below today's level will require massive deployments of nuclear reactors in non-fuel cycle states as well as fuel cycle states. The projected energy demands of non-fuel cycle states will not be met solely through the deployment of Light Water Reactors (LWRs) in those states without using up the world's resources of fissile material (e.g., known plus speculative virgin uranium resources = 15 million tonnes). The present U.S. policy is focused upon domestic deployment of large-scale LWRs and sodium-cooled fast spectrum Advanced Burner Reactors (ABRs) working in a symbiotic relationship that burns existing fissile material while destroying the actinides which are generated. Other major nuclear nations are carrying out the development and deployment of SFR breeders as witness the planning for SFR breeder deployments in France, Japan, China, India, and Russia. Small (less that 300 MWe) and medium (300 to 700 MWe) size reactors are better suited to the growing economies and infrastructures of many non-fuel cycle states and developing nations. For those deployments, fast reactor converters which are fissile self-sufficient by creating as much fissile material as they consume are preferred to breeders that create more fissile material than they consume. Thus, there is a need for small and medium size fast reactors in non-fuel cycle states operating in a converter mode as well as large sodium-cooled fast breeders in fuel cycle states. Desired attributes for exportable small fast reactors include: proliferation resistance features such as restricted access to fuel; long core life further restricting access by reducing or eliminating the need for refueling; restricted potential to be misused in a breeding mode; fuel form that is unattractive in the safeguards sense; and a conversion ratio of unity to self-generate as much fissile material as is consumed. Desired attributes for exportable small reactor deployments in developing nations and remote sites also include: a small power level to match the smaller demand of towns or sites that are off-grid or on immature local grids; low enough cost to be economically competitive with alternative energy sources available to developing nation customers (e.g. diesel generators in remote locations); readily transported and assembled from transportable modules; simple to operate and highly reliable reducing plant operating staff requirements; as well as high reliability and passive safety reducing the number of accident initiators and need for safety systems as well as reducing the size of the exclusion and emergency planning zones. The Lead-Cooled Fast Reactor (LFR) has the desired attributes. An example of a small exportable LFR concept is the 20 MWe (45 MWt) Small Secure Transportable Autonomous Reactor (SSTAR) incorporating proliferation resistance, fissile selfsufficiency, autonomous load following, a high degree of passive safety, and supercritical carbon dioxide Brayton cycle energy conversion for high plant efficiency and improved economic competitiveness.

  19. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    SciTech Connect (OSTI)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01T23:59:59.000Z

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  20. Preoperational test report, recirculation condenser cooling systems

    SciTech Connect (OSTI)

    Clifton, F.T.

    1997-11-04T23:59:59.000Z

    This represents a preoperational test report for Recirculation Condenser Systems, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The four system provide condenser cooling water for vapor space cooling of tanks AY1O1, AY102, AZ1O1, AZ102. Each system consists of a valved piping loop, a pair of redundant recirculation pumps, a closed-loop evaporative cooling tower, and supporting instrumentation; equipment is located outside the farm on concrete slabs. Piping is routed to the each ventilation condenser inside the farm via below-grade concrete trenches. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System.

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

    DOE Patents [OSTI]

    Peterson, Per F.

    2013-05-14T23:59:59.000Z

    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.

  2. 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.

    1989-04-01T23:59:59.000Z

    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.

  3. 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)

    1984-01-01T23:59:59.000Z

    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.

  4. NASA's Marshall Space Flight Center Improves Cooling System Performanc...

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

    Improves Cooling System Performance Case study details Marshall Space Flight Center's innovative technologies to improve water efficiency and cooling performance for one of its...

  5. Validation of SCALE and the TRITON Depletion Sequence for Gas-Cooled Reactor Analysis

    SciTech Connect (OSTI)

    DeHart, Mark D [ORNL; Pritchard, Megan L [ORNL

    2008-01-01T23:59:59.000Z

    The very-high-temperature reactor (VHTR) is an advanced reactor concept that uses graphite-moderated fuel and helium gas as a coolant. At present there are two primary VHTR reactor designs under consideration for development: in the pebble-bed reactor, a core is loaded with 'pebbles' consisting of 6 cm diameter spheres, while in a high-temperature gas-cooled reactor, fuel rods are placed within prismatic graphite blocks. In both systems, fuel elements (spheres or rods) are comprised of tristructural-isotropic (TRISO) fuel particles. The TRISO particles are either dispersed in the matrix of a graphite pebble for the pebble-bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks for the prismatic concept. Two levels of heterogeneity exist in such fuel designs: (1) microspheres of TRISO particles dispersed in a graphite matrix of a cylindrical or spherical shape, and (2) neutron interactions at the rod-to-rod or sphere-to-sphere level. Such double heterogeneity (DH) provides a challenge to multigroup cross-section processing methods, which must treat each level of heterogeneity separately. A new capability to model doubly heterogeneous systems was added to the SCALE system in the release of Version 5.1. It was included in the control sequences CSAS and CSAS6, which use the Monte Carlo codes KENO V.a and KENO-VI, respectively, for three-dimensional neutron transport analyses and in the TRITON sequence, which uses the two-dimensional lattice physics code NEWT along with both versions of KENO for transport and depletion analyses. However, the SCALE 5.1 version of TRITON did not support the use of the DH approach for depletion. This deficiency has been addressed, and DH depletion will be available as an option in the upcoming release of SCALE 6. At present Oak Ridge National Laboratory (ORNL) staff are developing a set of calculations that may be used to validate SCALE for DH calculations. This paper discusses the results of calculations completed to date and the direction of future validation work.

  6. Reactor control rod timing system

    DOE Patents [OSTI]

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

    1982-01-01T23:59:59.000Z

    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.

    1982-02-09T23:59:59.000Z

    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. Passive cooling system for a vehicle

    DOE Patents [OSTI]

    Hendricks, Terry Joseph; Thoensen, Thomas

    2005-11-15T23:59:59.000Z

    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).

  9. Passive Cooling System for a Vehicle

    DOE Patents [OSTI]

    Hendricks, T. J.; Thoensen, T.

    2005-11-15T23:59:59.000Z

    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).

  10. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1994-12-31T23:59:59.000Z

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future.

  11. Process Cooling Pumping Systems Analysis

    E-Print Network [OSTI]

    Sherman, C.

    2008-01-01T23:59:59.000Z

    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...

  12. Interim status report on lead-cooled fast reactor (LFR) research and development.

    SciTech Connect (OSTI)

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G.; Li, N.; Hosemann, P.; Zhang, J.; Bolind, A.; LLNL; LANL; Univ. of Illinois

    2008-03-31T23:59:59.000Z

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigation of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup 15} (n/cm{sup 2}-s) and the initially 563 MWt PHENIX reactor attained 2.0 x 10{sup 15} (n/cm{sup 2}-s) before one of three intermediate cooling loops was shut down due to concerns about potential steam generator tube failures. The calculations do not assume a test assembly location for advanced fuels and materials irradiation in place of a fuel assembly (e.g., at the center of the core); the calculations have not examined whether it would be feasible to replace the central assembly by a test assembly location. However, having only fifteen driver assemblies implies a significant effect due to perturbations introduced by the test assembly. The peak neutron fast flux is low compared with the fast fluxes previously achieved in FFTF and PHENIX. Furthermore, the peak neutron fluence is only about half of the limiting value (4 x 10{sup 23} n/cm{sup 2}) typically used for ferritic steels. The results thus suggest that a larger power level (e.g., 400 MWt) and a larger core would be better for a TPP based upon the ELSY fuel assembly design and which can also perform irradiation testing of advanced fuels and materials. In particular, a core having a higher power level and larger dimensions would achieve a suitable average discharge burnup, peak fast flux, peak fluence, and would support the inclusion of one or more test assembly locations. Participation in the Generation IV International Forum Provisional System Steering Committee for the LFR is being maintained throughout FY 2008. Results from the analysis of samples previously exposed to flowing lead-bismuth eutectic (LBE) in the DELTA loop are summarized and a model for the oxidation/corrosion kinetics of steels in heavy liquid metal coolants was applied to systematically compare the calculated long-term (i.e., following several years of growth) oxide layer thicknesses of several steels.

  13. 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

  14. A 100 MWe advanced sodium-cooled fast reactor core concept

    SciTech Connect (OSTI)

    Kim, T. K.; Grandy, C.; Hill, R. N. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2012-07-01T23:59:59.000Z

    An Advanced sodium-cooled Fast Reactor core concept (AFR-100) was developed targeting a small electrical grid to be transportable to the plant site and operable for a long time without frequent refueling. The reactor power rating was strategically decided to be 100 MWe, and the core barrel diameter was limited to 3.0 m for transportability. The design parameters were determined by relaxing the peak fast fluence limit and bulk coolant outlet temperature to beyond irradiation experience assuming that advanced cladding and structural materials developed under US-DOE programs would be available when the AFR-100 is deployed. With a de-rated power density and U-Zr binary metallic fuel, the AFR-100 can maintain criticality for 30 years without refueling. The average discharge burnup of 101 MWd/kg is comparable to conventional design values, but the peak discharge fast fluence of {approx}6x10{sup 23} neutrons/cm{sup 2} is beyond the current irradiation experiences with HT-9 cladding. The evaluated reactivity coefficients provide sufficient negative feedbacks and the reactivity control systems provide sufficient shutdown margins. The integral reactivity parameters obtained from quasi-static reactivity balance analysis indicate that the AFR-100 meets the sufficient conditions for acceptable asymptotic core outlet temperature following postulated unprotected accidents. Additionally, the AFR-100 has sufficient thermal margins by grouping the fuel assemblies into eight orifice zones. (authors)

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

    SciTech Connect (OSTI)

    Lee Nelson

    2011-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1980-01-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Petroski, Robert C

    2008-01-01T23:59:59.000Z

    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 ...

  18. Desiccant Cooling Systems - A Review 

    E-Print Network [OSTI]

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

    1986-01-01T23:59:59.000Z

    or by solar systems. Comfort is achieved by reducing the moisture content of air by a solid or liquid desiccant and then reducing the temperature in an evaporative cooler (direct or indirect). Another system is one where the dehumidifier removes enough...

  19. Desiccant Cooling Systems - A Review

    E-Print Network [OSTI]

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

    1986-01-01T23:59:59.000Z

    or by solar systems. Comfort is achieved by reducing the moisture content of air by a solid or liquid desiccant and then reducing the temperature in an evaporative cooler (direct or indirect). Another system is one where the dehumidifier removes enough...

  20. Cooling load design tool for UFAD systems.

    E-Print Network [OSTI]

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

    2010-01-01T23:59:59.000Z

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

  1. Uncertainty Analysis for a De-pressurised Loss of Forced Cooling Event of the PBMR Reactor

    SciTech Connect (OSTI)

    Jansen van Rensburg, Pieter A.; Sage, Martin G. [PBMR, 1279 Mike Crawford Avenue, Centurion 0046 (South Africa)

    2006-07-01T23:59:59.000Z

    This paper presents an uncertainty analysis for a De-pressurised Loss of Forced Cooling (DLOFC) event that was performed with the systems CFD (Computational Fluid Dynamics) code Flownex for the PBMR reactor. An uncertainty analysis was performed to determine the variation in maximum fuel, core barrel and reactor pressure vessel (RPV) temperature due to variations in model input parameters. Some of the input parameters that were varied are: thermo-physical properties of helium and the various solid materials, decay heat, neutron and gamma heating, pebble bed pressure loss, pebble bed Nusselt number and pebble bed bypass flows. The Flownex model of the PBMR reactor is a 2-dimensional axisymmetrical model. It is simplified in terms of geometry and some other input values. However, it is believed that the model adequately indicates the effect of changes in certain input parameters on the fuel temperature and other components during a DLOFC event. Firstly, a sensitivity study was performed where input variables were varied individually according to predefined uncertainty ranges and the results were sorted according to the effect on maximum fuel temperature. In the sensitivity study, only seven variables had a significant effect on the maximum fuel temperature (greater that 5 deg. C). The most significant are power distribution profile, decay heat, reflector properties and effective pebble bed conductivity. Secondly, Monte Carlo analyses were performed in which twenty variables were varied simultaneously within predefined uncertainty ranges. For a one-tailed 95% confidence level, the conservatism that should be added to the best estimate calculation of the maximum fuel temperature for a DLOFC was determined as 53 deg. C. This value will probably increase after some model refinements in the future. Flownex was found to be a valuable tool for uncertainly analyses, facilitating both sensitivity studies and Monte Carlo analyses. (authors)

  2. ASSESSING POWER PLANT COOLING WATER INTAKE SYSTEM

    E-Print Network [OSTI]

    ASSESSING POWER PLANT COOLING WATER INTAKE SYSTEM ENTRAINMENT IMPACTS Prepared For: California be obvious that large studies like these require the coordinated work of many people. We would first like from the Duke Energy South Bay and Morro Bay power plants and the PG&E Diablo Canyon Power Plant

  3. 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

    2014-02-11T23:59:59.000Z

    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.

  4. Cooling system for a gas turbine

    DOE Patents [OSTI]

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

    2003-01-01T23:59:59.000Z

    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.

  5. 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)

    1991-01-01T23:59:59.000Z

    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).

  6. Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

    SciTech Connect (OSTI)

    Takeda, T.; Shimazu, Y.; Hibi, K.; Fujimura, K. [Research Inst. of Nuclear Engineering, Univ. of Fukui, 1cho-me 2gaiku 4, Kanawa-cho, Tsuruga-shi, Fukui 914-0055 (Japan)

    2012-07-01T23:59:59.000Z

    Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of this project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)

  7. Simulation of cooling systems in gas turbines

    SciTech Connect (OSTI)

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

    1996-04-01T23:59:59.000Z

    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.

  8. 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

  9. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    SciTech Connect (OSTI)

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23T23:59:59.000Z

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.

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

    DOE Patents [OSTI]

    Marra, John J

    2014-04-01T23:59:59.000Z

    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.

  11. CONCEPTUAL DESIGN OF A LUNAR REGOLITH CLUSTERED-REACTOR SYSTEM

    SciTech Connect (OSTI)

    John Darrell Bess

    2009-06-01T23:59:59.000Z

    It is proposed that a fast-fission, heatpipe-cooled, lunar-surface power reactor system be divided into subcritical units that could be launched safely without the incorporation of additional spectral shift absorbers or other complex means of control. The reactor subunits are to be emplaced directly into the lunar regolith utilizing the regolith not just for shielding but as the reflector material to increase the neutron economy of the system. While a single subunit cannot achieve criticality by itself, coordinated placement of additional subunits will provide a critical reactor system for lunar surface power generation. A lunar regolith clustered-reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of a slight increase in launch mass per rated power level and an overall reduction in neutron economy when compared to a single-reactor system. Additional subunits may be launched with future missions to increase the cluster size and power according to desired lunar base power demand and lifetime. The results address the potential uncertainties associated with the lunar regolith material and emplacement of the subunit systems. Physical distance between subunits within the clustered emplacement exhibits the most significant feedback regarding changes in overall system reactivity. Narrow, deep holes will be the most effective in reducing axial neutron leakage from the core. The variation in iron concentration in the lunar regolith can directly influence the overall system reactivity although its effects are less than the more dominant factors of subunit emplacement.

  12. 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...

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

    SciTech Connect (OSTI)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01T23:59:59.000Z

    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.

  14. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    SciTech Connect (OSTI)

    Barsell, A.W.

    1980-05-01T23:59:59.000Z

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  15. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    SciTech Connect (OSTI)

    Perret, G.; Pattupara, R. M. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Girardin, G. [Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland); Chawla, R. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland)

    2012-07-01T23:59:59.000Z

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

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

    Broader source: Energy.gov [DOE]

    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)

  17. Study plan for conducting a section 316(a) demonstration: K-Reactor cooling tower, Savannah River Site

    SciTech Connect (OSTI)

    Paller, M.H.

    1991-02-01T23:59:59.000Z

    The K Reactor at the Savannah River Site (SRS) began operation in 1954. The K-Reactor pumped secondary cooling water from the Savannah River and discharged directly to the Indian Grave Branch, a tributary of Pen Branch which flows to the Savannah River. During earlier operations, the temperature and discharge rates of cooling water from the K-reactor were up to approximately 70{degree}C and 400 cfs, substantially altering the thermal and flow regimes of this stream. These discharges resulted in adverse impacts to the receiving stream and wetlands along the receiving stream. As a component of a Consent Order (84-4-W as amended) with the South Carolina Department of Health and Environmental Control, the Department of Energy (DOE) evaluated the alternatives for cooling thermal effluents from K Reactor and concluded that a natural draft recirculating cooling tower should be constructed. The cooling tower will mitigate thermal and flow factors that resulted in the previous impacts to the Indian Grave/Pen Branch ecosystem. The purpose of the proposed biological monitoring program is to provide information that will support a Section 316(a) Demonstration for Indian Grave Branch and Pen Branch when K-Reactor is operated with the recirculating cooling tower. The data will be used to determine that Indian Grave Branch and Pen Branch support Balanced Indigenous Communities when K-Reactor is operated with a recirculating cooling tower. 4 refs., 1 fig. 1 tab.

  18. The Helium Cooling System and Cold Mass Support System for the MICE Coupling Solenoid

    E-Print Network [OSTI]

    2008-01-01T23:59:59.000Z

    18, No. 2, LBNL-63592 The Helium Cooling System and Coldmass support system and helium cooling system are discussedc and ? = 420 mm III. THE HELIUM COOLING SYSTEM The coupling

  19. APPLICATION OF DESIGN METHODOLOGY TO THE COOLING SYSTEM OF AN IN-LINE MACHINE VISION SYSTEM

    E-Print Network [OSTI]

    Shih, Albert J.

    APPLICATION OF DESIGN METHODOLOGY TO THE COOLING SYSTEM OF AN IN-LINE MACHINE VISION SYSTEM ....................................... 8 Figure 3. Cooling System Failure with Respect to Electrical Wiring................................. 8 Figure 4. Cooling System Failure with Respect to QD fitting

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

    SciTech Connect (OSTI)

    Weaver, Kevan Dean; Mac Donald, Philip Elsworth

    2002-10-01T23:59:59.000Z

    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.

  1. Shutdown system for a nuclear reactor

    DOE Patents [OSTI]

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

    1984-06-05T23:59:59.000Z

    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.

  2. 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)

    1984-01-01T23:59:59.000Z

    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.

  3. Cooling load design tool for UFAD systems.

    E-Print Network [OSTI]

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

    2010-01-01T23:59:59.000Z

    fraction (SPF) of cooling Supply Plenum SPF heat transfer bythrough the supply ple- Figure 2: Design day cooling loadsupply represent the????????????????????????????????????????????? air temperature, diffuser type and number, room setpoint instantaneous cooling

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

    SciTech Connect (OSTI)

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

    2010-02-01T23:59:59.000Z

    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. 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 geometry 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 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 bulk velocity) are approximately 4,300 and 12,400. Uncertainty analyses and a discussion of the standard problem are included. The measurements reveal developing, 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 is also presented.

  5. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    SciTech Connect (OSTI)

    Not Available

    1986-10-01T23:59:59.000Z

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

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

    SciTech Connect (OSTI)

    Wayne Moe

    2013-05-01T23:59:59.000Z

    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).

  7. A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration

    SciTech Connect (OSTI)

    Dubourg, M. [FRAMATOME, Paris-La Defense (France)

    1993-12-31T23:59:59.000Z

    This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable.

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

    E-Print Network [OSTI]

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

    2013-01-01T23:59:59.000Z

    Gain on Radiant Floor Cooling System Design. Proceedings ofWater-based radiant cooling systems are gaining popularityGain on Radiant Floor Cooling System Design. Proceedings of

  9. 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.

    2002-01-01T23:59:59.000Z

    the effect of heating and cooling system inefficiencies onwith inefficient heating and cooling systems in CaliforniaOperation of Residential Cooling Systems. Proceedings of the

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

    E-Print Network [OSTI]

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

    2014-01-01T23:59:59.000Z

    Olesen, Radiant floor cooling systems, ASHRAE Journal, 50 (radiant heating and cooling systems -- Part 2: Determinationradiant heating and cooling systems -- Part 4: Dimensioning

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

    E-Print Network [OSTI]

    Holtz, Michael J.

    2011-01-01T23:59:59.000Z

    and Hybrid Heating Cooling Systems Michael]. Holtz, WayneHYBRID HEATING AND COOLING SYSTEMS Michael J. Holtz Solarspace heating and cooling systems. It is based upon the mode

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

    E-Print Network [OSTI]

    Xu, TengFang T.

    2009-01-01T23:59:59.000Z

    Scalable Overhead Cooling Systems In Data Centers FinalFOR MODULAR, SCALABLE OVERHEAD COOLING SYSTEMS IN DATATHE CHARACTERISTICS OF MODULAR, SCALABLE COOLING SYSTEMS AND

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

    E-Print Network [OSTI]

    Adams, Barbara J

    2009-01-01T23:59:59.000Z

    Modular, Scalable Cooling Systems with Hot Aisle ContainmentMODULAR, SCALABLE COOLING SYSTEMS WITH HOT AISLE CONTAINMENTINFORMATION ON THE CHARACTERISTICS OF COOLING SYSTEMS AND

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

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01T23:59:59.000Z

    Scalable Liquid-Rack Cooling Systems in Data Centers FinalFOR A MODULAR, SCALABLE LIQUID-RACK COOLING SYSTEM IN DATAINFORMATION ON THE CHARACTERISTICS OF COOLING SYSTEMS,

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

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01T23:59:59.000Z

    Scalable Passive Cooling System in Data Centers Final ReportFOR A MODULAR, SCALABLE PASSIVE COOLING SYSTEM IN DATAINFORMATION ON THE CHARACTERISTICS OF COOLING SYSTEMS,

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

    E-Print Network [OSTI]

    Feng, Jingjuan; Schiavon, Stefano; Bauman, Fred

    2012-01-01T23:59:59.000Z

    Radiant Heating and Cooling Systems. Olesen, B. (2012). "surface heating and cooling systems: . Brussels, Europeanperformance in radiant cooling systems (Babiak, Olesen et

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

    E-Print Network [OSTI]

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

    2013-01-01T23:59:59.000Z

    B. 2008. “Radiant floor cooling systems. ” ASHRAE Journal 4.gain on radiant floor cooling system design. ” Proceedings,of designing radiant slab cooling systems, including load

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

    E-Print Network [OSTI]

    Bourassa, Norman; Haves, Philip; Huang, Joe

    2002-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01T23:59:59.000Z

    Liquid-Rack Cooling Systems in Data Centers Final Report ToCOOLING SYSTEM IN DATA CENTERS ..Liquid-Rack Cooling System in Data Centers 1 Problem

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

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01T23:59:59.000Z

    Scalable Passive Cooling System in Data Centers Final ReportCOOLING SYSTEM IN DATA CENTERS Scalable Passive Cooling System in Data Centers 1 Problem

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

    E-Print Network [OSTI]

    Xu, TengFang T.

    2009-01-01T23:59:59.000Z

    Overhead Cooling Systems In Data Centers Final Report To TheCOOLING SYSTEMS IN DATA CENTERS ..Scalable Overhead Cooling Systems in Data Centers 1 Problem

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

    E-Print Network [OSTI]

    Adams, Barbara J

    2009-01-01T23:59:59.000Z

    scalable cooling systems on overall data center energymodular cooling systems installed in a data center. Suchtraditional legacy data center cooling systems. 2 Technical

  3. 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

    2014-03-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Buongiorno, Jacopo, 1971-

    2001-01-01T23:59:59.000Z

    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 ...

  5. Sodium Cooled Fast Reactors and the Pyro-Process: Conversion of Nuclear Waste into a Fuel Source

    E-Print Network [OSTI]

    Belanger, David P.

    1 Sodium Cooled Fast Reactors and the Pyro-Process: Conversion of Nuclear Waste into a Fuel Source renewed interest amongst the nuclear science community as the debate over nuclear waste has increased .................................................................................27 2.1.2 Waste Minimization

  6. 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

    2013-01-01T23:59:59.000Z

    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 ...

  7. 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

    2010-01-01T23:59:59.000Z

    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 ...

  8. 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.)

    2014-01-01T23:59:59.000Z

    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 ...

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

    SciTech Connect (OSTI)

    Buden, D.

    1993-08-11T23:59:59.000Z

    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.

  11. The Helium Cooling System and Cold Mass Support System for the MICE Coupling Solenoid

    E-Print Network [OSTI]

    2008-01-01T23:59:59.000Z

    LBNL-63592 The Helium Cooling System and Cold Mass Supportsystem and helium cooling system are discussed in detail.420 mm III. THE HELIUM COOLING SYSTEM The coupling magnet is

  12. FULLY INTEGRATED ONE PHASE LIQUID COOLING SYSTEM FOR ORGANIC BOARDS

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    FULLY INTEGRATED ONE PHASE LIQUID COOLING SYSTEM FOR ORGANIC BOARDS D. May1 , B. Wunderle1 , F approach: A liquid cooling system has to be cus- tomised to its application. Further, it requires many dis in designing liquid cooling solu- tions are performance, reliability and price. To that end a one-phase liquid

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

    E-Print Network [OSTI]

    Moore, Eugene James Thomas

    2006-08-16T23:59:59.000Z

    abilities of system analysis codes, used to develop an understanding of light water reactor phenomenology, need to be proven for HTGRs. RELAP5-3D v2.3.6 is used to generate two reactor plant models for a code-to-code and a code-to-experiment benchmark...

  14. CCHP System with Interconnecting Cooling and Heating Network 

    E-Print Network [OSTI]

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

    2006-01-01T23:59:59.000Z

    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...

  15. CCHP System with Interconnecting Cooling and Heating Network

    E-Print Network [OSTI]

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

    2006-01-01T23:59:59.000Z

    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...

  16. Improving the Water Efficiency of Cooling Production System

    E-Print Network [OSTI]

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

    2010-01-01T23:59:59.000Z

    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...

  17. The Thermodynamic and Cost Benefits of Floating Cooling Systems 

    E-Print Network [OSTI]

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

    1983-01-01T23:59:59.000Z

    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...

  18. Cooling load differences between radiant and air systems

    E-Print Network [OSTI]

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

    2013-01-01T23:59:59.000Z

    radiant heat transfer for cooling load calculation.heat gain is well recognized by cooling load calculationload calculation approach for radiant systems, Corgnati [17] also tackled the direct radiant heat

  19. The Thermodynamic and Cost Benefits of Floating Cooling Systems

    E-Print Network [OSTI]

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

    1983-01-01T23:59:59.000Z

    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...

  20. Integrated exhaust gas recirculation and charge cooling system

    DOE Patents [OSTI]

    Wu, Ko-Jen

    2013-12-10T23:59:59.000Z

    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.

  1. A review of existing gas-cooled reactor circulators with application of the lessons learned to the new production reactor circulators

    SciTech Connect (OSTI)

    White, L.S.

    1990-07-01T23:59:59.000Z

    This report presents the results of a study of the lessons learned during the design, testing, and operation of gas-cooled reactor coolant circulators. The intent of this study is to identify failure modes and problem areas of the existing circulators so this information can be incorporated into the design of the circulators for the New Production Reactor (NPR)-Modular High-Temperature Gas Cooled Reactor (MHTGR). The information for this study was obtained primarily from open literature and includes data on high-pressure, high-temperature helium test loop circulators as well as the existing gas cooled reactors worldwide. This investigation indicates that trouble free circulator performance can only be expected when the design program includes a comprehensive prototypical test program, with the results of this test program factored into the final circulator design. 43 refs., 7 tabs.

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

    DOE Patents [OSTI]

    Jardine, D.M.

    1983-03-22T23:59:59.000Z

    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. Economizer refrigeration cycle space heating and cooling system and process

    DOE Patents [OSTI]

    Jardine, Douglas M. (Colorado Springs, CO)

    1983-01-01T23:59:59.000Z

    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.

  4. Polk power station syngas cooling system

    SciTech Connect (OSTI)

    Jenkins, S.D.

    1995-01-01T23:59:59.000Z

    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.

  5. Corrosion in HVDC valve cooling systems

    SciTech Connect (OSTI)

    Jackson, P.O.; Abrahamsson, B.; Gustavsson, D.; Igetoft, L.

    1997-04-01T23:59:59.000Z

    Stainless steel couplings in the main cooling water pipes of HVDC thyristor valves have been in use since 1983, with an overall satisfactory behavior. However, some water leakage due to corrosion below the sealing O-rings of the couplings was observed during 1992. An extensive investigation and follow-up worldwide showed a direct correlation between water quality and the corrosion rate of the stainless steel couplings. Recommendations are given about actions to be taken in order to maintain a long lifetime for the fine water systems.

  6. A method for measurement of delayed neutron parameters for liquid-metal-cooled power reactors

    SciTech Connect (OSTI)

    Vilim, R.B. [Argonne National Lab., IL (United States); Brock, R.W. [Babcock and Wilcox, Lynchburg, VA (United States)

    1996-06-01T23:59:59.000Z

    The trend toward increased reliance on passive features for power reactor safety makes it important to obtain the characteristics of the reactor system from measurements on the system. A method is described for solving for the delayed neutron parameters in a liquid-metal power reactor by fitting an analytic solution of the point-kinetics equations to the flux die-away from a dropped rod in an initially critical core. The method includes treatment of those conditions found in a power reactor that depart from those in a critical assembly experiment. These include a comparatively long rod drop time and a detector signal that instead of providing an integrated count rate is a sampled data signal proportional to the instantaneous fission power. The delayed neutron parameter values calculated from a rod drop experiment in the Experimental Breeder Reactor II are in agreement with values calculated using first principles and knowledge of core material composition and nuclear cross sections.

  7. 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

    2003-09-01T23:59:59.000Z

    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.

  8. 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

    2011-01-01T23:59:59.000Z

    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.

  9. Application of Gamma code coupled with turbomachinery models for high temperature gas-cooled reactors

    SciTech Connect (OSTI)

    Chang Oh

    2008-02-01T23:59:59.000Z

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

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

    E-Print Network [OSTI]

    Holtz, Michael J.

    2011-01-01T23:59:59.000Z

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

  11. TREAT (Transient Reactor Test Facility) reactor control rod scram system simulations and testing

    SciTech Connect (OSTI)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01T23:59:59.000Z

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs.

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    type fast reactor of the IV generation for regional powerELECTRA-FCC: a centre for Generation IV system research andunder the framework of generation-IV nuclear pro- grams or

  13. 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

    2008-01-01T23:59:59.000Z

    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 ...

  14. 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

    2013-03-01T23:59:59.000Z

    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.

  15. 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))

    1992-12-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    McDonald, Kirk

    and a remote PVSS supervision station connected via Ethernet. Operation Modes: Cooling of proximity cryogenicsCooling System for the MERIT High-Power Target Experiment Haug F., Pereira H., Silva P., Pezzetti M a free mercury jet inside a normal conducting pulsed 15 T capture solenoid magnet cooled with liquid

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

    SciTech Connect (OSTI)

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

    2006-04-01T23:59:59.000Z

    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.

  18. TREAT Upgrade Manual Reactor Control System and its interface with the Automatic Reactor Control System and the Plant Protection System

    SciTech Connect (OSTI)

    McDowell, W.P.

    1985-01-01T23:59:59.000Z

    The Transient Reactor Test Facility (TREAT) at Argonne National Laboratory is being upgraded to simulate extreme conditions in a reactor. This facility will be used to subject test assemblies of fuel bundles to very rapid and intense power transients. This paper will describe in detail the manual reactor control system and its interfaces with the plant protection system the automatic reactor control system.

  19. Cooling season study and economic analysis of a desiccant cooling system 

    E-Print Network [OSTI]

    Lee, James Howard

    1992-01-01T23:59:59.000Z

    model is ASME Journal of Solar Energy Engineering. Desiccant cooling offers significant advantages over vapor compression systems: I) Water and air are the only working fluids in desiccant systems so the environmental impact is minimal. 2) Energy...

  20. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program: Topical report I, selection of candidate alloys. Volume 3. Selection of surface coating/substrate systems for screening creep and structural stability studies

    SciTech Connect (OSTI)

    Not Available

    1980-06-20T23:59:59.000Z

    Considering the high temperature, low O/sub 2/, high C environment of operation in the Very High Temperature Reactor (VHTR) Systems, the utilization of coatings is envisaged to hold potential for extending component lifetimes through the formation of stable and continuous oxide films with enhanced resistance to C diffusion. A survey of the current state of technology for high temperature coatings has been performed. The usefulness of these coatings on the Mo, Ni, and Fe base alloys is discussed. Specifically, no coating substitute was identified for TZM other than the well known W-3 (pack silicide) and Al/sub 2/O/sub 3/ forming coatings were recommended for the Fe and Ni base structural materials. Recommendations as to coating types and processng have been made based on the predicted VHTR component size, shape, base metal and operational environment. Four tests designed to evaluate the effects of selected combinations of coatings and substrate matrices are recommended for consideration.

  1. 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.

    2007-03-08T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Monado, Fiber [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su'ud, Zaki; Waris, Abdul; Basar, Khairul [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-09-30T23:59:59.000Z

    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.

  3. Time-series investigation of anomalous thermocouple responses in a liquid-metal-cooled reactor

    SciTech Connect (OSTI)

    Gross, K.C.; Planchon, H.P.; Poloncsik, J.

    1988-03-24T23:59:59.000Z

    A study was undertaken using SAS software to investigate the origin of anomalous temperature measurements recorded by thermocouples (TCs) in an instrumented fuel assembly in a liquid-metal-cooled nuclear reactor. SAS macros that implement univariate and bivariate spectral decomposition techniques were employed to analyze data recorded during a series of experiments conducted at full reactor power. For each experiment, data from physical sensors in the tests assembly were digitized at a sampling rate of 2/s and recorded on magnetic tapes for subsequent interactive processing with CMS SAS. Results from spectral and cross-correlation analyses led to the identification of a flow rate-dependent electromotive force (EMF) phenomenon as the origin of the anomalous TC readings. Knowledge of the physical mechanism responsible for the discrepant TC signals enabled us to device and justify a simple correction factor to be applied to future readings.

  4. Method for fabricating wrought components for high-temperature gas-cooled reactors and product

    DOE Patents [OSTI]

    Thompson, Larry D. (San Diego, CA); Johnson, Jr., William R. (San Diego, CA)

    1985-01-01T23:59:59.000Z

    A method and alloys for fabricating wrought components of a high-temperature gas-cooled reactor are disclosed. These wrought, nickel-based alloys, which exhibit strength and excellent resistance to carburization at elevated temperatures, include aluminum and titanium in amounts and ratios to promote the growth of carburization resistant films while preserving the wrought character of the alloys. These alloys also include substantial amounts of molybdenum and/or tungsten as solid-solution strengtheners. Chromium may be included in concentrations less than 10% to assist in fabrication. Minor amounts of carbon and one or more carbide-forming metals also contribute to high-temperature strength.

  5. The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

    SciTech Connect (OSTI)

    Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W.; Verfondern, K.

    1988-01-01T23:59:59.000Z

    High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, licensing, and risk analyses. During off normal events, where temperatures may reach up to 1600/degree/C, the data show that no significant radionuclide releases from the fuel will occur.

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

    SciTech Connect (OSTI)

    Larry Demick

    2010-08-01T23:59:59.000Z

    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.

  7. Progress of the stochastic cooling system of the Collector Ring

    E-Print Network [OSTI]

    Dimopoulou, C; Bohm, R; Dolinskyy, O; Franzke, B; Hettrich, R; Maier, W; Menges, R; Nolden, F; Peschke, C; Petri, P; Steck, M; Thorndahl, L

    2013-01-01T23:59:59.000Z

    An overview of the recent achievements and ongoing developments for the stochastic cooling system of the Collector Ring is given. In focus are the hardware developments as well as the progress in predicting the system performance. The system operates in the frequency band 1-2 GHz, it has to provide fast 3D cooling of antiproton, rare isotope and stable heavy ion beams. The main challenges are (i) the cooling of antiprotons by means of cryogenic movable pick-up electrodes and (ii) the fast two-stage cooling (pre-cooling by the Palmer method, followed by the notch filter method) of the hot rare isotope beams (RIBs). Recently, a novel code for simulating the cooling process in the time domain has been developed at CERN. First results for the momentum cooling for heavy ions in the CR will be shown in comparison with results obtained in the frequency domain with the Fokker-Planck approach.

  8. TREAT Reactor Control and Protection System

    SciTech Connect (OSTI)

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

    1985-01-01T23:59:59.000Z

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.

  9. Self-actuating reactor shutdown system

    DOE Patents [OSTI]

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

    1988-01-01T23:59:59.000Z

    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.

  10. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    SciTech Connect (OSTI)

    C. Fiorina; N. E. Stauff; F. Franceschini; M. T. Wenner; A. Stanculescu; T. K. Kim; A. Cammi; M. E. Ricotti; R. N. Hill; T. A. Taiwo; M. Salvatores

    2013-12-01T23:59:59.000Z

    The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.

  11. HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

    SciTech Connect (OSTI)

    Gorensek, M.

    2011-07-06T23:59:59.000Z

    Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

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

    SciTech Connect (OSTI)

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28T23:59:59.000Z

    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.

  13. 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.

    2011-06-01T23:59:59.000Z

    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.

  14. auxiliary cooling system: Topics by E-print Network

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

    Heating and Power (CHP) systems. The concept (more) Liao, Xiaohong 2004-01-01 54 ENERGY SAVINGS AND THERMAL COMFORT OF SEPARATE SENSIBLE AND LATENT COOLING...

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

    Open Energy Info (EERE)

    Jump to: navigation, search Last modified on July 22, 2011. Project Title Cedarville School District Retrofit of Heating and Cooling Systems with Geothermal Heat Pumps and Ground...

  16. Distributed expert systems for nuclear reactor control

    SciTech Connect (OSTI)

    Otaduy, P.J.

    1992-12-01T23:59:59.000Z

    A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed.

  17. Distributed expert systems for nuclear reactor control

    SciTech Connect (OSTI)

    Otaduy, P.J.

    1992-01-01T23:59:59.000Z

    A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed.

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

    E-Print Network [OSTI]

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

    2013-01-01T23:59:59.000Z

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

  19. 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

    2013-12-01T23:59:59.000Z

    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.

  20. 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)

    2012-06-19T23:59:59.000Z

    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.

  1. Preliminary Study of Turbulent Flow in the Lower Plenum of a Gas-Cooled Reactor

    SciTech Connect (OSTI)

    T. Gallaway; D.P. Guillen; H.M. McIlroy, Jr.; S.P. Antal

    2007-09-01T23:59:59.000Z

    A preliminary study of the turbulent flow in a scaled model of a portion of the lower plenum of a gas-cooled advanced reactor concept has been conducted. The reactor is configured such that hot gases at various temperatures exit the coolant channels in the reactor core, where they empty into a lower plenum and mix together with a crossflow past vertical cylindrical support columns, then exit through an outlet duct. An accurate assessment of the flow behavior will be necessary prior to final design to ensure that material structural limits are not exceeded. In this work, an idealized model was created to mimic a region of the lower plenum for a simplified set of conditions that enabled the flow to be treated as an isothermal, incompressible fluid with constant properties. This is a first step towards assessing complex thermal fluid phenomena in advanced reactor designs. Once such flows can be computed with confidence, heated flows will be examined. Experimental data was obtained using three-dimensional Particle Image Velocimetry (PIV) to obtain non-intrusive flow measurements for an unheated geometry. Computational fluid dynamic (CFD) predictions of the flow were made using a commercial CFD code and compared to the experimental data. The work presented here is intended to be scoping in nature, since the purpose of this work is to identify improvements that can be made to subsequent computations and experiments. Rigorous validation of computational predictions will eventually be necessary for design and analysis of new reactor concepts, as well as for safety analysis and licensing calculations.

  2. Progress on the MICE Liquid Absorber Cooling and Cryogenic Distribution System

    E-Print Network [OSTI]

    2005-01-01T23:59:59.000Z

    PROGRESS ON THE MICE LIQUID ABSORBER COOLING AND CRYOGENIC DISTRIBUTION SYSTEMprogress made on the design of the cryogenic cooling system

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

    SciTech Connect (OSTI)

    Sonat Sen; Gilles Youinou

    2013-02-01T23:59:59.000Z

    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

  4. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    SciTech Connect (OSTI)

    Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

    1981-01-01T23:59:59.000Z

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

  5. Desiccant-based, heat actuated cooling assessment for DHC systems

    SciTech Connect (OSTI)

    DiBella, F.; Patch, K.; Becker, F.

    1989-10-01T23:59:59.000Z

    The goal of the project is to perform a conceptual design, systems analysis and case study evaluation of an application of a desiccant-based, heat actuated cooling system in a District Heating System. The results of this study will encourage the deployment of cooler transport temperatures in District Heating Systems. The proposed concept includes a liquid or solid desiccant-based air cooling and drying system that can be integrated with an existing HVAC system. 3 refs., 6 figs.

  6. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, April 1, 1983-September 30, 1983

    SciTech Connect (OSTI)

    Not Available

    1983-12-01T23:59:59.000Z

    An assessment of the HTGR opportunities from the year 2000 through 2045 was the principal activity on the Market Definition Task (WBS 03). Within the Plant Technology (WBS 13) task, there were activities to develop analytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. The activities in support of the HTGR-SC/C Lead Plant (WBS 30 and 31) were the participation in the Lead Plant System Engineering (LPSE) effort and the plant simulation task. The efforts on the Advanced HTGR systems was performed under the Modular Reactor Systems (MRS) (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors.

  7. Scanning tunneling microscope assembly, reactor, and system

    DOE Patents [OSTI]

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

    2014-11-18T23:59:59.000Z

    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.

  8. Debris trap in a turbine cooling system

    DOE Patents [OSTI]

    Wilson, Ian David (Clifton Park, NY)

    2002-01-01T23:59:59.000Z

    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.

  9. Westinghouse Small Modular Reactor nuclear steam supply system design

    SciTech Connect (OSTI)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01T23:59:59.000Z

    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 first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)

  10. REPRESENTATIVE SOURCE TERMS AND THE INFLUENCE OF REACTOR ATTRIBUTES ON FUNCTIONAL CONTAINMENT IN MODULAR HIGH-TEMPERATURE GAS-COOLED REACTORS

    SciTech Connect (OSTI)

    Petti, D. A.; Hobbins, R. R.; Lowry, Peter P.; Gougar, Hans

    2013-11-01T23:59:59.000Z

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

  11. Representative Source Terms and the Influence of Reactor Attributes on Functional Containment in Modular High-Temperature Gas-Cooled Reactors

    SciTech Connect (OSTI)

    D. A. Petti; Hans Gougar; Dick Hobbins; Pete Lowry

    2013-11-01T23:59:59.000Z

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

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

    E-Print Network [OSTI]

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

    2012-05-15T23:59:59.000Z

    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.

  13. SciTech Connect: Nuclear power reactor instrumentation systems...

    Office of Scientific and Technical Information (OSTI)

    of Publication: United States Language: English Subject: N79400* --Reactors--Reactor Control Systems; N46110 -- Instrumentation--Radiation Detection Instruments--General...

  14. Optimized Design of a Furnace Cooling System

    E-Print Network [OSTI]

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

    2013-01-01T23:59:59.000Z

    at higher temperatures. The second mechanism considers the introduction of forced argon convection. Argon is used in the process to mitigate part oxidation. Cycling argon through the furnace during cooling increases convection over the parts and removes heat...

  15. Optimized Design of a Furnace Cooling System 

    E-Print Network [OSTI]

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

    2013-01-01T23:59:59.000Z

    at higher temperatures. The second mechanism considers the introduction of forced argon convection. Argon is used in the process to mitigate part oxidation. Cycling argon through the furnace during cooling increases convection over the parts and removes heat...

  16. BSU GHP District Heating and Cooling System (Phase I)

    Broader source: Energy.gov [DOE]

    Project objectives: Create a campus geothermal heating and cooling system; Validate the cost savings associated with a geothermal system; Reduce emissions of CO2, CO, PM, SO2, NOx.

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

    E-Print Network [OSTI]

    Fiedel, Ethan R

    2011-01-01T23:59:59.000Z

    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 ...

  18. Special Property Assessment for Renewable Heating and Cooling Systems

    Broader source: Energy.gov [DOE]

    Title 8 of Maryland’s property tax code includes a state-wide special assessment for solar and geothermal heating and cooling systems. Under this provision, such systems are to be assessed at not...

  19. Heat pump system with selective space cooling

    DOE Patents [OSTI]

    Pendergrass, J.C.

    1997-05-13T23:59:59.000Z

    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.

  20. Heat pump system with selective space cooling

    DOE Patents [OSTI]

    Pendergrass, Joseph C. (Gainesville, GA)

    1997-01-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    OF LARGE FAST REACTORS Calculation examples A typicalMonte Carlo Reactor Physics Burnup Calculation Code. Tech.reactor core design from experience and coarse calculations

  2. Study on neutronic of very small Pb - Bi cooled no-onsite refueling nuclear power reactor (VSPINNOR)

    SciTech Connect (OSTI)

    Arianto, Fajar, E-mail: ariantofajar@gmail.com [Laboratory of Nuclear and Biophysics, Department of Physics, Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40132, Indonesia and Laboratory of Atom and Nuclear, Department of Physics, Diponegoro University, Jl. Prof. Soedarto, S.H., Tembala (Indonesia); Su'ud, Zaki, E-mail: szaki@fi.itba.c.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) (Indonesia); Zuhair [Center for Reactor Technology and Nuclear Safety, National Nuclear Energy Agency, Kawasan Puspiptek, Gedung No. 80, Serpong, Tangerang 15310 (Indonesia)

    2014-09-30T23:59:59.000Z

    A conceptual design study on Very Small Pb-Bi No-Onsite Refueling Cooled Nuclear Reactor (VSPINNOR) with Uranium nitride fuel using MCNPX program has been performed. In this design the reactor core is divided into three regions with different enrichment. At the center of the core is laid fuel without enrichment (internal blanket). While for the outer region using fuel enrichment variations. VSPINNOR fast reactor was operated for 10 years without refueling. Neutronic analysis shows optimized result of VSPINNOR has a core of 50 cm radius and 100 cm height with 300 MWth thermal power output at 60% fuel fraction that can be operated 18 years without refueling or fuel shuffling.

  3. NEUTRON ACTIVATION COOL-DOWN OF THE TOKAMAK FUSION TEST REACTOR

    E-Print Network [OSTI]

    involved the safe handling and processing about 100g of tritium. This resulted in manageable long concrete Test Cell showing the relative locations of the vessel, neutral beam injection systems, the vacuum. INTRODUCTION The Tokamak Fusion Test Reactor (TFTR) began high power deuterium-tritium (D-T) fueled operations

  4. 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, E-mail: fiber.monado@gmail.com [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Aziz, Ferhat [National Nuclear Energy Agency of Indonesia (BATAN) (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-02-12T23:59:59.000Z

    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.

  5. Control system for a small fission reactor

    DOE Patents [OSTI]

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

    1986-01-01T23:59:59.000Z

    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. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

  6. 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)

    2011-03-01T23:59:59.000Z

    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.

  7. 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)

    2010-02-23T23:59:59.000Z

    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.

  8. Technical basis for extending storage of the UK's advanced gas-cooled reactor fuel

    SciTech Connect (OSTI)

    Hambley, D.I. [National Nuclear Laboratory, Sellafield, Seascale, Cumbria, CA20 1PG (United Kingdom)

    2013-07-01T23:59:59.000Z

    The UK Nuclear Decommissioning Agency has recently declared a date for cessation of reprocessing of oxide fuel from the UK's Advanced Gas-cooled Reactors (AGRs). This will fundamentally change the management of AGR fuel: from short term storage followed by reprocessing to long term fuel storage followed, in all likelihood, by geological disposal. In terms of infrastructure, the UK has an existing, modern wet storage asset that can be adapted for centralised long term storage of dismantled AGR fuel under the required pond water chemistry. No AGR dry stores exist, although small quantities of fuel have been stored dry as part of experimental programmes in the past. These experimental programmes have shown concerns about corrosion rates.

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

    SciTech Connect (OSTI)

    Larry Demick

    2011-08-01T23:59:59.000Z

    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.

  10. 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.

    2013-09-03T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Phillip Mills

    2012-02-01T23:59:59.000Z

    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. Seismicity and seismic response of the Soviet-designed VVER (Water-cooled, Water moderated Energy Reactor) reactor plants

    SciTech Connect (OSTI)

    Ma, D.C.; Gvildys, J.; Wang, C.Y.; Spencer, B.W.; Sienicki, J.J.; Seidensticker, R.W.; Purvis, E.E. III

    1989-01-01T23:59:59.000Z

    On March 4, 1977, a strong earthquake occurred at Vrancea, Romania, about 350 km from the Kozloduy plant in Bulgaria. Subsequent to this event, construction of the unit 2 of the Armenia plant was delayed over two years while seismic features were added. On December 7, 1988, another strong earthquake struck northwest Armenia about 90 km north of the Armenia plant. Extensive damage of residential and industrial facilities occurred in the vicinity of the epicenter. The earthquake did not damage the Armenia plant. Following this event, the Soviet government announced that the plant would be shutdown permanently by March 18, 1989, and the station converted to a fossil-fired plant. This paper presents the results of the seismic analyses of the Soviet-designed VVER (Water-cooled, Water moderated Energy Reactor) plants. Also presented is the information concerning seismicity in the regions where VVERs are located and information on seismic design of VVERs. The reference units are the VVER-440 model V230 (similar to the two units of the Armenia plant) and the VVER-1000 model V320 units at Kozloduy in Bulgaria. This document provides an initial basis for understanding the seismicity and seismic response of VVERs under seismic events. 1 ref., 9 figs., 3 tabs.

  13. Closed loop air cooling system for combustion turbines

    DOE Patents [OSTI]

    Huber, D.J.; Briesch, M.S.

    1998-07-21T23:59:59.000Z

    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. 1 fig.

  14. Closed loop air cooling system for combustion turbines

    DOE Patents [OSTI]

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

    1998-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    None

    2010-10-01T23:59:59.000Z

    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.

  16. Control system for a small fission reactor

    DOE Patents [OSTI]

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

    1985-02-08T23:59:59.000Z

    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.

  17. The Windscale Advanced Gas Cooled Reactor (WAGR) Decommissioning Project A Close Out Report for WAGR Decommissioning Campaigns 1 to 10 - 12474

    SciTech Connect (OSTI)

    Halliwell, Chris [Sellafield Ltd, Sellafield (United Kingdom)

    2012-07-01T23:59:59.000Z

    The reactor core of the Windscale Advanced Gas-Cooled Reactor (WAGR) has been dismantled as part of an ongoing decommissioning project. The WAGR operated until 1981 as a development reactor for the British Commercial Advanced Gas cooled Reactor (CAGR) power programme. Decommissioning began in 1982 with the removal of fuel from the reactor core which was completed in 1983. Subsequently, a significant amount of engineering work was carried out, including removal of equipment external to the reactor and initial manual dismantling operations at the top of the reactor, in preparation for the removal of the reactor core itself. Modification of the facility structure and construction of the waste packaging plant served to provide a waste route for the reactor components. The reactor core was dismantled on a 'top-down' basis in a series of 'campaigns' related to discrete reactor components. This report describes the facility, the modifications undertaken to facilitate its decommissioning and the strategies employed to recognise the successful decommissioning of the reactor. Early decommissioning tasks at the top of the reactor were undertaken manually but the main of the decommissioning tasks were carried remotely, with deployment systems comprising of little more than crane like devices, intelligently interfaced into the existing structure. The tooling deployed from the 3 tonne capacity (3te) hoist consisted either purely mechanical devices or those being electrically controlled from a 'push-button' panel positioned at the operator control stations, there was no degree of autonomy in the 3te hoist or any of the tools deployed from it. Whilst the ATC was able to provide some tele-robotic capabilities these were very limited and required a good degree of driver input which due to the operating philosophy at WAGR was not utilised. The WAGR box proved a successful waste package, adaptable through the use of waste box furniture specific to the waste-forms generated throughout the various decommissioning campaigns. The use of low force compaction for insulation and soft wastes provided a simple, robust and cost effective solution as did the direct encapsulation of LLW steel components in the later stages of reactor decommissioning. Progress through early campaigns was good, often bettering the baseline schedule, especially when undertaking the repetitive tasks seen during Neutron Shield and Graphite Core decommissioning, once the operators had become experienced with the equipment, though delays became more pronounced, mainly as a result of increased failures due to the age and maintainability of the RDM and associated equipment. Extensive delays came about as a result of the unsupported insulation falling away from the pressure vessel during removal and the inability of the ventilation system to manage the sub micron particulate generated during IPOPI cutting operations, though the in house development of revised and new methodologies ultimately led to the successful completion of PV and I removal. In a programme spanning over 12 years, the decommissioning of the reactor pressure vessel and core led to the production 110 ILW and 75 LLW WAGR boxes, with 20 LLW ISO freight containers of primary reactor wastes, resulting in an overall packaged volume of approximately 2500 cubic metres containing the estimated 460 cubic metres of the reactor structure. (authors)

  18. Protected air-cooled condenser for the Clinch River Breeder Reactor Plant

    SciTech Connect (OSTI)

    Louison, R.; Boardman, C.E.

    1981-05-29T23:59:59.000Z

    The long term residual heat removal for the Clinch River Breeder Reactor Plant (CRBRP) is accomplished through the use of three protected air-cooled condensers (PACC's) each rated at 15M/sub t/ following a normal or emergency shutdown of the reactor. Steam is condensed by forcing air over the finned and coiled condenser tubes located above the steam drums. The steam flow is by natural convection. It is drawn to the PACC tube bundle for the steam drum by the lower pressure region in the tube bundle created from the condensing action. The concept of the tube bundle employs a unique patented configuration which has been commercially available through CONSECO Inc. of Medfore, Wisconsin. The concept provides semi-parallel flow that minimizes subcooling and reduces steam/condensate flow instabilities that have been observed on other similar heat transfer equipment such as moisture separator reheaters (MSRS). The improved flow stability will reduce temperature cycling and associated mechanical fatigue. The PACC is being designed to operate during and following the design basis earthquake, depressurization from the design basis tornado and is housed in protective building enclosure which is also designed to withstand the above mentioned events.

  19. 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)

    2013-07-01T23:59:59.000Z

    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.

  20. Systems Issues in Nuclear Reactor Safety

    E-Print Network [OSTI]

    de Weck, Olivier L.

    Systems Issues in Nuclear Reactor Safety Commissioner George ApostolakisCommissioner George Apostolakis U.S. Nuclear Regulatory Commission CmrApostolakis@nrc.gov MIT SDM Conference on Systems Thinking, source, and special nuclear materials to ensure adequate protection of public health and safety, 3

  1. Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

    SciTech Connect (OSTI)

    Poncet, Bernard [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)] [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)

    2013-07-01T23:59:59.000Z

    About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

  2. 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.

    1992-02-01T23:59:59.000Z

    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.

  3. THE OPERATION OF A FORCED TWO PHASE COOLING SYSTEM ON A LARGE SUPERCONDUCTING MAGNET

    E-Print Network [OSTI]

    Green, M.A.

    2010-01-01T23:59:59.000Z

    A FORCED TWO PHASE COOLING SYSTEM ON A LARGE SUPERCONDUCTINGOPERATION O A FORCED T O PHASE COOLING SYSTEM F W O A LARGEof a forced two phase cooling system on a two meter diameter

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

    E-Print Network [OSTI]

    Holtz, Michael J.

    2011-01-01T23:59:59.000Z

    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

  5. Regulatory analysis for the resolution of Generic Issue 143: Availability of chilled water system and room cooling

    SciTech Connect (OSTI)

    Leung, V.T.

    1993-12-01T23:59:59.000Z

    This report presents the regulatory analysis for Generic Issue (GI-143), {open_quotes}Availability of Chilled Water System and Room Cooling.{close_quotes} The heating, ventilating, and air conditioning (HVAC) systems and related auxiliaries are required to provide control of environmental conditions in areas in light water reactor (LWR) plants that contain safety-related equipment. In some plants, the HVAC and chilled water systems serve to maintain a suitable environment for both safety and non-safety-related areas. Although some plants have an independent chilled water system for the safety-related areas, the heat removal capability often depends on the operability of other supporting systems such as the service water system or the component cooling water system. The operability of safety-related components depends upon operation of the HVAC and chilled water systems to remove heat from areas containing the equipment. If cooling to dissipate the heat generated is unavailable, the ability of the safety-related equipment to operate as intended cannot be assured. Typical components or areas in the nuclear power plant that could be affected by the failure of cooling from HVAC or chilled water systems include the (1) emergency switchgear and battery rooms, (2) emergency diesel generator room, (3) pump rooms for residual heat removal, reactor core isolation cooling, high-pressure core spray, and low-pressure core spray, and (4) control room. The unavailability of such safety-related equipment or areas could cause the core damage frequency (CDF) to increase significantly.

  6. Pressure loadings of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) reactor release mitigation structures from large-break LOCAs

    SciTech Connect (OSTI)

    Sienicki, J.J.; Horak, W.C. (Argonne National Lab., IL (USA); Brookhaven National Lab., Upton, NY (USA))

    1989-01-01T23:59:59.000Z

    Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs.

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

    SciTech Connect (OSTI)

    Ronald Farris; David Gertman; Jacques Hugo

    2014-03-01T23:59:59.000Z

    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.

  8. Desiccant-based, heat-actuated cooling assessment for DHC (District Heating and Cooling) systems

    SciTech Connect (OSTI)

    Patch, K.D.; DiBella, F.A.; Becker, F.E.

    1990-07-01T23:59:59.000Z

    An assessment has been completed of the use of desiccant-based, heat-actuated cooling for District Heating and Cooling (DHC) systems, showing that such desiccant-based cooling (DBC) systems are generally applicable to District Heating (DH) systems. Since the DH system only has to supply hot water (or steam) to its customers, systems that were designed as conventional two-pipe DH systems can now be operated as DHC systems without major additional capital expense. Desiccant-based DHC systems can be operated with low-grade DH-supplied heat, at temperatures below 180{degree}F, without significant loss in operating capacity, relative to absorption chillers. During this assessment, a systems analysis was performed, an experimental investigation was conducted, developmental requirements for commercializing DBC systems were examined, and two case studies were conducted. As a result of the case studies, it was found that the operating cost of a DBC system was competitive with or lower than the cost of purchasing DHC-supplied chilled water. However, because of the limited production volume and the current high capital costs of desiccant systems, the payback period is relatively long. In this regard, through the substitution of low-cost components specifically engineered for low-temperature DHC systems, the capital costs should be significantly reduced and overall economics made attractive to future users. 17 figs.

  9. 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

    2003-01-01T23:59:59.000Z

    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:

  10. adsorption cooling system: Topics by E-print Network

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

    one... Muller, M.R.; Muller, M.B. 2012-01-01 11 The Thermodynamic and Cost Benefits of Floating Cooling Systems Texas A&M University - TxSpace Summary: Historically, a fixed...

  11. air cooling system: Topics by E-print Network

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

    coil unit (FCU) and 100% fresh air are used... Mohamed, E.; Abdalla, K. N. 2010-01-01 8 Cold side thermal energy storage system for improved operation of air cooled power plants...

  12. air cooling systems: Topics by E-print Network

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

    coil unit (FCU) and 100% fresh air are used... Mohamed, E.; Abdalla, K. N. 2010-01-01 8 Cold side thermal energy storage system for improved operation of air cooled power plants...

  13. System and method for pre-cooling of buildings

    DOE Patents [OSTI]

    Springer, David A.; Rainer, Leo I.

    2011-08-09T23:59:59.000Z

    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.

  14. Double dark state cooling in a three level system

    E-Print Network [OSTI]

    Javier Cerrillo; Alex Retzker; Martin B. Plenio

    2011-12-29T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    with burnup of a depleted-uranium fueled sodium-cooled B&Bwith burnup of a depleted-uranium fueled sodium-cooled B&Bbalance integral of a depleted-uranium fueled sodium-cooled

  16. Cooling system for three hook ring segment

    DOE Patents [OSTI]

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

    2014-08-26T23:59:59.000Z

    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.

  17. Cavity cooling of an ensemble spin system

    E-Print Network [OSTI]

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

    2014-02-24T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Nelson, Karl Einar; Burks, Morgan T

    2014-04-22T23:59:59.000Z

    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.

  19. Hybrid Molten Salt Reactor (HMSR) System Study

    SciTech Connect (OSTI)

    Woolley, Robert D [PPPL; Miller, Laurence F [PPPL

    2014-04-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Peterson, Per

    2012-10-30T23:59:59.000Z

    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.

  1. Automated Test Coverage Measurement for Reactor Protection System Software

    E-Print Network [OSTI]

    Automated Test Coverage Measurement for Reactor Protection System Software Implemented in Function in implementing safety critical systems such as nuclear reactor protection systems. We have defined new structural- ing a case study using test cases prepared by domain experts for reactor protection system software

  2. INVESTIGATION OF FUNDAMENTAL THERMAL-HYDRAULIC PHENOMENA IN ADVANCED GAS-COOLED REACTORS

    SciTech Connect (OSTI)

    INVESTIGATION OF FUNDAMENTAL THERMAL-HYDRAULIC PHE

    2006-09-01T23:59:59.000Z

    INL LDRD funded research was conducted at MIT to experimentally characterize mixed convection heat transfer in gas-cooled fast reactor (GFR) core channels in collaboration with INL personnel. The GFR for Generation IV has generated considerable interest and is under development in the U.S., France, and Japan. One of the key candidates is a block-core configuration first proposed by MIT, has the potential to operate in Deteriorated Turbulent Heat Transfer (DTHT) regime or in the transition between the DTHT and normal forced or laminar convection regime during post-loss-of-coolant accident (LOCA) conditions. This is contrary to most industrial applications where operation is in a well-defined and well-known turbulent forced convection regime. As a result, important new need emerged to develop heat transfer correlations that make possible rigorous and accurate predictions of Decay Heat Removal (DHR) during post LOCA in these regimes. Extensive literature review on these regimes was performed and a number of the available correlations was collected in: (1) forced laminar, (2) forced turbulent, (3) mixed convection laminar, (4) buoyancy driven DTHT and (5) acceleration driven DTHT regimes. Preliminary analysis on the GFR DHR system was performed and using the literature review results and GFR conditions. It confirmed that the GFR block type core has a potential to operate in the DTHT regime. Further, a newly proposed approach proved that gas, liquid and super critical fluids all behave differently in single channel under DTHT regime conditions, thus making it questionable to extrapolate liquid or supercritical fluid data to gas flow heat transfer. Experimental data were collected with three different gases (nitrogen, helium and carbon dioxide) in various heat transfer regimes. Each gas unveiled different physical phenomena. All data basically covered the forced turbulent heat transfer regime, nitrogen data covered the acceleration driven DTHT and buoyancy driven DTHT, helium data covered the mixed convection laminar, acceleration driven DTHT and the laminar to turbulent transition regimes and carbon dioxide data covered the returbulizing buoyancy driven DTHT and non-returbulizing buoyancy induced DTHT. The validity of the data was established using the heat balance and the uncertainty analysis. Based on experimental data, the traditional threshold for the DTHT regime was updated to account for phenomena observed in the facility and a new heat transfer regime map was proposed. Overall, it can be stated that substantial reduction of heat transfer coefficient was observed in DTHT regime, which will have significant impact on the core and DHR design of passive GFR. The data were compared to the large number of existing correlations. None of the mixed convection laminar correlation agreed with the data. The forced turbulent and the DTHT regime, Celeta et al. correlation showed the best fit with the data. However, due to larger ratio of the MIT facility compared to the Celeta et al. facility and the returbuliziation due to the gas characteristics, the correlation sometimes under-predicts the heat transfer coefficient. Also, since Celeta et al. correlation requires the information of the wall temperature to evaluate the heat transfer coefficient, it is difficult to apply this correlation directly for predicting the wall temperature. Three new sets of correlation that cover all heat transfer regimes were developed. The bas

  3. Dynamic Impregnator Reactor System (Poster)

    SciTech Connect (OSTI)

    Not Available

    2012-09-01T23:59:59.000Z

    IBRF poster developed for the IBRF showcase. Describes the multifarious system designed for complex feedstock impregnation and processing. IBRF feedstock system has several unit operations combined into one robust system that provides for flexible and staged process configurations, such as spraying, soaking, low-severity pretreatment, enzymatic hydrolysis, fermentation, concentration/evaporation, and distillation.

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    to large conventional sodium fast reactors (SFR). TerraPowerincrease for a typical sodium fast reactor fuel rod geometryof the new Russian sodium fast reactor BN-800 [111]. The

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    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

  6. Reactor control rod timing system. [LMFBR

    DOE Patents [OSTI]

    Wu, P.T.K.

    1980-03-18T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01T23:59:59.000Z

    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)

  8. Fault-tolerant reactor protection system

    DOE Patents [OSTI]

    Gaubatz, D.C.

    1997-04-15T23:59:59.000Z

    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.

  9. Fault-tolerant reactor protection system

    DOE Patents [OSTI]

    Gaubatz, Donald C. (Cupertino, CA)

    1997-01-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    4 Reactivity feedback of large fast reactors 4.1temperature . . . . . . . . . . . . . . . . . . Fast reactorfission gas plenum212 Conventional fast reactor core design

  11. Vertical Pretreatment Reactor System (Poster)

    SciTech Connect (OSTI)

    Not Available

    2012-09-01T23:59:59.000Z

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

  12. Challenges in the Development of Advanced Reactors

    SciTech Connect (OSTI)

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

    2012-08-01T23:59:59.000Z

    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.

  13. System and method for cooling a combustion gas charge

    DOE Patents [OSTI]

    Massey, Mary Cecelia; Boberg, Thomas Earl

    2010-05-25T23:59:59.000Z

    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.

  14. 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.

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

    DOE Patents [OSTI]

    North, William Edward (Winter Springs, FL)

    2000-01-01T23:59:59.000Z

    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.

  16. 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)

    2002-01-01T23:59:59.000Z

    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.

  17. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01T23:59:59.000Z

    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)

  18. Gas cooled fast reactor control rod drive mechanism deceleration unit. Test program

    SciTech Connect (OSTI)

    Wagner, T.H.

    1981-10-01T23:59:59.000Z

    This report presents the results of the airtesting portion of the proof-of-principle testing of a Control Rod Scram Deceleration Device developed for use in the Gas Cooled Fast Reactor (GCFR). The device utilizes a grooved flywheel to decelerate the translating assembly (T/A). Two cam followers on the translating assembly travel in the flywheel grooves and transfer the energy of the T/A to the flywheel. The grooves in the flywheel are straight for most of the flywheel length. Near the bottom of the T/A stroke the grooves are spiraled in a decreasing slope helix so that the cam followers accelerate the flywheel as they transfer the energy of the falling T/A. To expedite proof-of-principle testing, some of the materials used in the fabrication of certain test article components were not prototypic. With these exceptions the concept appears to be acceptable. The initial test of 300 scrams was completed with only one failure and the failure was that of a non-prototypic cam follower outer sleeve material.

  19. 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

    2013-11-19T23:59:59.000Z

    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.

  20. Thermal Hydraulic Analyses for Coupling High Temperature Gas-Cooled Reactor to Hydrogen Plant

    SciTech Connect (OSTI)

    C.H. Oh; R. Barner; C. B. Davis; S. Sherman; P. Pickard

    2006-08-01T23:59:59.000Z

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. 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 cycle-efficiency 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. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The evaluations determined which configurations and coolants are the most promising from thermalhydraulic and efficiency points of view.

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    breeder reactors typically operate with an inner core of high fissile content surrounded by breeding blankets

  2. 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)

    2013-07-01T23:59:59.000Z

    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)

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

    E-Print Network [OSTI]

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

    2013-01-01T23:59:59.000Z

    heat transfer is handled in traditional cooling load calculationheat gain is well recognized by cooling load calculationload calculations for radiant systems should use the ASHRAE heat

  4. 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.

    2010-09-28T23:59:59.000Z

    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.

  5. Microchannel Reactor System for Catalytic Hydrogenation

    SciTech Connect (OSTI)

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

    2010-12-22T23:59:59.000Z

    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.

  6. Integrating externally developed systems for SNS Linac cooling and vacuum.

    SciTech Connect (OSTI)

    Marroquin, P. S. (Pilar S.)

    2001-01-01T23:59:59.000Z

    External contractors are developing the local cooling and vacuum control systems for the Spallation Neutron Source (SNS) linac. Soon these systems will be integrated into the facility-wide controls system. Allen-Bradley Logix5000 series programmable controllers, populated with appropriate input/output modules, were selected as the local controllers. These controllers will be interfaced to the facility-wide control system via VME systems with PowerPC processors running the Wind River VxWorks operating system and Experimental Physics and Industrial Control System (EPICS) front-end controller software. This paper describes the interface and integration issues driven by project, cooling system and vacuum system requirements and hardware selections.

  7. 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)

    2007-06-19T23:59:59.000Z

    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.

  8. Single Channel Testing for Characterization of the Direct Gas Cooled Reactor and the SAFE-100 Heat Exchanger

    SciTech Connect (OSTI)

    Bragg-Sitton, S.M. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Propulsion Research Center, NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States); Kapernick, R. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Godfroy, T.J. [Propulsion Research Center, NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States)

    2004-02-04T23:59:59.000Z

    Experiments have been designed to characterize the coolant gas flow in two space reactor concepts that are currently under investigation by NASA Marshall Space Flight Center and Los Alamos National Laboratory: the direct-drive gas-cooled reactor (DDG) and the SAFE-100 heatpipe-cooled reactor (HPR). For the DDG concept, initial tests have been completed to measure pressure drop versus flow rate for a prototypic core flow channel, with gas exiting to atmospheric pressure conditions. The experimental results of the completed DDG tests presented in this paper validate the predicted results to within a reasonable margin of error. These tests have resulted in a re-design of the flow annulus to reduce the pressure drop. Subsequent tests will be conducted with the re-designed flow channel and with the outlet pressure held at 150 psi (1 MPa). Design of a similar test for a nominal flow channel in the HPR heat exchanger (HPR-HX) has been completed and hardware is currently being assembled for testing this channel at 150 psi. When completed, these test programs will provide the data necessary to validate calculated flow performance for these reactor concepts (pressure drop and film temperature rise)

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

    E-Print Network [OSTI]

    Adams, Barbara J

    2009-01-01T23:59:59.000Z

    The supply water temperature to the cooling modules used ininlet supply air temperatures, as was the cooling module’sCooling System 2 be evaluated when operating with higher supply

  10. 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

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

    SciTech Connect (OSTI)

    Su'ud, Zaki, E-mail: szaki@fi.itba.c.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) (Indonesia); Sekimoto, H., E-mail: hsekimot@gmail.com [Research Lab. For Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo (Japan)

    2014-09-30T23:59:59.000Z

    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.

  12. SILER: Seismic-Initiated events risk mitigation in Lead-cooled Reactors

    SciTech Connect (OSTI)

    Forni, M. [ENEA, Via Martin di Monte Sole 4, 40129 Bologna (Italy); De Grandis, S. [SINTEC, Via Santo Stefano 20, 40125 Bologna (Italy)

    2012-07-01T23:59:59.000Z

    SILER is a Collaborative Project, partially funded by the European Commission, aimed at studying the risk associated to seismic initiated events in Generation IV Heavy Liquid Metal reactors and developing adequate protection measures. The attention is focused on the evaluation of the effects of earthquakes (with particular regards to beyond design seismic events) and to the identification of mitigation strategies, acting both on structures and components design (as well as on the development of seismic isolation devices) which can also have positive effects on economics, leading to an high level of plant design standardization. Attention is also devoted to the identification of plant layout solutions able to avoid risks of radioactive release from both the core and other structures (i.e. the spent fuel storage pools). Specific effort is paid to the development of guidelines and design recommendations for addressing the seismic issue in next generation reactor systems. In addition, consideration will be devoted to transfer the knowledge developed in the project to Generation III advanced systems, in line with the objective of the SNE-TP SRA to support present and future Light Water Reactors and their further development, for which safety issues are key aspects to be addressed. Note, in this respect, that the benefits of base isolation in terms of response to design seismic actions are already widely recognized for Generation III LWRs, along with the possibility of a significant standardization of structural and equipment design. SILER activities started on October 1 st 2011 and are carried out by 18 partners: ENEA (Italy, Coordinator), AREVA NP SAS (France), SCK-CEN (Belgium), FIP Industriale (Italy), MAURER SOHENE (Germany), EC-JRC (Ispra (Italy)), SINTEC (Italy), KTH (Sweden), BOA-BKT (Germany), IDOM (Spain), ANSALDO (Italy), IPUL (Latvia), NUMERIA (Italy), VCE (Austria), SRS (Italy), CEA (France), EA (Spain), NUVIA (France). (authors)

  13. Laser system for secondary cooling of {sup 87}Sr atoms

    SciTech Connect (OSTI)

    Khabarova, K Yu; Slyusarev, S N; Strelkin, S A; Belotelov, G S; Kostin, A S; Pal'chikov, Vitaly G; Kolachevsky, Nikolai N

    2012-11-30T23:59:59.000Z

    A laser system with a narrow generation line for secondary laser cooling of {sup 87}Sr atoms has been developed and investigated. It is planned to use ultracold {sup 87}Sr atoms loaded in an optical lattice in an optical frequency standard. To this end, a 689-nm semiconductor laser has been stabilised using an external reference ultrastable cavity with vibrational and temperature compensation near the critical point. The lasing spectral width was 80 Hz (averaging time 40 ms), and the frequency drift was at a level of 0.3 Hz s{sup -1}. Comparison of two independent laser systems yielded a minimum Allan deviation: 2 Multiplication-Sign 10{sup -14} for 300-s averaging. It is shown that this system satisfies all requirements necessary for secondary cooling of 87Sr atoms using the spectrally narrow {sup 1}S{sub 0} - {sup 3}P{sub 1} transition ({lambda} = 689 nm). (cooling of atoms)

  14. Staged membrane oxidation reactor system

    DOE Patents [OSTI]

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

    2013-04-16T23:59:59.000Z

    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.

  15. Staged membrane oxidation reactor system

    DOE Patents [OSTI]

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

    2012-09-11T23:59:59.000Z

    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.

  16. Staged membrane oxidation reactor system

    DOE Patents [OSTI]

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

    2014-05-20T23:59:59.000Z

    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.

  17. Thermal Storage with Conventional Cooling Systems

    E-Print Network [OSTI]

    McGee, E. E.

    1990-01-01T23:59:59.000Z

    demand which results in lower electrical costs. The effectiveness of this 'Thermal Retention System" is determined by its design characteristics, its operational efficiency and comparative system analysis. Today's computer technology has provided...

  18. Transient Accident Analysis of a Supercritical Carbon Dioxide Brayton Cycle Energy Converter Coupled to an Autonomous Lead-Cooled Fast Reactor

    SciTech Connect (OSTI)

    Moisseytsev, Anton; Sienicki, James J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2006-07-01T23:59:59.000Z

    The Supercritical Carbon Dioxide (S-CO{sub 2}) Brayton Cycle is a promising advanced alternative to the Rankine saturated steam cycle and recuperated gas Brayton cycle for the energy converters of specific reactor concepts belonging to the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. A new plant dynamics analysis computer code has been developed for simulation of the S-CO{sub 2} Brayton cycle coupled to an autonomous, natural circulation Lead-Cooled Fast Reactor (LFR). The plant dynamics code was used to simulate the whole-plant response to accident conditions. The specific design features of the reactor concept influencing passive safety are discussed and accident scenarios are identified for analysis. Results of calculations of the whole-plant response to loss-of-heat sink, loss-of-load, and pipe break accidents are demonstrated. The passive safety performance of the reactor concept is confirmed by the results of the plant dynamics code calculations for the selected accident scenarios. (authors)

  19. Transient accident analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to an autonomous lead-cooled fast reactor.

    SciTech Connect (OSTI)

    Moisseytsev, A.; Sienicki, J. J.; Nuclear Engineering Division

    2008-08-01T23:59:59.000Z

    The supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle is a promising advanced alternative to the Rankine steam cycle and recuperated gas Brayton cycle for the energy converters of specific reactor concepts belonging to the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. A new plant dynamics analysis computer code has been developed for simulation of the S-CO{sub 2} Brayton cycle coupled to an autonomous, natural circulation lead-cooled fast reactor (LFR). The plant dynamics code was used to simulate the whole-plant response to accident conditions. The specific design features of the reactor concept influencing passive safety are discussed and accident scenarios are identified for analysis. Results of calculations of the whole-plant response to loss-of-heat sink, loss-of-load, and pipe break accidents are demonstrated. The passive safety performance of the reactor concept is confirmed by the results of the plant dynamics code calculations for the selected accident scenarios.

  20. FISSION REACTORS KEYWORDS: high-temperature

    E-Print Network [OSTI]

    Yildiz, Bilge

    that is directly cou- pled to an advanced gas-cooled reactor (AGR) is pro- posed in this paper. The system features conversion system, and the progress in the electrolysis cell materials field can help the econom- ical by a supercritical CO2 ~SCO2! power conversion system that is directly coupled to an advanced gas-cooled reactor

  1. Radiation detector system having heat pipe based cooling

    DOE Patents [OSTI]

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

    2006-10-31T23:59:59.000Z

    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.

  2. 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

    1982-06-01T23:59:59.000Z

    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.

  3. 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)

    1986-01-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Mohsen Shadmehri; Jamshid Ghanbari

    2002-04-06T23:59:59.000Z

    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.

  5. The unique safety challenges of space reactor systems

    SciTech Connect (OSTI)

    Lanes, S.J. (Dept. of Energy, Washington, DC (United States)); Marshall, A.C. (Sandia National Lab., Albuquerque, NM (United States))

    1991-01-01T23:59:59.000Z

    Compact reactor systems can provide high levels of power for extended periods in space environments. Their relatively low mass and their ability to operate independently of their proximity to the sun make reactor power systems high desirable for many civilian and military space missions. The US Department of Energy is developing reactor system technologies to provide electrical power for space applications. In addition, reactors are now being considered to provide thermal power to a hydrogen propellant for nuclear thermal rocketry. Space reactor safety issues differ from commercial reactor issues, in some areas, because of very different operating requirements and environments. Accidents similar to those postulated for commercial reactors must be considered for space reactors during their operational phase. Safety strategies will need to be established that account for the consequences of the loss of essential power.

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

    SciTech Connect (OSTI)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01T23:59:59.000Z

    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.

  7. 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)

    2008-09-23T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

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

    1993-12-07T23:59:59.000Z

    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.

  9. 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)

    1993-01-01T23:59:59.000Z

    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.

  10. Development and validation of scale nuclear analysis methods for high temperature gas-cooled reactors

    SciTech Connect (OSTI)

    Gehin, Jess C [ORNL] [ORNL; Jessee, Matthew Anderson [ORNL] [ORNL; Williams, Mark L [ORNL] [ORNL; Lee, Deokjung [ORNL] [ORNL; Goluoglu, Sedat [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL; Ilas, Dan [ORNL] [ORNL; Bowman, Steve A [ORNL] [ORNL

    2010-01-01T23:59:59.000Z

    In support of the U.S. Nuclear Regulatory Commission, ORNL is updating the nuclear analysis methods and data in the SCALE code system to support modeling of HTGRs. Development activities include methods used for reactor physics, criticality safety, and radiation shielding. This paper focuses on the nuclear methods in support of reactor physics, which primarily include lattice physics for cross-section processing of both prismatic and pebble-bed designs, Monte Carlo depletion methods and efficiency improvements for double heterogeneous fuels, and validation against relevant experiments. These methods enhancements are being validated using available experimental data from the HTTR and HTR-10 startup and initial criticality experiments. Results obtained with three-dimensional Monte Carlo models of the HTTR initial core critical configurations with SCALE6/KENO show excellent agreement between the continuous energy and multigroup methods and the results are consistent with results obtained by others. A three-dimensional multigroup Monte Carlo model for the initial critical core of the HTR-10 has been developed with SCALE6/KENO based on the benchmark specifications included in the IRPhE Handbook. The core eigenvalue obtained with this model is in very good agreement with the corresponding value obtained with a consistent continuous energy MCNP5 core model.

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

    SciTech Connect (OSTI)

    J. M. Beck; L. F. Pincock

    2011-04-01T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Lee, Ching-Pang

    2014-12-30T23:59:59.000Z

    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.

  13. Weld monitor and failure detector for nuclear reactor system

    DOE Patents [OSTI]

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

    1987-01-01T23:59:59.000Z

    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.

  14. BETTER DUCT SYSTEMS FOR HOME HEATING AND COOLING.

    SciTech Connect (OSTI)

    ANDREWS,J.

    2001-01-01T23:59:59.000Z

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

    DOE Patents [OSTI]

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

    1982-08-19T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

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

    1998-01-01T23:59:59.000Z

    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.

  17. e-Cooling High Cavity & Cryomodule Systems, Inc.

    E-Print Network [OSTI]

    Beta Cavity & Cryomodule Final Design Review Cryomodule Design Brookhaven National Laboratory July 22;e-Cooling High Cavity & Cryomodule Advanced Energy Systems, Inc. Slide 7 of 24 Cavity Cold Model) 40.0 85.0 Wall Thickness (mm) 4 mm 3 mm Cavity Configuration Freq. Cells Tuner Load (400 kHz) Tuning

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

    E-Print Network [OSTI]

    Watson, Craig A.

    when the need is discovered, but a good preventive maintenance program will reduce the number. This fact sheet will emphasize corrective and preventive maintenance procedures for ventilation, evaporativeAE26 Maintenance Guide for Greenhouse Ventilation, Evaporative Cooling Heating Systems1 D. E

  19. Simulation of wind driven ventilative cooling systems for an apartment building in Beijing and Shanghai

    E-Print Network [OSTI]

    Chen, Qingyan "Yan"

    1 Simulation of wind driven ventilative cooling systems for an apartment building in Beijing., Glicksman, L.R. and Norford, L.K. 2002. "Simulation of wind driven ventilative cooling systems evaluation of two passive cooling strategies, daytime ventilation and night cooling, for a generic, six

  20. Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

    SciTech Connect (OSTI)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1984-06-01T23:59:59.000Z

    ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.

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

    E-Print Network [OSTI]

    Yan Feng; Bin Liu; J. Goree

    2011-04-19T23:59:59.000Z

    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.

  2. Developing, testing, evaluating and optimizing solar heating and cooling systems

    SciTech Connect (OSTI)

    Not Available

    1991-11-01T23:59:59.000Z

    The objective is to develop and test various integrated solar heating, cooling and domestic hot water systems, and to evaluate their performance. Systems composed of new, as well a previously tested, components are carefully integrated so that effects of new components on system performance can be clearly delineated. The SEAL-DOE program includes six tasks which have received funding for the 1991--1992 fifteen-month period. These include: (1) a project employing isothermal operation of air and liquid solar space hearing systems, (2) a project to build and test several generic solar water heaters, (3) a project that will evaluate advanced solar domestic hot water components and concepts and integrate them into solar domestic hot water systems, (4) a liquid desiccant cooling system development project, (5) a project that will perform system modeling and analysis work on solid desiccant cooling systems research, and (6) a management task. The objectives and progress in each task are described in this report. 6 figs., 2 tabs.

  3. Current cost and performance requirements for residential cool storage systems

    SciTech Connect (OSTI)

    Brown, D.R.; Spanner, G.E.

    1988-08-01T23:59:59.000Z

    This study defines the current cost and performance requirements for residential cool storage technologies based on the characteristics of conventional air conditioning equipment and residential time-of-day (TOD) rate structures existing during the 1986--1987 time frame. Currently, rate structures are changing rapidly. Given the volatility of rate structures, the establishment of cost goal is challenging. The goals presented in this study are based on the utility rate structure as of 1986. This study serves to define residential cool storage cost and performance requirements in the current economic environment as well as the many issues affecting the requirements for residential cool storage systems both now and in the future. The same methodology can be employed to establish long-run goals once future rate structures are adequately defined. 12 refs., 6 figs., 18 tabs.

  4. The TREAT upgrade manual reactor control system and its interface with the automatic reactor control system and the plant protection system

    SciTech Connect (OSTI)

    McDowell, W.P.

    1986-02-01T23:59:59.000Z

    The Transient Reactor Test Facility (TREAT) at Argonne National Laboratory is being upgraded to simulate extreme conditions in a reactor. This facility will be used to subject test assemblies of fuel bundles to very rapid and intense power transients. This paper describes in detail the manual reactor control system and its interfaces with the plant protection system the automatic reactor control system.

  5. Integrated systems analysis of the PIUS reactor

    SciTech Connect (OSTI)

    Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1993-11-01T23:59:59.000Z

    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.

  6. Method and system for powering and cooling semiconductor lasers

    DOE Patents [OSTI]

    Telford, Steven J; Ladran, Anthony S

    2014-02-25T23:59:59.000Z

    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. Redesigning Process Cooling Systems in the Plastics Industry

    E-Print Network [OSTI]

    Anderson, G. R.

    2006-01-01T23:59:59.000Z

    towers during the winter. Lifetime’s electric provider, Utah Power, manages a demand side management program (DSM) and hired etc Group, Inc to evaluate the cooling systems for potential energy efficiency improvements. etc Group, Inc and Lifetime... basketball systems in 1973 and became Lifetime Products in 1986. Lifetime introduced the first blow-molded plastic table in 1995. In the 120,000 square foot Clearfield, UT facility discussed in this paper, Lifetime blow molds plastic folding tables...

  8. advanced gas cooled graphite moderated reactor: Topics by E-print...

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

    temperatures during normal (more) Moore, Eugene James Thomas 2006-01-01 2 THORIUM FUEL CYCLES: A GRAPHITE-MODERATED MOLTEN SALT REACTOR Physics Websites Summary: ,...

  9. air-cooled graphite reactors: Topics by E-print Network

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

    an estimated 1,282 cubic yards of contaminated steel and filter elements from 4 THORIUM FUEL CYCLES: A GRAPHITE-MODERATED MOLTEN SALT REACTOR Physics Websites Summary: ,...

  10. Application of GRS method to evaluation of uncertainties of calculation parameters of perspective sodium-cooled fast reactor

    SciTech Connect (OSTI)

    Peregudov, A.; Andrianova, O.; Raskach, K.; Tsibulya, A. [Inst. for Physics and Power Engineering, Bondarenko Square 1, Obninsk 244033, Kaluga Region (Russian Federation)

    2012-07-01T23:59:59.000Z

    A number of recent studies have been devoted to the estimation of errors of reactor calculation parameters by the GRS (Generation Random Sampled) method. This method is based on direct sampling input data resulting in formation of random sets of input parameters which are used for multiple calculations. Once these calculations are performed, statistical processing of the calculation results is carried out to determine the mean value and the variance of each calculation parameter of interest. In our study this method is used for estimation of errors of calculation parameters (K{sub eff}, power density, dose rate) of a perspective sodium-cooled fast reactor. Neutron transport calculations were performed by the nodal diffusion code TRIGEX and Monte Carlo code MMK. (authors)

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

    E-Print Network [OSTI]

    Alhashimi, Tariq Yaqoob Sayed

    2014-12-02T23:59:59.000Z

    welded to the upper plenum which has two exhaust chimneys. Blowers are used to drive air through in-line heaters which are connected to the bottom end of the riser ducts. Experiments were conducted to measure the temperature spatial profile in the plenum...

  12. Experimental study on the operational and the cooling performance of the APR+ passive auxiliary feedwater system

    SciTech Connect (OSTI)

    Kang, K. H.; Bae, B. U.; Kim, S.; Cho, Y. J.; Park, Y. S.; Kim, B. D. [Korea Atomic Energy Research Inst., 150 Dukjin-dong, Yuseong-gu, Daejeon (Korea, Republic of)

    2012-07-01T23:59:59.000Z

    The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+ which is intended to completely replace the conventional active auxiliary feedwater system. The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by introducing a natural driving force mechanism; i.e., condensing steam in nearly-horizontal U-tubes submerged inside the passive condensation cooling tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, the separate effect test, PASCAL (PAFS Condensing Heat Removal Assessment Loop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. A single nearly-horizontal U-tube whose dimension is same as the prototypic U-tube of the APR+ PAFS is simulated in the PASCAL test. By performing the PASCAL test, the major thermal-hydraulic parameters such as local/overall heat transfer coefficients, fluid temperature inside the tube, wall temperature of the tube, and pool temperature distribution in the PCCT were produced not only to evaluate the current condensation heat transfer model but also to present database for the safety analysis related with the PAFS. (authors)

  13. anaerobic reactor systems: Topics by E-print Network

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

    Thermonuclear Experimental Reactor and the Nuclear Weapons Proliferation Implications of Thermonuclear-Fusion Energy Systems CiteSeer Summary: This paper contains two parts: (I) A...

  14. annular reactor system: Topics by E-print Network

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

    Thermonuclear Experimental Reactor and the Nuclear Weapons Proliferation Implications of Thermonuclear-Fusion Energy Systems CiteSeer Summary: This paper contains two parts: (I) A...

  15. Shutdown heat removal system reliability in thermal reactors

    SciTech Connect (OSTI)

    Sun, Y.H.; Bari, R.A.

    1980-01-01T23:59:59.000Z

    An analysis of the failure probability per year of the shutdown heat removal system (SHRS) at hot standby conditions for two thermal reactor designs is presented. The selected reactor designs are the Pressurized Water Reactor and the Nonproliferation Alternative System Assessment Program Heavy Water Reactor. Failures of the SHRS following the initiating transients of loss of offsite power and loss of main feedwater system are evaluated. Common mode failures between components are incorporated in this anlaysis via the ..beta..-factor method and the sensitivity of the system reliability to common mode failures is investigated parametrically.

  16. 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

    2006-02-07T23:59:59.000Z

    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.

  17. 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.

    2006-01-17T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01T23:59:59.000Z

    cooling system is well positioned for effectively cooling higher-density servers in today’s and futurecooling systems aim at significantly improving efficiency while addressing the thermal challenges, improving reliability, and allowing for future

  19. Preliminary results of a dynamic system model for a closed-loop Brayton cycle coupled to a nuclear reactor.

    SciTech Connect (OSTI)

    Wright, Steven Alan

    2003-06-01T23:59:59.000Z

    This paper describes preliminary results of a dynamic system model for a closed-loop Brayton-cycle that is coupled to a nuclear reactor. The current model assumes direct coupling between the reactor and the Brayton-cycle, however only minor additions are required to couple the Brayton-cycle through a heat exchanger to either a heat pipe reactor or a liquid metal cooled reactor. Few reactors have ever been coupled to closed Brayton-cycle systems. As such their behavior under dynamically varying loads, startup and shut down conditions, and requirements for safe and autonomous operation are largely unknown. Sandia National Laboratories has developed steady-state and dynamic models for closed-loop turbo-compressor systems (for space and terrestrial applications). These models are expected to provide a basic understanding of the dynamic behavior and stability of the coupled reactor and power generation loop. The model described in this paper is a lumped parameter model of the reactor, turbine, compressor, recuperator, radiator/waste-heat-rejection system and generator. More detailed models that remove the lumped parameter simplifications are also being developed but are not presented here. The initial results of the model indicate stable operation of the reactor-driven Brayton-cycle system and its ability to load-follow. However, the model also indicates some counter-intuitive behavior for the complete coupled system. This behavior will require the use of a reactor control system to select an appropriate reactor operating temperature that will optimize the performance of the complete spacecraft system. We expect this model and subsequent versions of it to provide crucial information in developing procedures for safe start up, shut down, safe-standby, and other autonomous operating modes. Ultimately, Sandia hopes to validate these models and to perform nuclear ground tests of reactor-driven closed Brayton-cycle systems in our nuclear research facilities.

  20. Quantum Friction: Cooling Quantum Systems with Unitary Time Evolution

    E-Print Network [OSTI]

    Aurel Bulgac; Michael McNeil Forbes; Kenneth J. Roche; Gabriel Wlaz?owski

    2013-05-29T23:59:59.000Z

    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.

  1. 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

    2011-05-25T23:59:59.000Z

    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.

  2. Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

    SciTech Connect (OSTI)

    Hikaru Hiruta; Gilles Youinou

    2013-09-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01T23:59:59.000Z

    cooling systems aim at significantly improving efficiency while addressing the thermal challenges, improving reliability, and allowing for future

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

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01T23:59:59.000Z

    Cooling System in Data Centers 1 Problem Statement Scientific and enterprise data centers, IT equipment product development, and research

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

    E-Print Network [OSTI]

    Xu, TengFang

    2009-01-01T23:59:59.000Z

    Cooling System in Data Centers 1 Problem Statement Scientific and enterprise data centers, IT equipment product development, and research

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

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    Florida, USA (1997). [34] P. HEJZLAR et. al. “Traveling WaveLaboratory, 2013. [76] P. HEJZLAR and C. B. DAVIS. “studies on TWRs Yarsky, Hejzlar, Driscoll (MIT) Gas-cooled

  7. Nuclear design of small-sized high temperature gas-cooled reactor for developing countries

    SciTech Connect (OSTI)

    Goto, M.; Seki, Y.; Inaba, Y.; Ohashi, H.; Sato, H.; Fukaya, Y.; Tachibana, Y. [Japan Atomic Energy Agency, 4002, Oarai-machi, Higashi Ibaraki-gun, Ibaraki-ken 311-1394 (Japan)

    2012-07-01T23:59:59.000Z

    Japan Atomic Energy Agency (JAEA) has started a conceptual design of a small-sized HTGR with 50 MW thermal power (HTR50S), which is a first-of-a-kind commercial or demonstration plant of a small-sized HTGR to be deployed in developing countries such as Kazakhstan in the 2020's. The nuclear design of the HTR50S is performed by upgrading the proven technology of the High Temperature Engineering Test Reactor (HTTR) to reduce the cost for the construction. In the HTTR design, twelve kinds of fuel enrichment was used to optimize the power distribution, which is required to make the maximum fuel temperature below the thermal limitation during the burn-up period. However, manufacture of many kinds of fuel enrichment causes increase of the construction cost. To solve this problem, the present study challenges the nuclear design by reducing the number of fuel enrichment to as few as possible. The nuclear calculations were performed with SRAC code system whose validity was proven by the HTTR burn-up data. The calculation results suggested that the optimization of the power distribution was reasonably achieved and the maximum fuel temperature was kept below the limitation by using three kinds of fuel enrichment. (authors)

  8. 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

  9. 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

  10. 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

  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. Improving the Efficiency of Your Process Cooling System

    E-Print Network [OSTI]

    Baker, R.

    2005-01-01T23:59:59.000Z

    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...

  14. Cooling system of an internal combustion engine having a turbo-charger

    SciTech Connect (OSTI)

    Hasegawa, M.; Fukuda, T.

    1986-09-02T23:59:59.000Z

    A cooling system of an internal combustion engine is described having a turbo-charger, comprising a cooling water circulation passageway filled with cooling water for cooling the engine including at least a cylinder head cooling portion, a cooling water circulation passageway for cooling the turbo-charger including a turbo-charger cooling portion, and means for supplying a part of the engine cooling water to the turbo-charger cooling water ciruclation passageway and returning it from there to the engine cooling water cirulation passageway, characterized in that the turbo-charger cooling portion is positioned at the same level or higher than the cylinder head cooling portion of the engine, the turbo-charger cooling water circulation passageway includes a water volume positioned at a level higher than the turbo-charger cooling portion. The volume is connected to a cooling water reservoir tank via a pressure relief valve which is opened when pressure in the volume exceeds a predetermined value to supply cooling water to the volume.

  15. Fluid sampling system for a nuclear reactor

    DOE Patents [OSTI]

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

    1994-01-01T23:59:59.000Z

    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.

  16. Fluid sampling system for a nuclear reactor

    DOE Patents [OSTI]

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

    1994-11-22T23:59:59.000Z

    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.

  17. 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

    2005-02-13T23:59:59.000Z

    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.

  18. Survival of zooplankton entrained into the cooling water system and supplemental cooling towers of a steam-electric generating station located on Galveston Bay, Texas

    E-Print Network [OSTI]

    Chase, Cathleen Louise

    1977-01-01T23:59:59.000Z

    is not an unlimited resource. Another method supplements the open ? cycle system with external cooling facilities, through which the heated water passes before it flows into the receiving body. Ex- ternal cooling facilities may be wet-cooling towers, dry-cooling...

  19. POOL WATER TREATMENT AND COOLING SYSTEM DESCRIPTION DOCUMENT

    SciTech Connect (OSTI)

    V. King

    2000-06-19T23:59:59.000Z

    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.

  20. 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

    2005-01-01T23:59:59.000Z

    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.