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Note: This page contains sample records for the topic "reactor cooling system" from the National Library of EnergyBeta (NLEBeta).
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1

Liquid metal cooled nuclear reactor plant system  

DOE Patents (OSTI)

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.

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

1993-01-01T23:59:59.000Z

2

Indirect passive cooling system for liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

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.

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

1990-01-01T23:59:59.000Z

3

Cooling system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1982-01-01T23:59:59.000Z

4

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

DOE Patents (OSTI)

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.

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

1991-01-01T23:59:59.000Z

5

Nuclear reactor cooling system decontamination reagent regeneration  

DOE Patents (OSTI)

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.

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

6

Natural circulating passive cooling system for nuclear reactor containment structure  

DOE Patents (OSTI)

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.

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

1990-01-01T23:59:59.000Z

7

Passive cooling system for nuclear reactor containment structure  

DOE Patents (OSTI)

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.

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

1989-01-01T23:59:59.000Z

8

Analysis of the Reactor Cavity Cooling System for Very High Temperature Gas-cooled Reactors Using Computational Fluid Dynamics Tools  

E-Print Network (OSTI)

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 an important heat removal system in case of accidents...

Frisani, Angelo

2011-08-08T23:59:59.000Z

9

Experimental and Computational Study of a Scaled Reactor Cavity Cooling System  

E-Print Network (OSTI)

The Very High Temperature Gas-Cooled Reactor (VHTR) is one of the next generation nuclear reactors designed to achieve high temperatures to support industrial applications and power generation. The Reactor Cavity Cooling System (RCCS) is a passive...

Vaghetto, Rodolfo

2013-11-25T23:59:59.000Z

10

A gas-cooled reactor surface power system  

Science Journals Connector (OSTI)

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed depending on the number of astronauts level of scientific activity and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

Ronald J. Lipinski; Steven A. Wright; Roger X. Lenard; Gary A. Harms

1999-01-01T23:59:59.000Z

11

Modification of the Core Cooling System of TRIGA 2000 Reactor  

SciTech Connect

To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24 deg. C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

Umar, Efrizon; Fiantini, Rosalina [National Nuclear Energy Agency of Indonesia, Jalan Tamansari 71, Bandung, 40132 (Indonesia)

2010-06-22T23:59:59.000Z

12

Experimental Studies of NGNP Reactor Cavity Cooling System With Water  

SciTech Connect

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.

Michael Corradini; Mark Anderson; Yassin Hassan; Akira Tokuhiro

2013-01-16T23:59:59.000Z

13

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

DOE Patents (OSTI)

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.

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

1980-06-06T23:59:59.000Z

14

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

SciTech Connect

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.

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

2014-07-14T23:59:59.000Z

15

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

SciTech Connect

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.

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

2010-11-02T23:59:59.000Z

16

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

SciTech Connect

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)

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

17

Method and apparatus for enhancing reactor air-cooling system performance  

DOE Patents (OSTI)

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.

Hunsbedt, Anstein (Los Gatos, CA)

1996-01-01T23:59:59.000Z

18

Method and apparatus for enhancing reactor air-cooling system performance  

DOE Patents (OSTI)

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.

Hunsbedt, A.

1996-03-12T23:59:59.000Z

19

Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system  

SciTech Connect

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 RVACS can be characterized as a surface-based decay heat removal system. It 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 the core volume) and decay heat removal capability (proportional to the vessel surface area). 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 environmental 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 or annular regions formed 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 among 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 density and therefore the reactor power can be significantly increased, without losing the passive heat removal feature. This paper introduces the concept of using DRACS to enhance VHTR passive safety and economics. Three design options with different cooling pipe locations are discussed. Analysis results from a lumped volume based model and CFD simulations are presented. (authors)

Zhao, H.; Zhang, H.; Zou, L. [Idaho National Laboratory (United States); Sun, X. [Ohio State Univ. (United States)

2012-07-01T23:59:59.000Z

20

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

DOE Patents (OSTI)

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.

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

1993-01-01T23:59:59.000Z

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


21

Heat pipe based passive emergency core cooling system for safe shutdown of nuclear power reactor  

Science Journals Connector (OSTI)

Abstract On March 11th, 2011, a natural disaster created by earthquakes and Tsunami caused a serious potential of nuclear reactor meltdown in Fukushima due to the failure of Emergency Core Cooling System (ECCS) powered by diesel generators. In this paper, heat pipe based ECCS has been proposed for nuclear power plants. The designed loop type heat pipe ECCS is composed of cylindrical evaporator with 62 vertical tubes, each 150 mm diameter and 6 m length, mounted around the circumference of nuclear fuel assembly and 21 m × 10 m × 5 m naturally cooled finned condenser installed outside the primary containment. Heat pipe with overall thermal resistance of 1.44 × 10?5 °C/W will be able to reduce reactor temperature from initial working temperature of 282 °C to below 250 °C within 7 h. The overall ECCS also includes feed water flooding of the core using elevated water tank for initial 10 min which will accelerate cooling of the core, replenish core coolant during loss of coolant accident and avoids heat transfer crisis phenomena during heat pipe start-up process. The proposed heat pipe system will operate in fully passive mode with high runtime reliability and therefore provide safer environment to nuclear power plants.

Masataka Mochizuki; Randeep Singh; Thang Nguyen; Tien Nguyen

2014-01-01T23:59:59.000Z

22

High Temperature Gas-Cooled Reactor Program. Modular HTGR systems design and cost summary. [Methane reforming; steam cycle-cogeneration  

SciTech Connect

This report provides a summary description of the preconceptual design and energy product costs of the modular High Temperature Gas-Cooled Reactor (HTGR). The reactor system was studied for two applications: (1) reforming of methane to produce synthesis gas and (2) steam cycle/cogeneration to produce process steam and electricity.

Not Available

1983-09-01T23:59:59.000Z

23

Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology  

Science Journals Connector (OSTI)

Abstract Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components.

Mukesh Kumar; Aranyak Chakravarty; A.K. Nayak; Hari Prasad; V. Gopika

2014-01-01T23:59:59.000Z

24

Design and optimization of the heat rejection system for a liquid cooled thermionic space nuclear reactor power system  

SciTech Connect

The heat transport subsystem for a liquid metal cooled thermionic space nuclear power system was modelled using algorithms developed in support of previous nuclear power system study programs, which date back to the SNAP-10A flight system. The model was used to define the optimum dimensions of the various components in the heat transport subsystem subjected to the constraints of minimizing mass and achieving a launchable package that did not require radiator deployment. The resulting design provides for the safe and reliable cooling of the nuclear reactor in a proven lightweight design.

Moriarty, M.P. (Rocketdyne Division, Rockwell International Corporation, 6633 Canoga Avenue, P.O. Box 7922, Canoga Park, California 91309-7922 (United States))

1993-01-15T23:59:59.000Z

25

Incorporating reliability analysis into the design of passive cooling systems with an application to a gas-cooled reactor  

Science Journals Connector (OSTI)

A time-dependent reliability evaluation of a two-loop passive decay heat removal (DHR) system was performed as part of the iterative design process for a helium-cooled fast reactor. The system was modeled using RELAP5-3D. The uncertainties in input parameters were assessed and were propagated through the model using Latin hypercube sampling. An important finding was the discovery that the smaller pressure loss through the DHR heat exchanger than through the core would make the flow to bypass the core through one DHR loop, if two loops operated in parallel. This finding is a warning against modeling only one lumped DHR loop and assuming that n of them will remove n times the decay power. Sensitivity analyses revealed that there are values of some input parameters for which failures are very unlikely. The calculated conditional (i.e., given the LOCA) failure probability was deemed to be too high leading to the identification of several design changes to improve system reliability. This study is an example of the kinds of insights that can be obtained by including a reliability assessment in the design process. It is different from the usual use of PSA in design, which compares different system configurations, because it focuses on the thermal–hydraulic performance of a safety function.

Francisco J. Mackay; George E. Apostolakis; Pavel Hejzlar

2008-01-01T23:59:59.000Z

26

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

E-Print Network (OSTI)

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...

Wu, Huali

2013-08-08T23:59:59.000Z

27

The computational-and-experimental investigation into the head-flow characteristic of the two-stage ejector for the emergency core cooling system of the NPP with a water-moderated water-cooled power reactor  

Science Journals Connector (OSTI)

The results of the computational-and-experimental investigation into the two-stage ejector for the emergency cooling system of the core of the water-moderated water-cooled power reactor. The results of experiment...

Yu. V. Parfenov

2013-09-01T23:59:59.000Z

28

Scaling Analysis for the Direct Reactor Auxillary Cooling System For AHTRS  

SciTech Connect

The Direct Reactor Auxiliary Cooling System (DRACS) is a passive heat removal system proposed for the Advanced High-Temperature Reactor (AHTR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops relying completely on buoyancy as the driving force. In the DRACS, two heat exchangers, namely, the DRACS Heat Exchanger (DHX) and the Natural Draft Heat Exchanger (NDHX) are used to couple these loops. In addition, a fluidic diode is employed to minimize the parasitic flow during normal operation of the reactor and to activate the DRACS in accidents. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for AHTRs built and tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of the scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional flow formulation, the governing equations are non-dimensionalized by introducing appropriate dimensionless parameters. The key dimensionless numbers that characterize the DRACS system are obtained straightforwardly from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has also been developed, which consists of the core scaling and loop scaling. The consistence between the core and loop scaling is examined through the reference volume ratio, which can be obtained from the core and loop scaling processes. The scaling methodology and similarity laws have been applied to obtain a design of the scaled-down high-temperature DRACS test facility (HTDF).

Lv, Q. NMN [Ohio State University] [Ohio State University; Wang, X. NMN [Ohio State University] [Ohio State University; Sun, X NMN [Ohio State University] [Ohio State University; Christensen, R. N. [Ohio State University] [Ohio State University; Blue, T. E. [Ohio State University] [Ohio State University; Yoder Jr, Graydon L [ORNL] [ORNL; Wilson, Dane F [ORNL] [ORNL; Subharwall, Piyush [Idaho National Laboratory (INL)] [Idaho National Laboratory (INL); Adams, I. [Ohio State University, Columbus] [Ohio State University, Columbus

2013-01-01T23:59:59.000Z

29

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

DOE Patents (OSTI)

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.

DeVolpi, A.

1984-07-20T23:59:59.000Z

30

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

SciTech Connect

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.

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

2009-03-01T23:59:59.000Z

31

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

E-Print Network (OSTI)

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 ...

Gibbs, Jonathan Paul

2008-01-01T23:59:59.000Z

32

Reactor water cleanup system  

DOE Patents (OSTI)

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.

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

1994-01-01T23:59:59.000Z

33

Reactor water cleanup system  

DOE Patents (OSTI)

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.

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

1994-12-20T23:59:59.000Z

34

Numerical simulation of a blanket cooling system for fusion reactor based on PWR conditions  

Science Journals Connector (OSTI)

The simulations of a blanket cooling system were presented to address the choice of cooling channel geometry and coolant input data which are related to blanket engineering implementation. This work was performed using computer aided design (CAD) and computational fluid dynamics (CFD) technology. Simulations were carried out for the blanket module with a size of 0.6 m × 0.45 m in toroidal plane, and the nuclear heat was applied on the cooling system at Pn (neutron wall load) of 5 MW/m2. The structure factors and input data of hydraulics were investigated to explore the optimal parameters to match the PWR condition. It was found that the inlet velocity of first wall (FW) channel should be within the range of 2.48–3.34 m/s. As a result, the temperature rise (TR) of the coolant in the FW channel would be 24–25 K. This leads to the remaining space for TR within the range of 15 K in the piping circuits. It also indicated that the FW plays an important role in TR (reaches 60% of the whole cooling system) due to its high level of Pn and heat flux in the zones. It was predicted that the nuclear heat inside blanket module could be removed completely by the piping circuits with an acceptable pipe bore and the related input data. Finally, a possible design range of cooling parameters was proposed in view of engineering feasibility and blanket neutronics design.

Changle Liu; Jianzhong Zhang; Yinfeng Zhu; Songlin Liu; Xuebin Ma; Peiming Chen

2013-01-01T23:59:59.000Z

35

Gas-cooled nuclear reactor  

DOE Patents (OSTI)

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.

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

1985-01-01T23:59:59.000Z

36

Development of Materials for Supercritical-Water-Cooled Reactor |  

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

Development of Materials for Supercritical-Water-Cooled Reactor Development of Materials for Supercritical-Water-Cooled Reactor Development of Materials for Supercritical-Water-Cooled Reactor 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 simplification, the R&D cost minimization and the flexibility for core design. As the demand for advanced nuclear system increases, Japanese R&D project started in 1999 aiming to provide technical information essential to demonstration of SCPR technologies through three sub-themes of 1. Plant conceptual design, 2. Thermal-hydraulics, and 3. Material. Although the material development is critical issue of SCWR development, previous studies were limited for the screening tests on commercial alloys

37

Alternative cooling resource for removing the residual heat of reactor  

SciTech Connect

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)

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

38

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

SciTech Connect

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.

Rodolfo Vaghetto; Luigi Capone; Yassin A. Hassan

2011-05-31T23:59:59.000Z

39

Solvent refined coal reactor quench system  

DOE Patents (OSTI)

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.

Thorogood, R.M.

1983-11-08T23:59:59.000Z

40

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)

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...

Vaghetto, Rodolfo

2012-07-16T23:59:59.000Z

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


41

Fuel Cycle System Analysis Implications of Sodium-Cooled Metal-Fueled Fast Reactor Transuranic Conversion Ratio  

SciTech Connect

If advanced fuel cycles are to include a large number of fast reactors (FRs), what should be the transuranic (TRU) conversion ratio (CR)? The nuclear energy era started with the assumption that they should be breeder reactors (CR > 1), but the full range of possible CRs eventually received attention. For example, during the recent U.S. Global Nuclear Energy Partnership program, the proposal was burner reactors (CR < 1). Yet, more recently, Massachusetts Institute of Technology's "Future of the Nuclear Fuel Cycle" proposed CR [approximately] 1. Meanwhile, the French company EDF remains focused on breeders. At least one of the reasons for the differences of approach is different fuel cycle objectives. To clarify matters, this paper analyzes the impact of TRU CR on many parameters relevant to fuel cycle systems and therefore spans a broad range of topic areas. The analyses are based on a FR physics parameter scan of TRU CR from 0 to [approximately]1.8 in a sodium-cooled metal-fueled FR (SMFR), in which the fuel from uranium-oxide-fueled light water reactors (LWRs) is recycled directly to FRs and FRs displace LWRs in the fleet. In this instance, the FRs are sodium cooled and metal fueled. Generally, it is assumed that all TRU elements are recycled, which maximizes uranium ore utilization for a given TRU CR and waste radiotoxicity reduction and is consistent with the assumption of used metal fuel separated by electrochemical means. In these analyses, the fuel burnup was constrained by imposing a neutron fluence limit to fuel cladding to the same constant value. This paper first presents static, time-independent measures of performance for the LWR [right arrow] FR fuel cycle, including mass, heat, gamma emission, radiotoxicity, and the two figures of merit for materials for weapon attractiveness developed by C. Bathke et al. No new fuel cycle will achieve a static equilibrium in the foreseeable future. Therefore, additional analyses are shown with dynamic, time-dependent measures of performance including uranium usage, TRU inventory, and radiotoxicity to evaluate the complex impacts of transition from the current uranium-fueled LWR system, and other more realistic impacts that may not be intuited from the time-independent steady-state conditions of the end-state fuel cycle. These analyses were performed using the Verifiable Fuel Cycle Simulation Model VISION. Compared with static calculations, dynamic results paint a different picture of option space and the urgency of starting a FR fleet. For example, in a static analysis, there is a sharp increase in uranium utilization as CR exceeds 1.0 (burner versus breeder). However, in dynamic analyses that examine uranium use over the next 1 to 2 centuries, behavior as CR crosses the 1.0 threshold is smooth, and other parameters such as the time required outside of reactors to recycle fuel become important. Overall, we find that there is no unambiguously superior value of TRU CR; preferences depend on the relative importance of different fuel cycle system objectives.

Steven J. Piet; Edward A. Hoffman; Samuel E. Bays; Gretchen E. Matthern; Jacob J. Jacobson; Ryan Clement; David W. Gerts

2013-03-01T23:59:59.000Z

42

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

E-Print Network (OSTI)

??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… (more)

Fong, Christopher J. (Christopher Joseph)

2008-01-01T23:59:59.000Z

43

Reliability study of a special decay heat removal system of a gas-cooled fast reactor demonstrator  

Science Journals Connector (OSTI)

Abstract The European roadmap toward the development of generation IV concepts addresses the safety and reliability assessment of the special system designed for decay heat removal of a gas-cooled fast reactor demonstrator (GFRD). The envisaged system includes the combination of both active and passive means to accomplish the fundamental safety function. Failure probabilities are calculated on various system configurations, according to either pressurized or depressurized accident events under investigation, and integrated with probabilities of occurrence of corresponding hardware components and natural circulation performance assessment. The analysis suggests the improvement of measures against common cause failures (CCF), in terms of an appropriate diversification among the redundant systems, to reduce the system failure risk. Particular emphasis is placed upon passive system reliability assessment, being recognized to be still an open issue, and the approach based on the functional reliability is adopted to address the point. Results highlight natural circulation as a challenging factor for the decay heat removal safety function accomplishment by means of passive devices. With the models presented here, the simplifying assumptions and the limited scenarios considered according to the level of definition of the design, where many systems are not yet established, one can conclude that attention has to be paid to the functional aspects of the passive system, i.e. the ones not pertaining to the “hardware” of the system. In this article the results of the analysis are discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The design diversity of the components undergoing \\{CCFs\\} can be effective for the improvement and some accident management measures are also possible by making use of the long grace period in GFRD.

Luciano Burgazzi

2014-01-01T23:59:59.000Z

44

Heat exchanger with auxiliary cooling system  

DOE Patents (OSTI)

A heat exchanger with an auxiliary cooling system capable of cooling a nuclear reactor should the normal cooling mechanism become inoperable. A cooling coil is disposed around vertical heat transfer tubes that carry secondary coolant therethrough and is located in a downward flow of primary coolant that passes in heat transfer relationship with both the cooling coil and the vertical heat transfer tubes. A third coolant is pumped through the cooling coil which absorbs heat from the primary coolant which increases the downward flow of the primary coolant thereby increasing the natural circulation of the primary coolant through the nuclear reactor.

Coleman, John H. (Salem Township, Westmoreland County, PA)

1980-01-01T23:59:59.000Z

45

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

E-Print Network (OSTI)

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 ...

Fong, Christopher J. (Christopher Joseph)

2008-01-01T23:59:59.000Z

46

Cooling System Analysis.  

E-Print Network (OSTI)

??ABSTRACT This master thesis report describes the behavior of a cooling system based on the power consumption and power losses during the velocity range. The… (more)

Cruz, Joăo Pedro Brás da

2012-01-01T23:59:59.000Z

47

Refueling Liquid-Salt-Cooled Very High-Temperature Reactors  

SciTech Connect

The liquid-salt-cooled very high-temperature reactor (LS-VHTR), also called the Advanced High-Temperature Reactor (AHTR), is a new reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. Depending upon goals, the peak coolant operating temperatures are between 700 and 1000 deg. C, with reactor outputs between 2400 and 4000 MW(t). Several fluoride salt coolants that are being evaluated have melting points between 350 and 500 deg. C, values that imply minimum refueling temperatures between 400 and 550 deg. C. At operating conditions, the liquid salts are transparent and have physical properties similar to those of water. A series of refueling studies have been initiated to (1) confirm the viability of refueling, (2) define methods for safe rapid refueling, and (3) aid the selection of the preferred AHTR design. Three reactor cores with different fuel element designs (prismatic, pebble bed, and pin-type fuel assembly) are being evaluated. Each is a liquid-salt-cooled variant of a graphite-moderated high-temperature reactor. The refueling studies examined applicable refueling experience from high-temperature reactors (similar fuel element designs) and sodium-cooled fast reactors (similar plant design with liquid coolant, high temperatures, and low pressures). The findings indicate that refueling is viable, and several approaches have been identified. The study results are described in this paper. (authors)

Forsberg, Charles W. [Oak Ridge National Laboratory, P.O. Box 2008 Oak Ridge, TN 37831 (United States); Peterson, Per F. [Nuclear Engineering Department, University of California at Berkeley, 6124a Etcheverry Hall, Berkeley, CA 94720 (United States); Cahalan, James E. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Enneking, Jeffrey A. [Areva NP (United States); Phil MacDonald [Consultant, Cedar Hill, TX (United States)

2006-07-01T23:59:59.000Z

48

THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS  

SciTech Connect

The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

David S. Duncan; Vondell J. Balls; Stephanie L. Austad

2008-09-01T23:59:59.000Z

49

Performance of Liquid Metals in Natural Circulation Cooled Nuclear Reactors  

SciTech Connect

The inherent safety capability of natural circulation makes reactor design more reliable. Additionally, the construction and operation of a nuclear power plant with natural circulation in the primary cooling circuit is an interesting alternative for nuclear plant designers, due to their lower operational and investment costs obtained by simplifying systems and controls. This paper deals with the feasibility of application of natural circulation in the primary cooling circuit of a liquid metal fast reactor. The methodology employed is a non-dimensional analysis, which describes the relationship between the physical properties and system variables. The performance criterion is bounded by a safety argument, referring to the maximum cladding temperature allowed during operation. The study considers several coolants, which can play a part in reactor cooling systems, such as lead, lead-bismuth and sodium. Bismuth and gallium are included in this analysis, in order to extend the range of properties for reference purposes. The results present a characterization of natural circulation flow in a reactor and compare the cooling capabilities from different liquid metals coolants. (authors)

Ceballos, Carlos; Lathouwers, Danny; Verkooijen, Adrian [Interfacultair Reactor Instituut, Technische Universiteit Delft, Mekelweg 15, Delft (Netherlands)

2004-07-01T23:59:59.000Z

50

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

E-Print Network (OSTI)

a tool for reactor design optimization, and for design ofdesign tool for reactor design optimization, and for designdesign tool for reactor design optimization, and for design

Scarlat, Raluca Olga

2012-01-01T23:59:59.000Z

51

Gas turbine cooling system  

DOE Patents (OSTI)

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).

Bancalari, Eduardo E. (Orlando, FL)

2001-01-01T23:59:59.000Z

52

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

E-Print Network (OSTI)

HTTR High Temperature engineering Test Reactor INET Institute of Nuclear Energy Technology LWR Light Water Reactor OKBM Test Design Bureau for Machine Building ORNL Oak Ridge National Laboratory RCCS Reactor Cavity Cooling System... to be at right angles to each other, ignoring an angular distribution of radiant heat.7 MORECA, used by ORNL, simulates accident scenarios for certain gas-cooled reactor types.7 INET conducts their analysis using Thermix, which performs two...

Moore, Eugene James Thomas

2006-08-16T23:59:59.000Z

53

Lessons Learned From Gen I Carbon Dioxide Cooled Reactors  

SciTech Connect

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.

David E. Shropshire

2004-04-01T23:59:59.000Z

54

Hydronic Radiant Cooling Systems  

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

4 4 Hydronic Radiant Cooling Systems Cooling nonresidential buildings in the U.S. contributes significantly to electrical power consumption and peak power demand. Part of the electrical energy used to cool buildings is drawn by fans transporting cool air through the ducts. The typical thermal cooling peak load component for California office buildings can be divided as follows: 31% for lighting, 13% for people, 14% for air transport, and 6% for equipment (in the graph below, these account for 62.5% of the electrical peak load, labeled "chiller"). Approximately 37% of the electrical peak power is required for air transport, and the remainder is necessary to operate the compressor. DOE-2 simulations for different California climates using the California

55

Design Study of Pb-Bi-Cooled and NaK-Cooled Small Reactors: PBWFR and DSFR  

SciTech Connect

The liquid lead-bismuth eutectic (Pb-Bi) has good compatibility with water, which is different from sodium. It is expected that the Pb-Bi could be used as a coolant of the deep sea fast reactor (DSFR) and the Pb-Bi- cooled direct contact boiling water small fast reactor (PBWFR). Physics analysis of the Pb-Bi-cooled small reactor cores with and without inner control rods was performed using the computer program of General Purpose Neutronics Code System (SRAC95) developed by Japan Atomic Energy Research Institute (JAERI). The coolant of Pb-Bi seems to be good as well as NaK for small reactors. (authors)

Otsubo, Akira; Takahashi, Minoru [N1-18, Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

2004-07-01T23:59:59.000Z

56

Heating and cooling system  

SciTech Connect

Heating and cooling of dwelling houses and other confined spaces is facilitated by a system in which thermal energy is transported between an air heating and cooling system in the dwelling and a water heat storage sink or source, preferably in the form of a swimming pool or swimming pool and spa combination. Special reversing valve circuitry and the use of solar collectors and liquid-to-liquid heat exchangers on the liquid side of the system , and special air valves and air modules on the air side of the system, enhance the system's efficiency and make it practical in the sense that systems employing the invention can utilize existing craft skills and building financing arrangements and building codes, and the like, without major modification.

Krumhansl, M.U.

1982-10-12T23:59:59.000Z

57

Review of High Temperature Water and Steam Cooled Reactor Concepts  

SciTech Connect

This review summarizes design concepts of supercritical-pressure water cooled reactors (SCR), nuclear superheaters and steam cooled fast reactors from 1950's to the present time. It includes water moderated supercritical steam cooled reactor, SCOTT-R and SC-PWR of Westinghouse, heavy water moderated light water cooled SCR of GE, SCLWR and SCFR of the University of Tokyo, B-500SKDI of Kurchatov Institute, CANDU -X of AECL, nuclear superheaters of GE, subcritical-pressure steam cooled FBR of KFK and B and W, Supercritical-pressure steam cooled FBR of B and W, subcritical-pressure steam cooled high converter by Edlund and Schultz and subcritical-pressure water-steam cooled FBR by Alekseev. This paper is prepared based on the previous review of SCR2000 symposium, and some author's comments are added. (author)

Oka, Yoshiaki [Nuclear Engineering Research Laboratory, The University of Tokyo, 3-1, Hongo 7-Chome, Bunkyo-ku (Japan)

2002-07-01T23:59:59.000Z

58

Testability of a heat pipe cooled thermionic reactor  

Science Journals Connector (OSTI)

As part of the Air Force Phillips Laboratory thermionics program Rocketdyne performed a design study for an in?core thermionic fuel element (TFE) heat pipe cooled reactor power system. This effort involved a testability evaluation that was performed starting with testing of individual components followed by testing at various stages of fabrication and concluding with full system acceptance and qualification testing. It was determined that the system could be thoroughly tested to ensure a high probability of successful operation in space after launch.

Richard E. Durand

1992-01-01T23:59:59.000Z

59

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

E-Print Network (OSTI)

Safety. The Accident at TEPCO’s Fukushima Nuclear Power2: Accident and Thermal Fluids Analysis PIRTs. (Nuclearmolten nuclear reactor core debris following accidents such

Scarlat, Raluca Olga

2012-01-01T23:59:59.000Z

60

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

E-Print Network (OSTI)

............................................................................................... 24 8 Flow diagram for an Advanced CANDU reactor ..................................... 27 9 Implementation of safeguards measures at a CANDU facility using video surveillance and radiation monitors... ................................................ 28 10 Implementation of safeguards measures at a CANDU facility using core discharge monitor.............................................................................. 28 11 Primary safeguards measures at MONJU Fast Reactor in Japan...

Gitau, Ernest Travis Ngure

2012-10-19T23:59:59.000Z

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


61

Non-linear Dynamical Reliability Analysis in the Very High Temperature Gas Cooled Reactor  

Science Journals Connector (OSTI)

A dynamic safety assessment has been developed for the passive system in the very high temperature gas cooled reactor (VHTR), where the operational data are deficient. It is needed to make use of the character...

Taeho Woo

2012-01-01T23:59:59.000Z

62

Flexible Conversion Ratio Fast Reactor Systems Evaluation  

SciTech Connect

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.

Neil Todreas; Pavel Hejzlar

2008-06-30T23:59:59.000Z

63

Fusion reactor systems  

Science Journals Connector (OSTI)

In this review we consider deuterium-tritium (D-T) fusion reactors based on four different plasma-confinement and heating approaches: the tokamak, the theta-pinch, the magnetic-mirror, and the laser-pellet system. We begin with a discussion of the dynamics of reacting plasmas and basic considerations of reactor power balance. The essential plasma physical aspects of each system are summarized, and the main characteristics of the corresponding conceptual power plants are described. In tokamak reactors the plasma densities are about 1020 m-3, and the ? values (ratio of plasma pressure to confining magnetic pressure) are approximately 5%. Plasma burning times are of the order of 100-1000 sec. Large superconducting dc magnets furnish the toroidal magnetic field, and 2-m thick blankets and shields prevent heat deposition in the superconductor. Radially diffusing plasma is diverted away from the first wall by means of null singularities in the poloidal (or transverse) component of the confining magnetic field. The toroidal theta-pinch reactor has a much smaller minor diameter and a much larger major diameter, and operates on a 10-sec cycle with 0.1-sec burning pulses. It utilizes shock heating from high-voltage sources and adabatic-compression heating powered by low-voltage, pulsed cryogenic magnetic or inertial energy stores, outside the reactor core. The plasma has a density of about 1022 m-3 and ? values of nearly unity. In the power balance of the reactor, direct-conversion energy obtained by expansion of the burning high-? plasma against the containing magnetic field is an important factor. No divertor is necessary since neutral-gas flow cools and replaces the "spent" plasma between pulses. The open-ended mirror reactor uses both thermal conversion of neutron energy and direct conversion of end-loss plasma energy to dc electrical power. A fraction of this direct-convertor power is then fed back to the ioninjection system to sustain the reaction and maintain the plasma. The average ion energy is 600 keV, plasma diameter 6 m, and the plasma beta 85%. The power levels of the three magnetic-confinement devices are in the 500-2000 MWe range, with the exception of the mirror reactor, for which the output is approximately 200 MWe. In Laser-Pellet reactors, frozen D-T pellets are ignited in a cavity which absorbs the electromagnetic, charged particle, and neutron energy from the fusion reaction. The confinement is "inertial," since the fusion reaction occurs during the disassembly of the heated pellet. A pellet-cavity unit would produce about 200 MWt in pulses with a repetition rate of the order of 10 sec-1. Such units could be clustered to give power plants with outputs in the range of 1000 MWe.

F. L. Ribe

1975-01-01T23:59:59.000Z

64

Overview of environmental control aspects for the gas-cooled fast reactor  

SciTech Connect

Environmental control aspects relating to release of radionuclides have been analyzed for the Gas-Cooled Fast Reactor (GCFR). Information on environmental control systems was obtained for the most recent GCFR designs, and was used to evaluate the adequacy of these systems. The GCFR has been designed by the General Atomic Company as an alternative to other fast breeder reactor designs, such as the Liquid Metal Fast Breeder Reactor (LMFBR). The GCFR design includes mixed oxide fuel and helium coolant. The environmental impact of expected radionuclide releases from normal operation of the GCFR was evaluated using estimated collective dose equivalent commitments resulting from 1 year of plant operation. The results were compared to equivalent estimates for the Light Water Reactor (LWR) and High-Temperature Gas-Cooled Reactor (HTGR). A discussion of uncertainties in system performances, tritium production rates, and radiation quality factors for tritium is included.

Nolan, A.M.

1981-05-01T23:59:59.000Z

65

A Semi-Passive Containment Cooling System Conceptual Design  

E-Print Network (OSTI)

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 ...

Liu, H.

66

Improved vortex reactor system  

DOE Patents (OSTI)

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.

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

1995-01-01T23:59:59.000Z

67

General features of direct-cycle, supercritical-pressure, light-water-cooled reactors  

SciTech Connect

The concept of direct-cycle, supercritical-pressure, light-water-cooled reactors is developed. Breeding is possible in the tight lattice core. The power output can be maximized in the fast converter reactor. The gross thermal efficiency of the high temperature reactor adopting Inconel as fuel cladding is expected to be 44.8%. The plant system is similar to the supercritical-fossil-fired power plant which adopts once-through type coolant circulation system. The volume and height of the containment are approximately half of the BWR. The basic safety principles follows those of LWRs. The reactor will solve the economic problems of LWR and LMFBR.

Oka, Y.; Koshizuka, S. [Univ. of Tokyo (Japan). Nuclear Engineering Research Lab.

1996-07-01T23:59:59.000Z

68

Emergency cooling system and method  

DOE Patents (OSTI)

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.

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

1994-01-04T23:59:59.000Z

69

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

SciTech Connect

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.

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

2011-05-01T23:59:59.000Z

70

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

DOE Patents (OSTI)

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.

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

1995-01-01T23:59:59.000Z

71

Desiccant Cooling Systems - A Review  

E-Print Network (OSTI)

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

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

1986-01-01T23:59:59.000Z

72

Reactor vessel support system  

DOE Patents (OSTI)

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.

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

1982-01-01T23:59:59.000Z

73

Temperature initiated passive cooling system  

DOE Patents (OSTI)

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.

Forsberg, C.W.

1994-11-01T23:59:59.000Z

74

Combustor liner cooling system  

DOE Patents (OSTI)

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.

Lacy, Benjamin Paul; Berkman, Mert Enis

2013-08-06T23:59:59.000Z

75

E-Print Network 3.0 - advanced water-cooled reactors Sample Search...  

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

in the Subcritical, Gas-Cooled Fast Transmutation Reactor Concept", Nucl... Cooled, Fast, Subcritical Advanced Burner ... Source: MIT Plasma Science and Fusion Center Collection:...

76

Heat pipe cooled reactors for multi-kilowatt space power supplies  

SciTech Connect

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.

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

1995-01-01T23:59:59.000Z

77

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

SciTech Connect

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.

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

2007-09-25T23:59:59.000Z

78

Small LBE-Cooled Fast Reactor for Expanding Market  

SciTech Connect

A long-life safe simple small portable proliferation-resistant reactor is expected to solve many problems associating future energy and globally environmental problems. From discussions on mainly neutronics and safety points it has been shown that the heavy liquid metal cooled fast reactor is the best candidate to satisfy the above requirements. A lead-bismuth-eutectic (LBE) cooled fast reactor LSPR (LBE-Cooled Long-Life Safe Simple Small Portable Proliferation-Resistant Reactor) has been designed and continues to be improved. In the present paper a recent version of LSPR is presented. The total power of the present design is 150 MWt (53 MWe). During whole reactor life of 12 years the excess reactivity required for burnup is very low, and negative coolant dilatation coefficient is confirmed. This characteristic together with some other characteristics makes unprotected loss of flow (ULOF) accident inherently safe. It can survive even simultaneous rod run-out transient over power (UTOP), ULOF and unprotected loss of heat sink (ULOHS) accident without the help of an operator or active device. (authors)

Hiroshi Sekimoto [Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Shinichi Makino [Nissan Motor Company Ltd. (Japan); Kunihiko Nakamura; Yoshio Kamishima [Advanced Reactor Technology Company Ltd. (Japan); Takashi Kawakita [Mitsubishi Heavy Industries Ltd. (Japan)

2002-07-01T23:59:59.000Z

79

Improved vortex reactor system  

DOE Patents (OSTI)

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.

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

1995-05-09T23:59:59.000Z

80

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

SciTech Connect

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

Moisseytsev, A.; Sienicki, J.J.

2004-10-06T23:59:59.000Z

Note: This page contains sample records for the topic "reactor cooling system" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Reactor & Nuclear Systems Publications | ORNL  

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

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

82

Three-dimensional imaging and precision metrology for liquid-salt-cooled reactors  

SciTech Connect

The liquid-salt-cooled very high temperature reactor, also called the Advanced High-Temperature Reactor (AHTR), is a new large high-temperature reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. The AHTR will require refueling, in-service inspection, and maintenance (RIM) with supporting instrumentation systems. The fluoride salts that are being evaluated as potential reactor coolants have melting points between 350 and 500 deg. C, values that imply minimum RIM temperatures between 400 and 550 deg. C. These salts are transparent over a wider range of the light spectrum than is water. The high temperatures, the optical characteristics of the coolant, and advances in metrology may enable the use of lasers to create three-dimensional images of the reactor interior to assist refueling, monitor vibrations in components, map fluid flow, and enable inspections of internal reactor components. A description of the reactor and an initial evaluation of the use of optical techniques for AHTR instrumentation are provided. (authors)

Forsberg, C. W. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6165 (United States); Varma, V. K.; Burgess, T. W. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6304 (United States)

2006-07-01T23:59:59.000Z

83

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

SciTech Connect

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)

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

84

High temperature gas cooled reactor steam-methane reformer design  

SciTech Connect

The concept of the long distance transportation of process heat energy from a High Temperature Gas Cooled Reactor (HTGR) heat source, based on the steam-methane reforming reaction, is being evaluated by the Department of Energy as an energy source/application for use early in the 21st century. This paper summaries the design of a helium heated steam reformer utilized in conjunction with an intermediate loop, 850/degree/C reactor outlet temperature, HTGR process heat plant concept. This paper also discusses various design considerations leading to the mechanical design features, the thermochemical performance, the materials selection and the structural design analysis. 12 refs.

Impellezzeri, J.R.; Drendel, D.B.; Odegaard, T.K.

1981-01-01T23:59:59.000Z

85

Solar-powered cooling system  

DOE Patents (OSTI)

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.

Farmer, Joseph C

2013-12-24T23:59:59.000Z

86

Compressor bleed cooling fluid feed system  

DOE Patents (OSTI)

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.

Donahoo, Eric E; Ross, Christopher W

2014-11-25T23:59:59.000Z

87

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

SciTech Connect

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.

Peters, Curtis D. (.)

2006-04-01T23:59:59.000Z

88

Evaluation of Hastelloy X for gas-cooled-reactor applications  

SciTech Connect

Hastelloy X is a potential structural material for use in gas-cooled reactor systems. In this application data are necessary on the mechanical properties of base metals and weldments under realistic service conditions. The test environment studied was helium that contained small amounts of H/sub 2/, CH/sub 4/, and CO. It is shown that this environment is carburizing with the kinetics of this process, becoming rapid above 800/sup 0/C. Suitable weldments of Hastelloy X were prepared by several processes; those weldments generally had properties similar to the base metal except for lower fracture strains under some conditions. Some samples were aged up to 20,000 h in the test gas and tested, and some creep tests on as-received material exceeded 40,000 h. The predominant effect of aging was the significant reduction of the fracture strains at ambient temperature; the strains were lower when the samples were aged in HTGR helium than when aged in inert gas. Under some conditions aging also increased the yield and ultimate tensile strength. Limited impact testing showed that the impact energy at 25/sup 0/C was reduced drastically by aging at 871 and 704/sup 0/C.

McCoy, H.E.; King, J.F.

1982-11-01T23:59:59.000Z

89

Fluoride Salt-Cooled High-Temperature Reactor Development Roadmap  

SciTech Connect

Fluoride salt-cooled high-temperature reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics and fully passive safety. This paper provides an overview of a technology development pathway for expeditious commercial deployment of first-generation FHRs. The paper describes the principal remaining FHR technology challenges and the development path needed to address the challenges. First-generation FHRs do not appear to require any technology breakthroughs, but will require significant technology development and demonstration. FHRs are currently entering early phase engineering development. As such, the development 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.

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

2014-01-01T23:59:59.000Z

90

Lamination cooling system formation method  

DOE Patents (OSTI)

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.

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

2012-06-19T23:59:59.000Z

91

Lamination cooling system formation method  

DOE Patents (OSTI)

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.

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

2009-05-12T23:59:59.000Z

92

Cooling System Basics | Department of Energy  

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

Cooling System Basics Cooling System Basics Cooling System Basics August 16, 2013 - 1:08pm Addthis Cooling technologies used in homes and buildings include ventilation, evaporative cooling, air conditioning, absorption cooling, and radiant cooling. Learn more about how these technologies work. Ventilation Ventilation allows air to move into and out of homes and buildings either by natural or mechanical means. Evaporative Cooling In dry climates, evaporative cooling or "swamp cooling" provides an experience like air conditioning, but with much lower energy use. An evaporative cooler uses the outside air's heat to evaporate water inside the cooler. The heat is drawn out of the air and the cooled air is blown into the space by the cooler's fan. Air Conditioning Air conditioners, which employ the same operating principles and basic

93

Cooling System Basics | Department of Energy  

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

Cooling System Basics Cooling System Basics Cooling System Basics August 16, 2013 - 1:08pm Addthis Cooling technologies used in homes and buildings include ventilation, evaporative cooling, air conditioning, absorption cooling, and radiant cooling. Learn more about how these technologies work. Ventilation Ventilation allows air to move into and out of homes and buildings either by natural or mechanical means. Evaporative Cooling In dry climates, evaporative cooling or "swamp cooling" provides an experience like air conditioning, but with much lower energy use. An evaporative cooler uses the outside air's heat to evaporate water inside the cooler. The heat is drawn out of the air and the cooled air is blown into the space by the cooler's fan. Air Conditioning Air conditioners, which employ the same operating principles and basic

94

Gas hydrate cool storage system  

DOE Patents (OSTI)

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)

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

1984-09-12T23:59:59.000Z

95

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

SciTech Connect

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)

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

96

Key Thermal Fluid Phenomena In Prismatic Gas-Cooled Reactors  

SciTech Connect

Several types of gas-cooled nuclear reactors have been suggested as part of the international Generation IV initiative with the proposed NGNP (Next Generation Nuclear Plant) as one of the main concepts [MacDonald et al., 2003]. Meaningful studies for these designs will require accurate, reliable predictions of material temperatures to evaluate the material capabilities; these temperatures depend on the thermal convection in the core and in other important components. Some of these reactors feature complex geometries and wide ranges of temperatures, leading to significant variations of the gas thermodynamic and transport properties plus possible effects of buoyancy during normal and reduced power operations and loss-of-flow (LOFA) and loss-of-coolant scenarios. Potential issues identified to date include ''hot streaking'' in the lower plenum evolving from ''hot channels'' in prismatic cores. In order to predict thermal hydraulic behavior of proposed designs effectively and efficiently, it is useful to identify the dominant phenomena occurring.

D. M. McEligot; G. E. McCreery; P. D. Bayless; T. D. Marshall

2005-06-01T23:59:59.000Z

97

Information technology equipment cooling system  

SciTech Connect

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.

Schultz, Mark D.

2014-06-10T23:59:59.000Z

98

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

E-Print Network (OSTI)

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 ...

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

2007-01-01T23:59:59.000Z

99

Cooling system for superconducting magnet  

DOE Patents (OSTI)

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.

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

1998-12-15T23:59:59.000Z

100

E-Print Network 3.0 - army gas-cooled reactor Sample Search Results  

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

ENABLING SUSTAINABLE NUCLEAR POWER Summary: and NRE Design Class., "Advances in the Subcritical, Gas-Cooled Fast Transmutation Reactor Concept", Nucl... . Tedder, J. Lackey, J....

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


101

Measurements of thermal-hydraulic parameters in liquid-metal-cooled fast-breeder reactors  

SciTech Connect

This paper discusses instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, sodium purity, and leakage. The paper identifies the overall instrumentation requirements for LMFBR's and those aspects of instrumentation which are unique or of special concern to LMFBR systems. It also gives an overview of the status of instrument design and performance.

Sackett, J.I.

1983-01-01T23:59:59.000Z

102

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

SciTech Connect

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.

Lee O. Nelson

2011-04-01T23:59:59.000Z

103

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

SciTech Connect

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.

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

104

The integration of cryogenic cooling systems with superconducting electronic systems  

E-Print Network (OSTI)

SCMAG-SIO The Integration of Cryogenic Cooling Systems With76SF0009S. The Integration of Cryogenic Cooling Systems WithAbstract- The need for cryogenic cooling has been critical

Green, Michael A.

2011-01-01T23:59:59.000Z

105

Helium circulator design considerations for modular high temperature gas-cooled reactor plant  

SciTech Connect

Efforts are in progress to develop a standard modular high temperature gas-cooled reactor (MHTGR) plant that is amenable to design certification and serial production. The MHTGR reference design, based on a steam cycle power conversion system, utilizes a 350 MW(t) annular reactor core with prismatic fuel elements. Flexibility in power rating is afforded by utilizing a multiplicity of the standard module. The circulator, which is an electric motor-driven helium compressor, is a key component in the primary system of the nuclear plant, since it facilitates thermal energy transfer from the reactor core to the steam generator; and, hence, to the external turbo-generator set. This paper highlights the helium circulator design considerations for the reference MHTGR plant and includes a discussion on the major features of the turbomachine concept, operational characteristics, and the technology base that exists in the U.S.

McDonald, C.F.; Nichols, M.K.

1987-01-01T23:59:59.000Z

106

Helium circulator design considerations for modular high temperature gas-cooled reactor plant  

SciTech Connect

Efforts are in progress to develop a standard modular high temperature gas-cooled reactor (MHTGR) plant that is amenable to design certification and serial production. The MHTGR reference design, based on a steam cycle power conversion system, utilizes a 350 MW(t) annular reactor core with prismatic fuel elements. Flexibility in power rating is afforded by utilizing a multiplicity of the standard module. The circulator, which is an electric motor-driven helium compressor, is a key component in the primary system of the nuclear plant, since it facilitates thermal energy transfer from the reactor core to the steam generator; and, hence, to the external turbo-generator set. This paper highlights the helium circulator design considerations for the reference MHTGR plant and includes a discussion on the major features of the turbomachine concept, operational characteristics, and the technology base that exists in the US.

McDonald, C.F.; Nichols, M.K.

1986-12-01T23:59:59.000Z

107

Parametric Investigation of Brayton Cycle for High Temperature Gas-Cooled Reactor  

SciTech Connect

The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. In this project, we are investigating helium Brayton cycles for the secondary side of an indirect energy conversion system. Ultimately we will investigate the improvement of the Brayton cycle using other fluids, such as supercritical carbon dioxide. Prior to the cycle improvement study, we established a number of baseline cases for the helium indirect Brayton cycle. These cases look at both single-shaft and multiple-shaft turbomachinary. The baseline cases are based on a 250 MW thermal pebble bed HTGR. The results from this study are applicable to other reactor concepts such as a very high temperature gas-cooled reactor (VHTR), fast gas-cooled reactor (FGR), supercritical water reactor (SWR), and others. In this study, we are using the HYSYS computer code for optimization of the helium Brayton cycle. Besides the HYSYS process optimization, we performed parametric study to see the effect of important parameters on the cycle efficiency. For these parametric calculations, we use a cycle efficiency model that was developed based on the Visual Basic computer language. As a part of this study we are currently investigated single-shaft vs. multiple shaft arrangement for cycle efficiency and comparison, which will be published in the next paper.The ultimate goal of this study is to use supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency to values great than that of the helium Brayton cycle. This paper includes preliminary calculations of the steady state overall Brayton cycle efficiency based on the pebble bed reactor reference design (helium used as the working fluid) and compares those results with an initial calculation of a CO2 Brayton cycle.

Chang Oh

2004-07-01T23:59:59.000Z

108

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

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

BSU GHP District Heating and Cooling System (Phase I) BSU GHP District Heating and Cooling System (Phase I) Project objectives: Create a campus geothermal heating and cooling...

109

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

E-Print Network (OSTI)

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

Harilal, S. S.

110

Sustained Recycle in Light Water and Sodium-Cooled Reactors  

SciTech Connect

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.

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

2010-11-01T23:59:59.000Z

111

ORIGEN-ARP Cross-Section Libraries for Magnox, Advanced Gas-Cooled, and VVER Reactor Designs  

SciTech Connect

Cross-section libraries for the ORIGEN-ARP system were extended to include four non-U.S. reactor types: the Magnox reactor, the Advanced Gas-Cooled Reactor, the VVER-440, and the VVER-1000. Typical design and operational parameters for these four reactor types were determined by an examination of a variety of published information sources. Burnup simulation models of the reactors were then developed using the SAS2H sequence from the Oak Ridge National Laboratory SCALE code system. In turn, these models were used to prepare the burnup-dependent cross-section libraries suitable for use with ORIGEN-ARP. The reactor designs together with the development of the SAS2H models are described, and a small number of validation results using spent-fuel assay data are reported.

Murphy, BD

2004-03-10T23:59:59.000Z

112

A new cylinder cooling system using oil  

SciTech Connect

The design of engine cylinders must satisfy two conflicting requirements, good cooling performance and ease of manufacture. A cooling system was designed to permit the circulation of engine lubricating oil as a coolant at high speed through grooves provided on the external periphery of the cylinder liner. Testing in an actual operating engine confirmed that this cooling system design not only provides better heat transfer and higher cooling performance but also simplifies the manufacturing of the cylinder since external cooling fins are not required. In this paper, the authors will discuss the cylinder cooling effect of the new cylinder cooling system, referring mainly to the test results of a single-cylinder motorcycle engine with lubricating oil from the crankcase used as the coolant.

Harashina, Kenichi; Murata, Katsuhiro; Satoh, Hiroshi; Shimizu, Yasuo; Hamamura, Masahiro

1995-12-31T23:59:59.000Z

113

Cooling load design tool for UFAD systems.  

E-Print Network (OSTI)

Underfloor Air Distribution (UFAD) Design Guide. Atlanta:Load Design Tool for Underfloor Air Distribution Systems. ”for design cooling loads in underfloor air distribution (

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

2010-01-01T23:59:59.000Z

114

Attrition reactor system  

DOE Patents (OSTI)

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.

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

1993-09-28T23:59:59.000Z

115

Reactor vessel support system. [LMFBR  

DOE Patents (OSTI)

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.

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

1980-05-09T23:59:59.000Z

116

Active Solar Heating and Cooling Systems Exemption  

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

Active solar heating and cooling systems may not be assessed at more than the value of a conventional system for property tax purposes. This law applies only to active solar systems and does not...

117

Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis  

SciTech Connect

This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

Ilas, Germina [ORNL; Ilas, Dan [ORNL; Kelly, Ryan P [ORNL; Sunny, Eva E [ORNL

2012-08-01T23:59:59.000Z

118

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

E-Print Network (OSTI)

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 ...

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

119

Hot gas path component cooling system  

DOE Patents (OSTI)

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.

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

2014-02-18T23:59:59.000Z

120

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)  

SciTech Connect

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

Yamada, K. [Vienna International Centre, P.O. Box 100, 1400 Vienna (Austria); Aksan, S. N. [International Atomic Energy Agency, 1400 Vienna (Austria)

2012-07-01T23:59:59.000Z

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


121

Optimization of actinide transmutation in innovative lead-cooled fast reactors  

E-Print Network (OSTI)

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 ...

Romano, Antonino, 1972-

2003-01-01T23:59:59.000Z

122

Containment system for supercritical water oxidation reactor  

DOE Patents (OSTI)

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.

Chastagner, P.

1994-07-05T23:59:59.000Z

123

Ventilation Systems for Cooling | Department of Energy  

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

Ventilation Systems for Cooling Ventilation Systems for Cooling Ventilation Systems for Cooling May 30, 2012 - 6:19pm Addthis Proper ventilation helps you save energy and money. | Photo courtesy of JD Hancock. Proper ventilation helps you save energy and money. | Photo courtesy of JD Hancock. Ventilation is the least expensive and most energy-efficient way to cool buildings. Ventilation works best when combined with methods to avoid heat buildup in your home. In some cases, natural ventilation will suffice for cooling, although it usually needs to be supplemented with spot ventilation, ceiling fans, and window fans. For large homes, homeowners might want to investigate whole house fans. Interior ventilation is ineffective in hot, humid climates where

124

Ventilation Systems for Cooling | Department of Energy  

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

Ventilation Systems for Cooling Ventilation Systems for Cooling Ventilation Systems for Cooling May 30, 2012 - 6:19pm Addthis Proper ventilation helps you save energy and money. | Photo courtesy of JD Hancock. Proper ventilation helps you save energy and money. | Photo courtesy of JD Hancock. Ventilation is the least expensive and most energy-efficient way to cool buildings. Ventilation works best when combined with methods to avoid heat buildup in your home. In some cases, natural ventilation will suffice for cooling, although it usually needs to be supplemented with spot ventilation, ceiling fans, and window fans. For large homes, homeowners might want to investigate whole house fans. Interior ventilation is ineffective in hot, humid climates where

125

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

SciTech Connect

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.

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

2010-11-01T23:59:59.000Z

126

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

E-Print Network (OSTI)

reactor core design 7 Assembly Design and Optimization codePhysics Optimization of Breed and Burn Fast Reactor Systems.OPTIMIZATION CODE (ADOPT) Table 7.4: SWR B&B Reference Reactor

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

127

Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)  

SciTech Connect

The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab.

Silady, F.A.; Millunzi, A.C.

1989-08-01T23:59:59.000Z

128

Experimental and CFD Analysis of Advanced Convective Cooling Systems  

SciTech Connect

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).

Yassin A. Hassan; Victor M. Ugaz

2012-06-27T23:59:59.000Z

129

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

SciTech Connect

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

Not Available

1980-02-01T23:59:59.000Z

130

Advanced Open-Cycle Desiccant Cooling System  

E-Print Network (OSTI)

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...

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

1989-01-01T23:59:59.000Z

131

Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1979-September 30, 1979  

SciTech Connect

The results of work performed from July 1, 1979 through September 30, 1979 on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

Not Available

1980-03-07T23:59:59.000Z

132

Assessment of passive safety system performance under gravity driven cooling system drain line break accident  

Science Journals Connector (OSTI)

Abstract A generation III+ Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features to improve safety and performance. Natural circulation allows the elimination of emergency injection pump and no operator action or alternating current (AC) power supply. The generation III+ BWR's passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS) and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A GDCS Drain Line Break (GDLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III+ BWR. The main results of PUMA GDLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the containment maximum pressure (266 kPa) was 36% lower than the safety limit (414 kPa). The minimum collapsed water level (1.496 m) before the GDCS injection was 8% lower than the TAF (1.623 m) but it was ensured that two-phase water level was higher than the TAF with no core uncovery.

J. Lim; J. Yang; S.W. Choi; D.Y. Lee; S. Rassame; T. Hibiki; M. Ishii

2014-01-01T23:59:59.000Z

133

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

SciTech Connect

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

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

2012-02-01T23:59:59.000Z

134

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

SciTech Connect

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.

Ian Mckirdy

2010-12-01T23:59:59.000Z

135

Direct conversion nuclear reactor space power systems  

SciTech Connect

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.

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

1982-08-01T23:59:59.000Z

136

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

SciTech Connect

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.

Jeffrey D Bryan

2009-09-01T23:59:59.000Z

137

Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies  

SciTech Connect

A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

Dixon, David 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); Hiatt, Matthew T. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, Texas A and M University, College Station, TX 77843 (United States); Poston, David I.; Kapernick, Richard J. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2006-01-20T23:59:59.000Z

138

Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980  

SciTech Connect

Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Included are the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.

Not Available

1980-06-25T23:59:59.000Z

139

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

E-Print Network (OSTI)

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 ...

Pope, Michael A. (Michael Alexander)

2006-01-01T23:59:59.000Z

140

Process Cooling Pumping Systems Analysis  

E-Print Network (OSTI)

their calculated design conditions. As observed, each system was running +1 pump more than required to meet the requirement. System 1 was overpumped by 25%, or 1100 GPM, and System 2 was overpumped by 37%, or 1850 GPM. Both systems may provide adequate heat...

Sherman, C.

2008-01-01T23:59:59.000Z

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


141

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

E-Print Network (OSTI)

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 ...

Minck, Matthew J. (Matthew Joseph)

2013-01-01T23:59:59.000Z

142

HIGH TEMPERATURE GAS-COOLED REACTOR KNOWLEDGE MANAGEMENT  

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

to assure the safety of the public. Such a request would mean that the reactor would go subcritical upon scram of the outer reflector control rods, and then, after 36 hours for...

143

Passive Cooling System for a Vehicle  

DOE Patents (OSTI)

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).

Hendricks, T. J.; Thoensen, T.

2005-11-15T23:59:59.000Z

144

Studies on Materials for Heavy-Liquid-Metal-Cooled Reactors in Japan  

SciTech Connect

Recent studies on materials for the development of lead-bismuth (Pb-Bi)-cooled fast reactors (FR) and accelerator-driven sub-critical systems (ADS) in Japan are reported. The measurement of the neutron cross section of Bi to produce {sup 210}Po, the removal experiment of Po contamination and steel corrosion test in Pb-Bi flow were performed in Tokyo Institute of Technology. A target material corrosion test was performed in the project of Transmutation Experimental Facility for ADS in Japan Atomic Energy Research Institute (JAERI). Steel corrosion test was started in Mitsui Engineering and Shipbuilding Co., LTD (MES). The feasibility study for FR cycle performed in Japan Nuclear Cycle Institute (JNC) are described. (authors)

Minoru Takahashi; Masayuki Igashira; Toru Obara; Hiroshi Sekimoto [Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Kenji Kikuchi [Japan Atomic Energy Research Institute (Japan); Kazumi Aoto [Japan Nuclear Cycle Development Institute (Japan); Teruaki Kitano [Mitsui Engineering and Shipbuilding Company Ltd., 6-4, Tsukiji 5-chome, Chuo-ku, Tokyo 104-8439 (Japan)

2002-07-01T23:59:59.000Z

145

Thermal storage for solar cooling using paired ammoniated salt reactors. Final report  

SciTech Connect

The objectives of the program were to investigate the feasibility of using various solid and liquid ammoniates in heat pump/thermal storage systems for space heating and cooling. The study included corrosion testing of selected metallic and non-metallic specimens in the ammoniates, subscale testing of the candidate ammoniates singly and in pairs, trade studies and conceptual design of a residential system, prototype testing, and ammoniation/deammoniation cyclic testing of manganese chloride. Results of the corrosion testing showed that problems exist with manganese and magnesium chloride ammoniates, except with the teflon which displayed excellent resistance in all environments. Also, all liquid ammoniates are unsuitable for use with uncoated carbon steel. Cycling of the manganese chloride between the high and low ammoniates does not affect its properties. However, the density change between the high and low ammoniates could cause packing problems in a reactor which constrains the salt volume. Subscale tests with solid ammoniates indicated that the heat transfer coefficient in a fixed bed reactor is low (approx. 1 Btu/h-ft/sup 2/-/sup 0/F). Therefore solid ammoniates are not practical because of the high heat exchanger cost requirement. Forced ammonia recirculation was tested as a means of increasing heat transfer rate in the fixed bed reactor with solid salts, but was not successful. Conversely, the subscale testing with liquid ammoniates produced heat transfer coefficients of 40 to 45 Btu/h-ft/sup 2/-/sup 0/F. Thus, the residential design was based on a liquid ammoniate/ammonia system using ammonium nitrate as the salt.

Not Available

1981-09-01T23:59:59.000Z

146

Rapid starting methanol reactor system  

DOE Patents (OSTI)

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.

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

147

New and Underutilized Technology: Evaporative Pre-Cooling Systems |  

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

Evaporative Pre-Cooling Systems Evaporative Pre-Cooling Systems New and Underutilized Technology: Evaporative Pre-Cooling Systems October 4, 2013 - 4:43pm Addthis The following information outlines key deployment considerations for evaporated pre-cooling systems within the Federal sector. Benefits Evaporative pre-cooling systems install ahead of the condenser to lower the condenser pressure. These systems can also work with an economizer. Evaporative pre-cooling reduces the requirement for energy intensive DX cooling. Application Evaporative pre-cooling systems are applicable in most building categories. Climate and Regional Considerations Evaporative pre-cooling systems are well suited in dry climates. Key Factors for Deployment Water usage needs to be taken into account in evaporative pre-cooling

148

New and Underutilized Technology: Evaporative Pre-Cooling Systems |  

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

Technology: Evaporative Pre-Cooling Systems Technology: Evaporative Pre-Cooling Systems New and Underutilized Technology: Evaporative Pre-Cooling Systems October 4, 2013 - 4:43pm Addthis The following information outlines key deployment considerations for evaporated pre-cooling systems within the Federal sector. Benefits Evaporative pre-cooling systems install ahead of the condenser to lower the condenser pressure. These systems can also work with an economizer. Evaporative pre-cooling reduces the requirement for energy intensive DX cooling. Application Evaporative pre-cooling systems are applicable in most building categories. Climate and Regional Considerations Evaporative pre-cooling systems are well suited in dry climates. Key Factors for Deployment Water usage needs to be taken into account in evaporative pre-cooling

149

ASSESSING POWER PLANT COOLING WATER INTAKE SYSTEM  

E-Print Network (OSTI)

ASSESSING POWER PLANT COOLING WATER INTAKE SYSTEM ENTRAINMENT IMPACTS Prepared For: California, Center for Ocean Health, Long Marine Lab GREGOR CAILLIET, Moss Landing Marine Laboratories DAVID MAYER be obvious that large studies like these require the coordinated work of many people. We would first like

150

Hastelloy-X for high-temperature gas-cooled reactor applications  

SciTech Connect

Hastelloy-X is a potential structural material for use in gas-cooled reactor systems. In this application, data are necessary on the mechanical properties of base metal and weldments under realistic service conditions. The test environment studied was helium that contained small amounts of H/sub 2/, CH/sub 4/, and CO. This environment was found to be carburizing, with the kinetics of this process becoming rapid above 800/sup 0/C. Suitable weldments of Hastelloy-X were prepared by several processes; those weldments generally had the same properties as base metal except for lower fracture strains under some conditions. Some samples were aged for up to 20 000 h in the test gas and tested, and some creep tests on as-received material exceeded 40 000 h. The predominant effects of aging were the significant reduction in the fracture strains at ambient temperature and the lower strains for samples aged in high-temperature gas-cooled reactor (HTGR) helium than for those aged in inert gas. Under some conditions, aging also resulted in increased yield and ultimate tensile strength. Creep tests failed to show the effects of environment, aging, or welding on the creep strength of Hastelloy-X; however, the fracture strains for weldments were generally lower than they were for base metal. Prior aging in inert gas for 20 000 h at 538 and 871/sup 0/C reduced the fatigue life slightly, but no difference was observed in the fatigue properties of samples aged in air and HTGR helium environments.

McCoy, H.E.; King, J.F.; Strizak, J.P.

1984-07-01T23:59:59.000Z

151

Hastelloy-X for high-temperature gas-cooled reactor applications  

SciTech Connect

Hastelloy-X is a potential structural material for use in gas-cooled reactor systems. In this application, data are necessary on the mechanical properties of base metal and weldments under realistic service conditions. The test environment studied was helium that contained small amounts of H/sub 2/, CH/sub 4/, and CO. This environment was found to be carburizing, with the kinetics of this process becoming rapid above 800/sup 0/C. Suitable weldments of Hastelloy-X were prepared by several processes; those weldments generally had the same properties as base metal except for lower fracture strains under some conditions. Some samples were aged for up to 20000 h in the test gas and tested, and some creep tests on as-received material exceeded 40000 h. The predominant effects of aging were the significant reduction in the fracture strains at ambient temperature and the lower strains for samples aged in high-temperature gas-cooled reactor (HTGR) helium than for those aged in inert gas. Under some conditions, aging also resulted in increased yield and ultimate tensile strength. Creep tests failed to show the effects of environment, aging, or welding on the creep strength of Hastelloy-X; however, the fracture strains for weldments were generally lower than they were for base metal. Prior aging in inert gas for 20000 h at 538 and 871/sup 0/C reduced the fatigue life slightly, but no difference was observed in the fatigue properties of samples aged in air and HTGR helium environments.

McCoy, H.E.; King, J.F.; Strizak, J.P.

1984-07-01T23:59:59.000Z

152

Cooling system for internal combustion engines  

SciTech Connect

A cooling system for an internal combustion engine is described comprising: a head-side water jacket and a block-side water jacket made independent of each other; and a radiator and a cooling fan shared between the two water jackets. The improvement comprises: a first cooling water conduit for connecting the outlet of the head-side water jacket and the inlet of the radiator; a mixing valve having two water inlets and one water outlet; a second cooling water conduit for connecting one of the water inlets of the mixing valve and the outlet of the radiator; a third conduit for connecting the water outlet of the block-side water jacket and the remaining one of the water inlets of the mixing valve; a water pump, a fourth conduit branched midway from the second conduit and connected with the water inlet of the head-side water jacket; an auxiliary water pump; a fifth conduit branched midway from the third conduit and connected with the first conduit; one-way valve; and a control unit for controlling the mixing ratio of the mixing valve, the displacement of the auxiliary water pump and the operation of the cooling fan.

Itakura, M.

1988-07-26T23:59:59.000Z

153

Method of fabricating a cooled electronic system  

DOE Patents (OSTI)

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.

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

154

Thermostatically controlled solar heating and cooling system  

SciTech Connect

This patent describes a solar heating and cooling system for simultaneously heating or cooling an ambient air system within a building, heating a hot water supply for domestic use within the building and heating or cooling a swimming pool adjacent the building comprising a building. This comprises a swimming pool as a primary water source, a solar connector connected to the swimming pool, a heat pump for controlling ambient air temperature within the building, an energy conservation unit connected to the heat pump and to the hot water supply for utilizing hot gases from the heat pump to heat water in the hot water supply and an air heat exchanger connected to the air system and to the heat pump for selectively heating or cooling air in the building. Also a water heat exchanger is connected to a water source for selectively transferring heat between the heat pump and the water source, a well as a secondary water source connected to the water heat exchanger.

Yovanofski, T.

1986-12-16T23:59:59.000Z

155

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

SciTech Connect

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

Vondy, D.R.

1984-07-01T23:59:59.000Z

156

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

SciTech Connect

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.

Lee Nelson

2011-09-01T23:59:59.000Z

157

Preliminary neutronics design of china lead-alloy cooled demonstration reactor (CLEAR-III) for nuclear waste transmutation  

SciTech Connect

China Lead-Alloy cooled Demonstration Reactor (CLEAR-III), which is the concept of lead-bismuth cooled accelerator driven sub-critical reactor for nuclear waste transmutation, was proposed and designed by FDS team in China. In this study, preliminary neutronics design studies have primarily focused on three important performance parameters including Transmutation Support Ratio (TSR), effective multiplication factor and blanket thermal power. The constraint parameters, such as power peaking factor and initial TRU loading, were also considered. In the specific design, uranium-free metallic dispersion fuel of (TRU-Zr)-Zr was used as one of the CLEAR-III fuel types and the ratio between MA and Pu was adjusted to maximize transmutation ratio. In addition, three different fuel zones differing in the TRU fraction of the fuel were respectively employed for this subcritical reactor, and the zone sizes and TRU fractions were determined such that the linear powers of these zones were close to each other. The neutronics calculations and analyses were performed by using Multi-Functional 4D Neutronics Simulation System named VisualBUS and nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library). In the preliminary design, the maximum TSRLLMA was {approx}11 and the blanket thermal power was {approx}1000 MW when the effective multiplication factor was 0.98. The results showed that good performance of transmutation could be achieved based on the subcritical reactor loaded with uranium-free fuel. (authors)

Chen, Z. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Southwest Science and Technology Univ., No.350 Shushanhu Road, Shushan District, Hefei, Anhui, 230031 (China); Chen, Y.; Bai, Y.; Wang, W.; Chen, Z.; Hu, L.; Long, P. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, Univ. of Science and Technology of China, Hefei, Anhui, 230031 (China)

2012-07-01T23:59:59.000Z

158

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

SciTech Connect

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)

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

2006-07-01T23:59:59.000Z

159

Hybrid optomechanical cooling by atomic $?$ systems  

E-Print Network (OSTI)

We investigate a hybrid quantum system consisting of a cavity optomechanical device optically coupled to an ultracold quantum gas. We show that the dispersive properties of the ultracold gas can be used to dramatically modify the optomechanical response of the mechanical resonator. We examine hybrid schemes wherein the mechanical resonator is coupled either to the motional or the spin degrees of freedom of the ultracold gas. In either case, we find an enhancement of more than two orders of magnitude in optomechanical cooling due to this hybrid interaction. Significantly, based on demonstrated parameters for the cavity optomechanical device, we identify regimes that enable the ground state cooling of the resonator from room temperature. In addition, the hybrid system considered here represents a powerful interface for the use of an ultracold quantum gas for state preparation, sensing and quantum manipulation of a mesoscopic mechanical resonator.

F. Bariani; S. Singh; L. F. Buchmann; M. Vengalattore; P. Meystre

2014-07-03T23:59:59.000Z

160

Low pressure cooling seal system for a gas turbine engine  

DOE Patents (OSTI)

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.

Marra, John J

2014-04-01T23:59:59.000Z

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


161

Reactor vessel annealing system  

DOE Patents (OSTI)

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).

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

162

Reactor and Nuclear Systems Division (RNSD)  

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

RNSD Home RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Staff Details (CV/Bios) Publications Org Chart Contact Us ORNL Staff Only Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Reactor and Nuclear Systems Division News Highlights U.S. Rep. Fleischmann touts ORNL as national energy treasure Martin Peng wins Fusion Power Associates Leadership Award

163

Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor  

SciTech Connect

The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

C. B. Davis

2006-07-01T23:59:59.000Z

164

Chapter 7 - Test Cell Cooling Water and Exhaust Gas Systems  

Science Journals Connector (OSTI)

Part 1 considers the thermodynamics of water cooling systems, water quality, typical cooling water circuits, and engine coolant control units. Also covered are the commissioning cooling circuits, thermal shock, and chilled water systems. Part 2 covers the design of test cell exhaust systems, exhaust silencers, exhaust gas volume flow, exhaust silencers, and exhaust cowls. Part 3 briefly covers the testing of turbochargers.

A.J. Martyr; M.A. Plint

2012-01-01T23:59:59.000Z

165

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

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

166

Modular high temperature gas-cooled reactor plant design duty cycle. Revision 3  

SciTech Connect

This document defines the Plant Design Duty Cycle (PCDC) for the Modular High Temperature Gas-cooled Reactor (MHTGR). The duty cycle is a set of events and their design number of occurrences over the life of the plant for which the MHTGR plant shall be designed to ensure that the plant meets all the top-level requirements. The duty cycle is representative of the types of events to be expected in multiple reactor module-turbine plant configurations of the MHTGR. A synopsis of each PDDC event is presented to provide an overview of the plant response and consequence. 8 refs., 1 fig., 4 tabs.

Chan, T.

1989-12-31T23:59:59.000Z

167

Fuel performance models for high-temperature gas-cooled reactor core design  

SciTech Connect

Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10/sup -4/ fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience.

Stansfield, O.M.; Simon, W.A.; Baxter, A.M.

1983-09-01T23:59:59.000Z

168

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

SciTech Connect

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).

Not Available

1986-10-01T23:59:59.000Z

169

MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents  

SciTech Connect

The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR.

Ball, S.J. (Oak Ridge National Lab., TN (United States))

1991-10-01T23:59:59.000Z

170

Maintenance Guide for Greenhouse Ventilation, Evaporative Cooling Heating Systems1  

E-Print Network (OSTI)

condensation in winter, reduced life and reliability of ventilation equipment, and high repair bills cooling and heating systems. VENTILATION SYSTEMS The operating efficiency of a ventilation fan can be pockets of stagnant air, inadequate cooling from evaporative cooling pads, high heating expenses, heavy

Watson, Craig A.

171

CCHP System with Interconnecting Cooling and Heating Network  

E-Print Network (OSTI)

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...

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

2006-01-01T23:59:59.000Z

172

The Thermodynamic and Cost Benefits of Floating Cooling Systems  

E-Print Network (OSTI)

. The application of a floating cooling concept to evaporative heat rejection systems can have significant impact on improving plant performance. The floating cooling concept refers to the optimization of yearly plant output and energy consumption by taking...

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

1983-01-01T23:59:59.000Z

173

Improving the Water Efficiency of Cooling Production System  

E-Print Network (OSTI)

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...

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

2010-01-01T23:59:59.000Z

174

NASA's Marshall Space Flight Center Improves Cooling System Performance  

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

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

175

Corrosion in HVDC valve cooling systems  

SciTech Connect

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.

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

1997-04-01T23:59:59.000Z

176

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

SciTech Connect

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.

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

2014-03-01T23:59:59.000Z

177

Shutdown system for a nuclear reactor  

DOE Patents (OSTI)

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.

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

178

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

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

cooling systems that are providing 30%-70% energy and cost savings for homeowners in Jordan, New York. Demand for these systems is growing; nationally, shipments of geothermal...

179

Economizer refrigeration cycle space heating and cooling system and process  

DOE Patents (OSTI)

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.

Jardine, D.M.

1983-03-22T23:59:59.000Z

180

Economizer refrigeration cycle space heating and cooling system and process  

DOE Patents (OSTI)

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.

Jardine, Douglas M. (Colorado Springs, CO)

1983-01-01T23:59:59.000Z

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


181

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

SciTech Connect

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

Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro [Nuclear Hydrogen and Heat Application Research Center, Japan Atomic Energy Agency, Oarai-machi, Ibaraki-ken, 311-1394 (Japan)

2012-06-06T23:59:59.000Z

182

Physical Similitude in Hierarchical Engineered Systems  

E-Print Network (OSTI)

reliability analysis into the design of passive cooling systems with an application to a gas-cooled reactor.

Blandford, Edward David

2010-01-01T23:59:59.000Z

183

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

SciTech Connect

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.

Buden, D.

1993-08-11T23:59:59.000Z

184

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

E-Print Network (OSTI)

for Modular, Scalable Liquid-Rack Cooling Systems in DataFOR A MODULAR, SCALABLE LIQUID-RACK COOLING SYSTEM IN DATA3 M ODULAR LIQUID - RACK COOLING

Xu, TengFang

2009-01-01T23:59:59.000Z

185

Cooling load design tool for UFAD systems.  

E-Print Network (OSTI)

ratio of time between Fan Coil Units Perimeter Zone Linearand underfloor fan coil units. cooling contribution of

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

2010-01-01T23:59:59.000Z

186

Special Property Assessment for Renewable Heating and Cooling Systems |  

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

Special Property Assessment for Renewable Heating and Cooling Special Property Assessment for Renewable Heating and Cooling Systems Special Property Assessment for Renewable Heating and Cooling Systems < Back Eligibility Commercial Industrial Residential Savings Category Heating & Cooling Commercial Heating & Cooling Solar Heating Program Info State Maryland Program Type Property Tax Incentive Rebate Amount Eligible property is assessed at no more than the value of a conventional system Provider Department of Assessments and Taxation 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 more than the value of a conventional system for property tax purposes if no conventional system

187

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

E-Print Network (OSTI)

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

Wang, Jing, 1976-

2004-01-01T23:59:59.000Z

188

The CANDU Reactor System: An Appropriate Technology  

Science Journals Connector (OSTI)

The CANDU Reactor System: An Appropriate Technology...Chalk River, Ontario, Canada K0J 1J0 CANDU power reactors are characterized by the combination...breeder. These and other features make the CANDU system an appropriate technology for countries...

J. A. L. Robertson

1978-02-10T23:59:59.000Z

189

Performance analysis of a Pb-Bi cooled fast reactor - PEACER-300 in proliferation resistance and transmutation aspects  

SciTech Connect

A design study of 850 MWt lead-bismuth cooled reactor cores is performed to maximize the transmutation of both TRU nuclides in homogeneous fuel pin and long-lived fission products in separate target pins. Transmutation of minor actinide under a closed recycling was analyzed with assumption that decontamination factors in pyro-reprocessing plant data be reasonably high. The optimized design parameter were chosen as of a flat core shape with 50 cm in active core height and 5 m in core diameter, loaded with 17 x 17 arrayed fuel assemblies. A pitch to diameter ratio is 2.2, operating coolant temperature range is 300 deg. C-400 deg. C, and core consists of 3 different enrichment zones with one year cycle length. In safety aspects, this core design satisfied large negative temperature feedback coefficients, and sufficient shutdown margin by primary shutdown system with 20 B{sub 4}C control assemblies and by secondary shutdown system with 40 w/o enriched 12 B{sub 4}C control assemblies. Performance of designed core showed a high transmutation capability with support ratio of 2.085 and less TEX values than other reactor types. Better proliferation resistance could be achieved than other reactor types. (authors)

Lim, J. Y.; Kim, M. H. [Dept. of Nuclear Engineering, Kyung Hee Univ., Yongin-shi, Gyeonggi-do, 449-701 (Korea, Republic of)

2006-07-01T23:59:59.000Z

190

Research and Development of High Temperature Light Water Cooled Reactor Operating at Supercritical-Pressure in Japan  

SciTech Connect

This paper summarizes the status and future plans of research and development of the high temperature light water cooled reactor operating at supercritical-pressure in Japan. It includes; the concept development; material for the fuel cladding; water chemistry under supercritical pressure; thermal hydraulics of supercritical fluid; and the conceptual design of core and plant system. Elements of concept development of the once-through coolant cycle reactor are described, which consists of fuel, core, reactor and plant system, stability and safety. Material studies include corrosion tests with supercritical water loops and simulated irradiation tests using a high-energy transmission electron microscope. Possibilities of oxide dispersion strengthening steels for the cladding material are studied. The water chemistry research includes radiolysis and kinetics of supercritical pressure water, influence of radiolysis and radiation damage on corrosion and behavior on the interface between water and material. The thermal hydraulic research includes heat transfer tests of single tube, single rod and three-rod bundles with a supercritical Freon loop and numerical simulations. The conceptual designs include core design with a three-dimensional core simulator and sub-channel analysis, and balance of plant. (authors)

Yoshiaki Oka [Nuclear Engineering Research Laboratory, The University of Tokyo, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 112-0006 (Japan); Katsumi Yamada [Isogo Nuclear Engineering Center, Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan)

2004-07-01T23:59:59.000Z

191

BSU GHP District Heating and Cooling System (Phase I)  

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

BSU GHP District Heating and Cooling System (Phase I) James Lowe Ball State University May 03, 2010 This presentation does not contain any proprietary confidential, or otherwise...

192

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

Open Energy Info (EERE)

Last modified on July 22, 2011. Project Title Cedarville School District Retrofit of Heating and Cooling Systems with Geothermal Heat Pumps and Ground Source Water Loops Project...

193

Design and simulation of solar absorption cooling systems.  

E-Print Network (OSTI)

??The absorption chiller modeling is then exploited in a computer tool that gives valuable information on the planning of solar cooling systems. The design of… (more)

NURZIA, GIOVANNI

2008-01-01T23:59:59.000Z

194

APPLICATION OF TURBOMACHINERY IN SOLAR-ASSISTED RANKINE COOLING SYSTEMS  

E-Print Network (OSTI)

machinery. and the solar-assisted approach at these higherevaluation of the solar-assisted Rankine cycle could beTURBOMACHINER Y IN SOLAR - ASSISTED RANKINE COOLING SYSTEMS

Leech, J.

2010-01-01T23:59:59.000Z

195

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

DOE Patents (OSTI)

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.

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

1985-01-01T23:59:59.000Z

196

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

SciTech Connect

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.

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

1988-01-01T23:59:59.000Z

197

Mechanical properties of welds in commercial alloys for high-temperature gas-cooled reactor components  

SciTech Connect

Weld properties of Hastelloy-X, Incoloy alloy 800H (with and without Inconel-82 cladding), and 2 1/4 Cr-1 Mo are being studied to provide design data to support the development of steam generator, core auxiliary heat exchanger, and metallic thermal barrier components of the high-temperature gas-cooled reactor (HTGR) steam cycle/cogeneration plant. Tests performed include elevated-temperature creep rupture tests and tensile tests. So far, data from the literature and from relatively short-term tests at GA Technologies Inc. indicate that the weldments are satisfactory for HTGR application.

Lindgren, J.R.; Li, C.C.; Ryder, R.H.; Thurgood, B.E.

1984-07-01T23:59:59.000Z

198

Methods for nondestructive testing of austenitic high-temperature gas-cooled reactor components  

SciTech Connect

Safety-relevant components of high-temperature gas-cooled reactor components are mostly fabricated in nickel-based alloys and austenitic materials like Inconel-617, Hastelloy-X, Nimonic-86, or Incoloy-800H. Compared to ferritic steels, these austenitic materials can have a coarse-grained microstructure, especially in weldments and castings. Coarse-grained or elastic anisotropic materials are difficult to inspect with ultrasonics due to strong attenuation, high noise level (scattering, ''grass'' indications), and sound beam distortions (skewing, splitting, and mode conversion). Only few results dealing with the nondestructive testing of nickel-based alloys are known. The problem area, solutions, and first experiences are reported.

Gobbels, K.; Kapitza, H.

1984-09-01T23:59:59.000Z

199

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

SciTech Connect

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.

Gorensek, M.

2011-07-06T23:59:59.000Z

200

High-temperature gas-cooled-reactor steam-methane reformer design  

SciTech Connect

The concept of the long distance transportation of process heat energy from a High Temperature Gas Cooled Reactor (HTGR) heat source, based on the steam reforming reaction, is currently being evaluated as an energy source/application for use early in the 21st century. The steam-methane reforming reaction is an endothermic reaction at temperatures approximately 700/sup 0/C and higher, which produces hydrogen, carbon monoxide and carbon dioxide. The heat of the reaction products can then be released, after being pumped to industrial site users, in a methanation process producing superheated steam and methane which is then returned to the reactor plant site. In this application the steam reforming reaction temperatures are produced by the heat energy from the core of the HTGR through forced convection of the primary or secondary helium circuit to the catalytic chemical reactor (steam reformer). This paper summarizes the design of a helium heated steam reformer utilized in conjunction with a 1170 MW(t) intermediate loop, 850/sup 0/C reactor outlet temperature, HTGR process heat plant concept. This paper also discusses various design considerations leading to the mechanical design features, the thermochemical performance, materials selection and the structural design analysis.

Impellezzeri, J.R.; Drendel, D.B.; Odegaard, T.K.

1981-01-20T23:59:59.000Z

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


201

Self-actuating reactor shutdown system  

DOE Patents (OSTI)

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.

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

1988-01-01T23:59:59.000Z

202

Property:Distributed Generation System Heating-Cooling Application | Open  

Open Energy Info (EERE)

Heating-Cooling Application Heating-Cooling Application Jump to: navigation, search This is a property of type Page. Pages using the property "Distributed Generation System Heating-Cooling Application" Showing 21 pages using this property. D Distributed Generation Study/10 West 66th Street Corp + Domestic Hot Water +, Space Heat and/or Cooling + Distributed Generation Study/Aisin Seiki G60 at Hooligans Bar and Grille + Domestic Hot Water + Distributed Generation Study/Arrow Linen + Domestic Hot Water + Distributed Generation Study/Dakota Station (Minnegasco) + Space Heat and/or Cooling +, Other + Distributed Generation Study/Elgin Community College + Space Heat and/or Cooling +, Domestic Hot Water + Distributed Generation Study/Emerling Farm + Domestic Hot Water +, Process Heat and/or Cooling +

203

SciTech Connect: Nuclear power reactor instrumentation systems...  

Office of Scientific and Technical Information (OSTI)

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

204

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

SciTech Connect

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.

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

205

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

SciTech Connect

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.

Jorge Navarro

2013-12-01T23:59:59.000Z

206

Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors  

SciTech Connect

A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

Kasten, P.R.; Bartine, D.E.

1981-01-01T23:59:59.000Z

207

Remediation of a large contaminated reactor cooling reservoir: Resolving and environmental/regulatory paradox  

SciTech Connect

This paper presents a case study of a former reactor cooling water reservoir, PAR Pond, located Savannah River Site. PAR Pond, a 2640 acre, man-made reservoir was built in 1958 and until 1988, received cooling water from two DOE nuclear production reactors, P and R. The lake sediments were contaminated with low levels of radiocesium (CS-137) and transuranics in the late 1950s and early 1960s because of leaking fuel elements. Elevated levels of mercury accumulated in the sediments from pumping water from the Savannah River to maintain a full pool. PAR Ponds` stability, size, and nutrient content made a significant, unique, and highly studied ecological resource for fish and wildlife populations until it was partially drained in 1991 due to a depression in the downslope of the earthen dam. The drawdown, created 1340 acres of exposed, radioactively contaminated sediments along 33 miles of shoreline. This led US EPA to declare PAR Pond as a CERCLA operable unit subject to remediation. The drawdown also raised concerns for the populations of aquatic plants, fish, alligators, and endangered species and increased the potential for off-site migration of contaminated wildlife from contact with the exposed sediments. Applicable regulations, such as NEPA and CERCLA, require wetland loss evaluations, human health and ecological risk assessments, and remediation feasibility studies. DOE is committed to spending several million dollars to repair the dam for safety reasons, even though the lake will probably not be used for cooling purposes. At the same time, DOE must make decisions whether to refill and expend additional public funds to maintain a full pool to reduce the risks defined under CERCLA or spend hundreds of millions in remediation costs to reduce the risks of the exposed sediments.

Bowers, J.A.: Gladden, J.B.; Hickey, H.M.; Jones, M.P.; Mackey, H.E.; Mayer, J.J. [Westinghouse Savannah River Co., Aiken, SC (United States); Doswell, A. [USDOE, Washington, DC (United States)

1994-05-01T23:59:59.000Z

208

New Fuel Cycle and Fuel Management Options in Heavy Liquid Metal-Cooled Reactors  

Science Journals Connector (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Fuel Management of Reactors Other Than Light Water Reactors

Ehud Greenspan; Pavel Hejzlar; Hiroshi Sekimoto; Georgy Toshinsky; David Wade

209

Heat pump system with selective space cooling  

DOE Patents (OSTI)

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.

Pendergrass, J.C.

1997-05-13T23:59:59.000Z

210

Heat pump system with selective space cooling  

DOE Patents (OSTI)

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.

Pendergrass, Joseph C. (Gainesville, GA)

1997-01-01T23:59:59.000Z

211

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors  

SciTech Connect

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01T23:59:59.000Z

212

Cooling system early-stage design tool for naval applications  

E-Print Network (OSTI)

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 ...

Fiedel, Ethan R

2011-01-01T23:59:59.000Z

213

Closed loop air cooling system for combustion turbines  

DOE Patents (OSTI)

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.

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

1998-01-01T23:59:59.000Z

214

Investigations of Alternative Steam Generator Location and Flatter Core Geometry for Lead-Cooled Fast Reactors  

SciTech Connect

This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MW{sub th} power. This was compared to a similar design, but with the steam generators located in the downcomers. The transients investigated were Total-Loss-of-Power and unprotected Loss-Of-Flow. It is shown that this reactor peaks at 1041 K after 29 hours during a Total-Loss-Of-Power accident. The difference between locating the steam generators in the risers and the downcomers is insignificant for this accident type. During an unprotected Loss-Of-Flow accident at full power, the core outlet temperature stabilizes at 1010 K, which is 337 K above nominal outlet temperature. The second investigation concerns a 1426 MW{sub th} critical reactor where the influence of the core height versus the core outlet temperature is studied during an unprotected Loss-Of-Flow and Total-Loss-Of-Power accident. A pancake type core geometry of 1.0 m height and 5.8 m diameter, is compared to a compact core of 2 m height and 4.5 m diameter. Moderators, like BeO and hydrides, and their influence on safety coefficients and burnup swings are also presented. Both cores incinerate transuranics from spent LWR fuel with minor actinide fraction of 5%. We show that LFRs can be designed both to breed and burn transuranics from LWRs. It is shown that the hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. The computational fluid dynamics code STAR-CD was used for all thermal hydraulic calculations, and the MCNP and MCB for neutronics, and burn-up calculations. (authors)

Carlsson, Johan; Tucek, Kamil; Wider, Hartmut [Joint Research Centre, Institute for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands)

2006-07-01T23:59:59.000Z

215

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

SciTech Connect

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.

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

2013-11-01T23:59:59.000Z

216

Energy efficient operation strategy design for the combined cooling, heating and power system.  

E-Print Network (OSTI)

??Combined cooling, heating and power (CCHP) systems are known as trigeneration systems, designed to provide electricity, cooling and heating simultaneously. The CCHP system has become… (more)

Liu, Mingxi

2012-01-01T23:59:59.000Z

217

Redesigning Process Cooling Systems in the Plastics Industry  

E-Print Network (OSTI)

REDESIGNING PROCESS COOLING SYSTEMS IN THE PLASTICS INDUSTRY Glen R Anderson - Senior Energy Analyst - etc Group, Inc - Salt Lake City, UT ABSTRACT Lifetime Products grew their plastics division rapidly starting in the mid 1990’s.... During this growth, their support systems were designed with one thing in mind – ensuring adequate capacity. Energy consumption was a much lower priority with their process cooling systems, resulting in inefficient chillers, oversized pumps...

Anderson, G. R.

2006-01-01T23:59:59.000Z

218

Power Plant Cooling Systems: Policy Alternatives  

Science Journals Connector (OSTI)

...contrast, provide convenient field laboratories for examining...barrels per day of additional oil equivalent would be required...num-ber of citations is the cumulative total and does not include...of a Cooling Lake Fishery, Illinois Natural History Survey, project...

John Z. Reynolds

1980-01-25T23:59:59.000Z

219

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

DOE Patents (OSTI)

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.

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

2013-09-03T23:59:59.000Z

220

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

DOE Patents (OSTI)

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.

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

2011-03-01T23:59:59.000Z

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


221

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

SciTech Connect

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.

Phillip Mills

2012-02-01T23:59:59.000Z

222

High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics  

SciTech Connect

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.

Larry Demick

2011-08-01T23:59:59.000Z

223

Solar cooling systems for climate change mitigation: A review  

Science Journals Connector (OSTI)

Abstract The impact of global warming and an increase in the indoor cooling equipments using energy sources other than conventional energy have become very attractive because they can reduce consumption of fossil fuels as well as harmful emissions in to the environment. The solar energy is one of the readily available forms of renewable energy which can be used to operate the cooling equipments depending on the geographical location of the area where solar cooling system needs to be installed. The effectiveness of solar cooling also needs to be evaluated based on various performance indicating parameters. However, different types of solar collectors also need to be evaluated in order to find out their feasibility for cooling applications. Thus in this article review of different types of solar cooling technologies have been carried out. The study reveals that evacuated tube collectors are best option for solar cooling where as desiccant cooling helps in improving the indoor air quality. Also, thermal energy storage and ejector based solar cooling efficiently improves the performance besides energy saving.

S. Anand; A. Gupta; S.K. Tyagi

2015-01-01T23:59:59.000Z

224

Preoperational test report, primary ventilation condenser cooling system  

SciTech Connect

This represents the preoperational test report for the Primary Ventilation Condenser Cooling System, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system uses a closed chilled water piping loop to provide offgas effluent cooling for tanks AY101, AY102, AZ1O1, AZ102; the offgas is cooled from a nominal 100 F to 40 F. Resulting condensation removes tritiated vapor from the exhaust stack stream. The piping system includes a package outdoor air-cooled water chiller with parallel redundant circulating pumps; the condenser coil is located inside a shielded ventilation equipment cell. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System.

Clifton, F.T.

1997-10-29T23:59:59.000Z

225

Transient thermal analysis of a space reactor power system  

E-Print Network (OSTI)

Thermoelectric Power Conversion Module Heat Pipe Radiator Module . Auxiliary Modules . Flow of Calculation . Transient Test Cases Studied Summary . 10 10 CHAPTER II. ENERGY EQUATION FINITE DIFFERENCING . . 12 Energy Equation for a Solid Finite..., but this stud~ uses a generic liquid metal cooled fast reactor concept as the model to test the code. The space power svstem to be modeled consists of a liquid lithium cooled fast reactor, primarv and secondary loops svith a sell-induced thermoelectric...

Gaeta, Michael J.

1988-01-01T23:59:59.000Z

226

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

SciTech Connect

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.

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

227

Cooling system for three hook ring segment  

DOE Patents (OSTI)

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.

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

2014-08-26T23:59:59.000Z

228

Evaluation of Indirect Combined Cycle in Very High Temperature Gas--Cooled Reactor  

SciTech Connect

The U.S. Department of Energy and Idaho National Laboratory are developing a very high temperature reactor to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is twofold: (a) efficient, low-cost energy generation and (b) hydrogen production. Although a next-generation plant could be developed as a single-purpose facility, early designs are expected to be dual purpose, as assumed here. A dual-purpose design with a combined cycle of a Brayton top cycle and a bottom Rankine cycle was investigated. An intermediate heat transport loop for transporting heat to a hydrogen production plant was used. Helium, CO2, and a helium-nitrogen mixture were studied to determine the best working fluid in terms of the cycle efficiency. The relative component sizes were estimated for the different working fluids to provide an indication of the relative capital costs. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the cycle were performed to determine the effects of varying conditions in the cycle. This gives some insight into the sensitivity of the cycle to various operating conditions as well as trade-offs between efficiency and component size. Parametric studies were carried out on reactor outlet temperature, mass flow, pressure, and turbine cooling.

Chang Oh; Robert Barner; Cliff Davis; Steven Sherman; Paul Pickard

2006-10-01T23:59:59.000Z

229

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

SciTech Connect

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.

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

230

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

SciTech Connect

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)

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

231

Control system for a small fission reactor  

DOE Patents (OSTI)

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.

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

1985-02-08T23:59:59.000Z

232

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

SciTech Connect

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)

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

2012-07-01T23:59:59.000Z

233

Improving the liquid-cooling systems of power units and technological equipment  

Science Journals Connector (OSTI)

Processes in the liquid cooling systems of power units and technological equipment are considered. Criteria ... of the energy and resource aspects of the cooling systems.

V. A. Zhukov

2011-12-01T23:59:59.000Z

234

Active noise canceling system for mechanically cooled germanium radiation detectors  

DOE Patents (OSTI)

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.

Nelson, Karl Einar; Burks, Morgan T

2014-04-22T23:59:59.000Z

235

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

SciTech Connect

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.

Ronald Farris; David Gertman; Jacques Hugo

2014-03-01T23:59:59.000Z

236

ERANOS 2.1 : International Code System for GEN IV Fast Reactor Analysis  

SciTech Connect

The aim of the paper is to present the new release of the European Reactor Analysis Optimized code System, ERANOS 2.1. This version has been developed and validated to establish a suitable basis for reliable neutronic calculations of current, as well as advanced fast reactor cores of the GEN IV International Forum. The latest version of the ERANOS code and data system, ERANOS 2.1, contains all of the functions required for reference and design calculations of Liquid Metal Fast Reactors (LMFR's), with extended capabilities for treating advanced reactor fuel subassemblies and cores of Gas Cooled Fast Reactors (GCFR's). The releases of recent, modern neutron data libraries: JENDL-3.3 in 2002, JEFF-3.1 in 2005 and soon ENDF/B-VII are also available. (authors)

Ruggieri, J.M.; Tommasi, J.; Lebrat, J.F.; Suteau, C.; Plisson-Rieunier, D.; De Saint Jean, C.; Rimpault, G.; Sublet, J.C. [CEA/DEN/DER/SPRC/LEPH, CEA/Cadarache, 13108 Saint Paul Lez Durance, Cedex (France)

2006-07-01T23:59:59.000Z

237

Dynamic Impregnator Reactor System (Poster), NREL (National Renewable Energy Laboratory)  

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

Several unit operations are combined into Several unit operations are combined into one robust system, off ering fl exible and staged process confi gurations in one vessel. Spraying, soaking, low-severity pretreat- ment, enzymatic hydrolysis, fermentation, concentration/evaporation, and distillation are amongst its many capabilities. * 1,900 L Horizontal Paddle Blender Vessel with Sidewall Liquid Drains * 6-60 rpm / 50 HP Tri-Directional Agitator * 3.4 bar & Vacuum ASME Design, 316L Stainless Steel * Heating/Cooling Jacket using Water or Steam * 150 L Chemical Mix Tank & Pump with Spray Injectors * Vent Condenser with Collection Tank and Vacuum Pump Dynamic Impregnator Reactor System Multifaceted system designed for complex feedstock impregnation and processing Integrated Biorefi nery Research Facility | NREL * Golden, Colorado | December 15, 2011 | NREL/PO-5100-56156

238

Design of an Actinide-Burning, Lead or Lead-Bismuth Cooled Reactor that Produces Low-Cost Electricity  

SciTech Connect

The purpose of this Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology (MIT) University Research Consortium (URC) project is to investigate the suitability of lead or lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. Work has been accomplished in four major areas of research: core neutronic design, material compatibility, plant engineering, and coolant activation. In the area of core neutronic design, the reactivity vs. burnup and discharge isotopics of both non-fertile and fertile fuels were evaluated. An innovative core for pure actinide burning that uses streaming, fertile-free fuel assemblies was studied in depth. This particular core exhibits excellent reactivity performance upon coolant voiding, even for voids that occur in the core center, and has a transuranic (TRU) destruction rate that is comparable to the proposed accelerator transmutation of waste (ATW) facility. These studies suggest that a core can be designed to achieve a long life while maintaining safety and minimizing waste. In the area of material compatibility studies, an experimental apparatus for the investigation of the flow-assisted dissolution and precipitation (corrosion) of potential fuel cladding and structural materials has been designed and built at the INEEL. The INEEL forced-convection corrosion cell consists of a small heated vessel with a shroud and gas flow system. The corrosion cell is being used to test steel that is commercially available in the United States to temperatures above 650°C. Progress in plant engineering was made for two reactor concepts, one utilizing an indirect cycle with heat exchangers and the other utilizing a direct-contact steam cycle. The evaluation of the indirect cycle designs has investigated the effects of various parameters to increase electric production at full power. For the direct-contact reactor, major issues related to the direct-contact heat transfer rate and entrainment and carryover of liquid lead-bismuth to the turbine have been identified and analyzed. An economic analysis approach was also developed to determine the cost of electricity production in the lead-bismuth reactor. The approach will be formulated into a model and applied to develop scientific cost estimates for the different reactor designs and thus aid in the selection of the most economic option. In the area of lead-bismuth coolant activation, the radiological hazard was evaluated with particular emphasis on the direct-contact reactor. In this system, the lack of a physical barrier between the primary and secondary coolant favors the release of the alpha-emitter Po?210 and its transport throughout the plant. Modeling undertaken on the basis of the scarce information available in the literature confirmed the importance of this issue, as well as the need for experimental work to reduce the uncertainties on the basic characteristics of volatile polonium chemical forms.

Mac Donald, Philip Elsworth; Weaver, Kevan Dean; Davis, Cliff Bybee; MIT folks

2000-07-01T23:59:59.000Z

239

Heating and Cooling System Support Equipment Basics | Department of Energy  

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

Heating and Cooling System Support Equipment Basics Heating and Cooling System Support Equipment Basics Heating and Cooling System Support Equipment Basics July 30, 2013 - 3:28pm Addthis Thermostats and ducts provide opportunities for saving energy. Dehumidifying heat pipes provide a way to help central air conditioners and heat pumps dehumidify air. Electric and gas meters allow users to track energy use. Thermostats Programmable thermostats can store and repeat multiple daily settings. Users can adjust the times heating or air-conditioning is activated according to a pre-set schedule. Visit the Energy Saver website for more information about thermostats and control systems in homes. Ducts Efficient and well-designed duct systems distribute air properly throughout a building, without leaking, to keep all rooms at a comfortable

240

Modeling and Simulation of a Solar Assisted Desiccant Cooling System  

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

Modeling and Simulation of a Solar Assisted Desiccant Cooling System Modeling and Simulation of a Solar Assisted Desiccant Cooling System Speaker(s): Chadi Maalouf Date: December 2, 2004 - 12:00pm Location: Bldg. 90 Seminar Host/Point of Contact: Peng Xu Increased living standards and high occupants comfort demands lead to a growth in air conditioning market. This results in high energy consumption and high CO2 emissions. For these reasons, the solar desiccant cooling system is proposed as an alternative to traditional air conditioning systems. This system comprises a desiccant wheel containing Lithium Chloride in tandem with a rotating heat exchanger and two humidifiers on both supply and return air. The required regeneration temperature for the desiccant wheel varies between 40oC and 70oC which makes possible the use

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241

Mining Gold from your Cooling Water System  

E-Print Network (OSTI)

to be achieved. GPM 2 /GPM 1 = RPM 2 /RPM 1 Equation (1) (RPM 2 /RPM 1 ) 3 = HP 2 /HP 1 Equation (2) ESL-IE-07-05-25 Proceedings from the Twenty-ninth Industrial Energy Technology Conference, New Orleans, LA, May 8-11, 2007. COOLING WATER PUMPING Pumping... Apr May Jun Jul Aug Sep Oct Nov Months Ri ver l eve l ( f t ) 0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 80.00 90.00 T e mp er at ur e ( F) Average River Level Average River Temperature ESL-IE-07-05-25 Proceedings from the Twenty...

Mendez, T.

242

Dynamic Impregnator Reactor System (Poster)  

SciTech Connect

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.

Not Available

2012-09-01T23:59:59.000Z

243

Cedarville School District Retrofit of Heating and Cooling Systems with  

Open Energy Info (EERE)

School District Retrofit of Heating and Cooling Systems with School District Retrofit of Heating and Cooling Systems with Geothermal Heat Pumps and Ground Source Water Loops Geothermal Project 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 Source Water Loops Project Type / Topic 1 Recovery Act - Geothermal Technologies Program: Ground Source Heat Pumps Project Type / Topic 2 Topic Area 1: Technology Demonstration Projects Project Description - Improve the indoor air quality and lower the cost of cooling and heating the buildings that make up the campus of Cedarville High School, Middle School and Elementary School. - Provide jobs, and reduce requirements of funds for the capital budget of the School District, and thus give relief to taxpayers in this rural region during a period of economic recession. - The new Heat Pumps will be targeted to perform at very high efficiency with EER (energy efficiency ratios) of 22+/-. System capacity is planned at 610 tons. - Remove unusable antiquated existing equipment and systems from the campus heating and cooling system, but utilize ductwork, piping, etc. where feasible. The campus is served by antiquated air conditioning units combined with natural gas, and with very poor EER estimated at 6+/-. - Monitor for 3 years the performance of the new systems compared to benchmarks from the existing system, and provide data to the public to promote adoption of Geothermal technology. - The Geothermal installation contractor is able to provide financing for a significant portion of project funding with payments that fall within the energy savings resulting from the new high efficiency heating and cooling systems.

244

Diversified emergency core cooling in CANDU with a passive moderator heat rejection system  

SciTech Connect

Contemporary Canada deuterium uranium (CANDU) reactor designs employ redundancy and a considerable level of diversity in the safety systems. Thus redundancy and a high level of diversity exist in the two safety shutdown systems, which, in turn, are independent of the reactor regulating system, Also, a considerable level of diversity exists in the two emergency core cooling systems: the emergency coolant injection system and the CANDU low-pressure low-temperature moderator, which, in surrounding the fuel channels, can act as a redundant heat sink. However, common cooling water and common electrical supplies are used for heat rejection from the emergency coolant and from the moderator. Scenarios involving failures in these common supplies are the main contributors to the CANDU core-melt frequency of 4 x 10{sup -6} yr{sup -1}. The purpose of this work is to develop a passive moderator heat rejection system that, in avoiding the use of pumps and electrical supplies, enhances the independence of the moderator as a heat sink.

Spinks, N.J. [Atomic Energy of Canada Limited, Ontario (Canada)

1995-12-31T23:59:59.000Z

245

Modular High-Temperature Gas-Cooled Reactor short term thermal response to flow and reactivity transients  

SciTech Connect

The analyses reported here have been conducted at the Oak Ridge National Laboratory (ORNL) for the US Nuclear Regulatory Commission's (NRC's) Division of Regulatory Applications of the Office of Nuclear Regulatory Research. The short-term thermal response of the Modular High-Temperature Gas-Cooled Reactor (MHTGR) is analyzed for a range of flow and reactivity transients. These include loss of forced circulation (LOFC) without scram, moisture ingress, spurious withdrawal of a control rod group, hypothetical large and rapid positive reactivity insertion, and a rapid core cooling event. The coupled heat transfer-neutron kinetics model is also described.

Cleveland, J.C.

1988-01-01T23:59:59.000Z

246

Vertical Pretreatment Reactor System (Poster)  

SciTech Connect

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

Not Available

2012-09-01T23:59:59.000Z

247

POWER CYCLE AND STRESS ANALYSES FOR HIGH TEMPERATURE GAS-COOLED REACTOR  

SciTech Connect

The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with three turbines and four compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with three stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to various operating conditions as well as trade offs between efficiency and capital cost. Parametric studies were carried out on reactor outlet temperature, mass flow, pressure, and turbine cooling. Recommendations on the optimal working fluid for each configuration were made. Engineering analyses were performed for several configurations of the intermediate heat transport loop that transfers heat from the nuclear reactor to the hydrogen production plant. The analyses evaluated parallel and concentric piping arrangements and two different working fluids, including helium and a liquid salt. The thermal-hydraulic analyses determined the size and insulation requirements for the hot and cold leg pipes in the different configurations. Mechanical analyses were performed to determine hoop stresses and thermal expansion characteristics for the different configurations. Economic analyses were performed to estimate the cost of the various configurations.

Oh, Chang H; Davis, Cliff; Hawkes, Brian D; Sherman, Steven R

2007-05-01T23:59:59.000Z

248

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- ing a case study using test cases prepared by domain experts for reactor protection system software) are widely used to implement safety- critical systems such as nuclear reactor protection systems, testing

249

A Peltier cooling system for SiPM temperature stabilization  

E-Print Network (OSTI)

A Peltier cooling system for SiPM temperature stabilization von Simon Nieswand Bachelorarbeit auĂ?en 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

Hebbeker, Thomas

250

A COOLING SYSTEM FOR BUIDINGS USING WIND ENERGY  

E-Print Network (OSTI)

A COOLING SYSTEM FOR BUIDINGS USING WIND ENERGY Hamid Daiyan Islamic Azad University - Semnan in dray land, and only uses wind energy for conditioning. It technologies date back over 1000 years. Wind system, Wind energy, Temperature Fig.1 Wind tower of Doulat-Abad garden of Yazd with it's altitude is 33

251

Westinghouse Small Modular Reactor balance of plant and supporting systems design  

SciTech Connect

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)

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

252

CONTROL SYSTEM FOR SOLAR HEATING and COOLING  

E-Print Network (OSTI)

sensors and control valves used in our generalized experimental system. The experimental solarsensors are remotely located at critical (in terms of decision-making) locations in the solar

Dols, C.

2010-01-01T23:59:59.000Z

253

Geometric effect on cooling power and performance of an integrated thermoelectric generation-cooling system  

Science Journals Connector (OSTI)

Abstract Geometric design of an integrated thermoelectric generation-cooling system is performed numerically using a finite element method. In the system, a thermoelectric cooler (TEC) is powered directly by a thermoelectric generator (TEG). Two different boundary conditions in association with the effects of contact resistance and heat convection on system performance are taken into account. The results suggest that the characteristics of system performance under varying TEG length are significantly different from those under altering TEC length. When the TEG length is changed, the entire behavior of system performance depends highly on the boundary conditions. On the other hand, the maximum distributions of cooling power and coefficient of performance (COP) are exhibited when the TEC length is altered, whether the hot surface of TEG is given by a fixed temperature or heat transfer rate. The system performance will be reduced once the contact resistance and heat convection are considered. When the lengths of TEG and TEC vary, the maximum reduction percentages of system performance are 12.45% and 18.67%, respectively. The numerical predictions have provided a useful insight into the design of integrated TEG–TEC systems.

Wei-Hsin Chen; Chien-Chang Wang; Chen-I Hung

2014-01-01T23:59:59.000Z

254

Heat exchanger design considerations for high temperature gas-cooled reactor (HTGR) plants  

SciTech Connect

Various aspects of the high-temperature heat exchanger conceptual designs for the gas turbine (HTGR-GT) and process heat (HTGR-PH) plants are discussed. Topics include technology background, heat exchanger types, surface geometry, thermal sizing, performance, material selection, mechanical design, fabrication, and the systems-related impact of installation and integration of the units in the prestressed concrete reactor vessel. The impact of future technology developments, such as the utilization of nonmetallic materials and advanced heat exchanger surface geometries and methods of construction, is also discussed.

McDonald, C.F.; Vrable, D.L.; Van Hagan, T.H.; King, J.H.; Spring, A.H.

1980-02-01T23:59:59.000Z

255

CONTROL SYSTEM FOR SOLAR HEATING and COOLING  

E-Print Network (OSTI)

Solar Energy Society Meeting, Los Angeles, California, Julysolar in- solation measuring stations in northern and central California (California 94720 August 1975 A control system is being developed that will be capable of operating solar

Dols, C.

2010-01-01T23:59:59.000Z

256

Wind turbine generators having wind assisted cooling systems and cooling methods  

DOE Patents (OSTI)

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.

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

257

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

SciTech Connect

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.

Scheele, Randall D.; Casella, Andrew M.

2010-09-28T23:59:59.000Z

258

Microchannel Reactor System for Catalytic Hydrogenation  

SciTech Connect

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.

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

2010-12-22T23:59:59.000Z

259

Comparative Study Between Air-Cooled and Water-Cooled Condensers of the Air-Conditioning Systems  

E-Print Network (OSTI)

consumptions. The cooling capacities for WC and AC systems were 373 and 278 tons-of- refrigeration, respectively. It was found that for the same cooling production, the peak power demand and the daily energy consumption of the WC system were 45 and 32% less...

Maheshwari, G. P.; Mulla Ali, A. A.

2004-01-01T23:59:59.000Z

260

Staged membrane oxidation reactor system  

DOE Patents (OSTI)

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.

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

2012-09-11T23:59:59.000Z

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


261

Staged membrane oxidation reactor system  

DOE Patents (OSTI)

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.

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

2014-05-20T23:59:59.000Z

262

Staged membrane oxidation reactor system  

DOE Patents (OSTI)

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.

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

2013-04-16T23:59:59.000Z

263

BETTER DUCT SYSTEMS FOR HOME HEATING AND COOLING.  

SciTech Connect

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.

ANDREWS,J.

2001-01-01T23:59:59.000Z

264

Reactor operation safety information document  

SciTech Connect

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)

Not Available

1990-01-01T23:59:59.000Z

265

Analysis and Development of A Robust Fuel for Gas-Cooled Fast Reactors  

SciTech Connect

The focus of this effort was on the development of an advanced fuel for gas-cooled fast reactor (GFR) applications. This composite design is based on carbide fuel kernels dispersed in a ZrC matrix. The choice of ZrC is based on its high temperature properties and good thermal conductivity and improved retention of fission products to temperatures beyond that of traditional SiC based coated particle fuels. A key component of this study was the development and understanding of advanced fabrication techniques for GFR fuels that have potential to reduce minor actinide (MA) losses during fabrication owing to their higher vapor pressures and greater volatility. The major accomplishments of this work were the study of combustion synthesis methods for fabrication of the ZrC matrix, fabrication of high density UC electrodes for use in the rotating electrode process, production of UC particles by rotating electrode method, integration of UC kernels in the ZrC matrix, and the full characterization of each component. Major accomplishments in the near-term have been the greater characterization of the UC kernels produced by the rotating electrode method and their condition following the integration in the composite (ZrC matrix) following the short time but high temperature combustion synthesis process. This work has generated four journal publications, one conference proceeding paper, and one additional journal paper submitted for publication (under review). The greater significance of the work can be understood in that it achieved an objective of the DOE Generation IV (GenIV) roadmap for GFR Fuel—namely the demonstration of a composite carbide fuel with 30% volume fuel. This near-term accomplishment is even more significant given the expected or possible time frame for implementation of the GFR in the years 2030 -2050 or beyond.

Knight, Travis W

2010-01-31T23:59:59.000Z

266

Fracture mechanics investigations on high-temperature gas-cooled reactor materials  

SciTech Connect

The prototype nuclear process heat plant and the high-temperature gas-cooled reactor need materials that can withstand temperatures up to 1223 K (950/sup 0/C). An elaboration of fracture mechanics concepts that holds for the complete temperature regime must consider all possible phenomena like creep damage and precipitation during exposure, etc. In tests on the Inconel-617, Hastelloy-X, and Nimonic-86 alloys with respect to fatigue crack growth, creep crack growth, and toughness (J integral R curves) up to 1273 K (1000/sup 0/C), the first creep crack growth results were obtained in helium to compare with the air results. It was shown that pure fatigue crack growth behavior can be described by linear elastic fracture mechanics up to 1273 K. An example of Hastelloy-X at 1223 K proves that evaluating fatigue crack growth according to the J intergral concept gives, within a small scatterband, the same results as by following the linear elastic concept. Hastelloy-X shows a decreasing fracture toughness with increasing temperatures. It is emphasized that the J integral concept holds only if creep deformation can be neglected. The experimental evidence at highest temperatures shows that the J integral R curve is not at all similar to that found at lower temperatures under ideal conditions. Creep crack growth for Nimonic-86 at 1073 less than or equal to T/K less than or equal to 1273 shows that crack growth at 1223 K in helium is found to be larger than in air. Problems arise when correlating the creep crack growth results. The application of the energy rate integral C* seems promising, but this has yet to be proven. A combination of long-term creep with fatigue crack growth is presently impossible.

Krompholz, K.; Bodmann, E.; Gnirss, G.K.; Huthmann, H.

1984-08-01T23:59:59.000Z

267

Turbine airfoil with an internal cooling system having vortex forming turbulators  

DOE Patents (OSTI)

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.

Lee, Ching-Pang

2014-12-30T23:59:59.000Z

268

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

SciTech Connect

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.

J. M. Beck; L. F. Pincock

2011-04-01T23:59:59.000Z

269

Advanced Nuclear Reactor Systems – An Indian Perspective  

Science Journals Connector (OSTI)

The Indian nuclear power programme envisages use of closed nuclear fuel cycle and thorium utilisation as its mainstay for its sustainable growth. The current levels of deployment of nuclear energy in India need to be multiplied nearly hundred fold to reach levels of electricity generation that would facilitate the country to achieve energy independence as well as a developed status. The Indian thorium based nuclear energy systems are being developed to achieve sustainability in respect of fuel resource along with enhanced safety and reduced waste generation. Advanced Heavy Water Reactor and its variants have been designed to meet these objectives. The Indian High Temperature Reactor programme also envisages use of thorium-based fuel with advanced levels of passive safety features.

Ratan Kumar Sinha

2011-01-01T23:59:59.000Z

270

Systems analysis of the CANDU 3 Reactor  

SciTech Connect

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

Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

1993-07-01T23:59:59.000Z

271

Method and system for powering and cooling semiconductor lasers  

DOE Patents (OSTI)

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.

Telford, Steven J; Ladran, Anthony S

2014-02-25T23:59:59.000Z

272

Correcting Aberrations in Complex Magnet Systems for Muon Cooling Channels  

SciTech Connect

Designing and simulating complex magnet systems needed for cooling channels in both neutrino factories and muon colliders requires innovative techniques to correct for both chromatic and spherical aberrations. Optimizing complex systems, such as helical magnets for example, is also difficult but essential. By using COSY INFINITY, a differential algebra based code, the transfer and aberration maps can be examined to discover what critical terms have the greatest influence on these aberrations.

J.A. Maloney, B. Erdelyi, A. Afanaciev, R.P. Johnson, Y.S. Derbenev, V.S. Morozov

2011-03-01T23:59:59.000Z

273

Analysis of advanced solar hybrid desiccant cooling systems for buildings  

SciTech Connect

This report describes an assessment of the energy savings possible from developing hybrid desiccant/vapor-compression air conditioning systems. Recent advances in dehumidifier design for solar desiccant cooling systems have resulted in a dehumidifier with a low pressure drop and high efficiency in heat and mass transfer. A recent study on hybrid desiccant/vapor compression systems showed a 30%-80% savings in resource energy when compared with the best conventional systems with vapor compression. A system consisting of a dehumidifier with vapor compression subsystems in series was found to be the simplest and best overall performer.

Schlepp, D.; Schultz, K.

1984-10-01T23:59:59.000Z

274

Simulation and analysis of district-heating and -cooling systems  

SciTech Connect

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

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

1983-03-01T23:59:59.000Z

275

Development of a system model for advanced small modular reactors.  

SciTech Connect

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.

Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

2014-01-01T23:59:59.000Z

276

Advanced thermionic reactor systems design code  

SciTech Connect

An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance.

Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C. (Department of Nuclear Engineering, Radiation Center, C116, Oregon State University, Corvallis, Oregon 97331-5902 (US))

1991-01-01T23:59:59.000Z

277

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010  

SciTech Connect

The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

Snead, Lance Lewis [ORNL; Besmann, Theodore M [ORNL; Collins, Emory D [ORNL; Bell, Gary L [ORNL

2010-10-01T23:59:59.000Z

278

Weld monitor and failure detector for nuclear reactor system  

DOE Patents (OSTI)

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.

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

1987-01-01T23:59:59.000Z

279

Nuclear reactor with makeup water assist from residual heat removal system  

DOE Patents (OSTI)

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.

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

1993-01-01T23:59:59.000Z

280

Preliminary neutronic studies for the liquid-salt-cooled very hightemperature reactor (LS-VHTR).  

SciTech Connect

Preliminary neutronic studies have been performed in order to provide guidelines to the design of a liquid-salt cooled Very High Temperature Reactor (LS-VHTR) using Li{sub 2}BeF{sub 4} (FLiBe) as coolant and a solid cylindrical core. The studies were done using the lattice codes (WIMS8 and DRAGON) and the linear reactivity model to estimate the core reactivity balance, fuel composition, discharge burnup, and reactivity coefficients. An evaluation of the lattice codes revealed that they give very similar accuracy as the Monte Carlo MCNP4C code for the prediction of the fuel element multiplication factor (kinf) and the double heterogeneity effect of the coated fuel particles in the graphite matrix. The loss of coolant from the LS-VHTR core following coolant voiding was found to result in a positive reactivity addition, due primarily to the removal of the strong neutron absorber Li-6. To mitigate this positive reactivity addition and its impact on reactor design (positive void reactivity coefficient), the lithium in the coolant must be enriched to greater than 99.995% in its Li-7 content. For the reference LS-VHTR considered in this work, it was found that the magnitude of the coolant void reactivity coefficient (CVRC) is quite small (less than $1 for 100% voiding). The coefficient was found to become more negative or less positive with increase in the lithium enrichment (Li-7 content). It was also observed that the coefficient is positive at the beginning of cycle and becomes more negative with increasing burnup, indicating that by using more than one fuel batch, the coefficient could be made negative at the beginning of cycle. It might, however, still be necessary at the beginning of life to design for a negative CVRC value. The study shows that this can be done by using burnable poisons (erbium is a leading candidate) or by changing the reference assembly design (channel dimensions) in order to modify the neutron spectrum. Parametric studies have been performed to attain targeted cycle length of 18 months and discharge burnup greater than 100 GWd/t with a constraint on the uranium enrichment (less than 20% to support non-proliferation goals). The results show that the required uranium enrichment and discharge burnup increase with the number of batches. The three-batch scheme is, however, impractical because the required uranium enrichment is greater than 20%. The required enrichment is smallest for the one-batch case, but its discharge burnup is smaller than the target value. Therefore, the two-batch scheme is desirable to satisfy simultaneously the target cycle length and discharge burnup. It was additionally shown that to increase the core power density to 150% of the reference core value, the required uranium enrichment is less than 20% in the single-batch scheme. This higher power density might not be achievable in the two- or three-batch schemes because the fuel enrichment would exceed 20%.

Kim, T. K.; Taiwo, T. A.; Yang, W. S.

2005-10-05T23:59:59.000Z

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


281

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

SciTech Connect

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)

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

282

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

DOE Patents (OSTI)

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.

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

1982-08-19T23:59:59.000Z

283

A neutron poison tritium breeding controller applied to a water cooled fusion reactor model  

Science Journals Connector (OSTI)

Abstract The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist. This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into the water coolant of a stratified blanket model, depending on the difference between the required tritium excess inventory and the measured tritium excess inventory. The compounds effectively reduce the amount of low energy neutrons available to react with lithium compounds, thus reducing the tritium breeding ratio. This controller reduces the amount of tritium being produced at the start of the reactor's lifetime and increases the rate of tritium production towards the end of its lifetime. Thus, a relatively stable tritium production level may be maintained, allowing the control system to minimize the stored tritium with obvious safety benefits. The FATI code (Fusion Activation and Transport Interface) will be used to perform the tritium breeding and controller calculations.

L.W.G. Morgan; L.W. Packer

2014-01-01T23:59:59.000Z

284

Gas-Cooled Fast Reactor Program. Annual progress report for period ending December 31, 1979  

SciTech Connect

Information on the GCFR reactor is presented concerning the Core Flow Test Loop; shielding and physics; pressure vessel and closure studies; and irradiation program.

Gat, U.; Kasten, P.R.

1980-11-01T23:59:59.000Z

285

CFD Simulation and Analysis of the Combined Evaporative Cooling and Radiant Ceiling Air-conditioning System  

E-Print Network (OSTI)

, and the ceiling cooling system deals with the other part of sensible loads in the air-conditioned zone, so that the condensation on radiant panels and the insufficiency of cooling capacity can be avoided. The cooling water at 18? used in the cooling coils...

Xiang, H.; Yinming, L.; Junmei, W.

2006-01-01T23:59:59.000Z

286

Improving the Efficiency of Your Process Cooling System  

E-Print Network (OSTI)

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...

Baker, R.

2005-01-01T23:59:59.000Z

287

POOL WATER TREATMENT AND COOLING SYSTEM DESCRIPTION DOCUMENT  

SciTech Connect

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.

V. King

2000-06-19T23:59:59.000Z

288

Hybrid Molten Salt Reactor (HMSR): Method and System to fully...  

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

Hybrid Molten Salt Reactor (HMSR): Method and System to fully fission actinides for electric power production without fuel enrichment, fabrication, or reprocessing A method for...

289

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

SciTech Connect

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.

Hikaru Hiruta; Gilles Youinou

2013-09-01T23:59:59.000Z

290

Control of Noise in Power Station Cooling Tower Systems  

Science Journals Connector (OSTI)

Power?station cooling tower systems must handle large volumes of water and air with large potential energy in the water flows and the requirement for large fans. To minimize the noise generated at power station sites use is made of efficient tower fill materials dual low?speed fans (which shifts the spectrum and lowers mid?frequency noise level) and barrier effects in tower location and orientation. Conventional noise control measures such as mufflers are avoided because of the required increase in pressure across the fan and the high initial cost for quieting large towers. The use of natural draft towers is discussed and it is shown that although the low?frequency noise may be reduced the noise levels at typical property line locations are of the same order of magnitude as that for conventional mechanical cooling towers. Since cooling towers at power stations are required as an environmental (thermal) pollution control measure a trade?off between temperature rise of local water supplies versus increases in community noise becomes a critical factor.

Lewis S. Goodfriend

1973-01-01T23:59:59.000Z

291

The Helium Cooling System and Cold Mass Support System for theMICE Coupling Solenoid  

SciTech Connect

The MICE cooling channel consists of alternating threeabsorber focus coil module (AFC) and two RF coupling coil module (RFCC)where the process of muon cooling and reacceleration occurs. The RFCCmodule comprises a superconducting coupling solenoid mounted around fourconventional conducting 201.25 MHz closed RF cavities and producing up to2.2T magnetic field on the centerline. The coupling coil magnetic fieldis to produce a low muon beam beta function in order to keep the beamwithin the RF cavities. The magnet is to be built using commercialniobium titanium MRI conductors and cooled by pulse tube coolers thatproduce 1.5 W of cooling capacity at 4.2 K each. A self-centering supportsystem is applied for the coupling magnet cold mass support, which isdesigned to carry a longitudinal force up to 500 kN. This report willdescribe the updated design for the MICE coupling magnet. The cold masssupport system and helium cooling system are discussed indetail.

Wang, L.; Wu, H.; Li, L.K.; Green, M.A.; Liu, C.S.; Li, L.Y.; Jia, L.X.; Virostek, S.P.

2007-08-27T23:59:59.000Z

292

A review of solar photovoltaic panel cooling systems with special reference to Ground coupled central panel cooling system (GC-CPCS)  

Science Journals Connector (OSTI)

Abstract The efficiency of wafer-based crystalline as well as Thin film Solar photovoltaic cells get reduced with increase of panel temperature. It is noted that the efficiency drops by about 0.5% for increase of 1 °C of panel temperature. It is necessary to operate them at low temperatures in order to keep the PV module electrical efficiency at acceptable level. Therefore need for a low-cost cooling system for the Solar panels is felt. The cooling of Solar PV panels is a problem of great practical significance. It has the potential to reduce the cost of solar energy in three ways. First, cooling can improve the electrical production of standard flat plate PV modules. Second, cooling makes possible the use of concentrating PV systems. Finally, the heat removed by the PV cooling system can be used for domestic or industrial use. Various Solar PV panel cooling systems have been developed in past. A new system for cooling of Solar PV panels called the Ground-Coupled Central Panel Cooling System (GC-CPCS) is installed and operational at the Energy Park of Rajiv Gandhi Proudyogiki Vishwavidyalaya (RGPV), Bhopal, India. This paper discusses various Solar PV panel cooling technologies and the testing and performance of the newly developed Ground-Coupled Central Panel Cooling System (GC-CPCS). This study is unique in three ways. First, it introduces the concept of Central cooling of Solar PV panels. Second, Ground Coupled Heat Exchanger is used for the first time in any Solar PV Panel Cooling System. Third, the testing is done on site under actual solar irradiance.

Amit Sahay; V.K. Sethi; A.C. Tiwari; Mukesh Pandey

2015-01-01T23:59:59.000Z

293

Nuclear reactor fuel rod attachment system  

DOE Patents (OSTI)

A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

Christiansen, David W. (Kennewick, WA)

1982-01-01T23:59:59.000Z

294

Fluid sampling system for a nuclear reactor  

DOE Patents (OSTI)

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.

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

1994-01-01T23:59:59.000Z

295

Optimal Scheduling for Biocide and Heat Exchangers Maintenance Towards Environmentally Friendly Seawater Cooling Systems  

E-Print Network (OSTI)

FOR SEAWATER-COOLED POWER AND DESALINATION PLANTS....................................................... 127 5.1 Overview .............................................................................................. 127 5.2 Introduction... 5.2 Representation of a Once-Thorough Cooling System................................ 141 5.3 An Overall Representation of the Power/Desalination Plant ..................... 152 5.4 The Cooling System for the Case Study...

Binmahfouz, Abdullah

2012-10-19T23:59:59.000Z

296

Preliminary Analysis of a Solar Heat Pump System with Seasonal Storage for Heating and Cooling  

E-Print Network (OSTI)

and cooling were set up, which is responsible for the space heating and cooling and domestic hot water for a residential block. Through hourly simulation, the performance and the economics of such systems were analyzed, for the different tank volumes...

Yu, G.; Chen, P.; Dalenback, J.

2006-01-01T23:59:59.000Z

297

Analysis of recoverable waste heat of circulating cooling water in hot-stamping power system  

Science Journals Connector (OSTI)

This article studies the possibility of using heat pump instead of cooling tower to decrease temperature and recover waste heat of circulating cooling water of power system. Making use of heat transfer theory ......

Panpan Qin; Hui Chen; Lili Chen; Chong Wang…

2013-08-01T23:59:59.000Z

298

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production  

SciTech Connect

The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

2002-01-01T23:59:59.000Z

299

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

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.

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

2005-01-01T23:59:59.000Z

300

Reactor noise analysis applications in NPP I and C systems  

SciTech Connect

Reactor noise analysis techniques are used in many NPPs on a routine basis as 'inspection tools' to get information on the dynamics of reactor processes and their instrumentation in a passive, non-intrusive way. The paper discusses some of the tasks and requirements an NPP has to take to implement and to use the full advantages of reactor noise analysis techniques. Typical signal noise analysis applications developed for the monitoring of the reactor shutdown system and control system instrumentation of the Candu units of Ontario Power Generation and Bruce Power are also presented. (authors)

Gloeckler, O. [International Atomic Energy Agency, Wagramer Strosse 5, A-1400 Vienna, Austria Ontario Power Generation, 230 Westney Road South, Ajax, Ont. L1S 7R3 (Canada)

2006-07-01T23:59:59.000Z

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


301

E-Print Network 3.0 - auxiliary cooling system Sample Search...  

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

Panels Summary: Heating System Preheat - Solar thermal 80-gal tank, electric auxiliary heating Active, indirect forced... -circulation system for cool climates Four solar thermal...

302

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

E-Print Network (OSTI)

be r lea-ed in present pow r reactors. Currently t ere are large st&&ckpiles of depleted uranium. Tt is the oh iective ? therefore, of the bre der reactor. to take this depleted uranium&, use it as a blanket, of fertile material, ccnvert it to a... not he mined, but simply that utility manages nt !)ou'd then h- te an economy c choice as to mine uranium or to use plu tonium an! mate ial from the stockpile of depleted uranium. The doubling rime target is sele. cted by considering th growtl! rate...

Sobey, Thomas Milburn

2012-06-07T23:59:59.000Z

303

Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK{sup TM}  

SciTech Connect

The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK{sup TM} (Simulink, 2004). SIMULINK{sup TM} is a development environment packaged with MatLab{sup TM} (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion components such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK{sup TM} models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK{sup TM} modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator)

Wright, Steven A.; Sanchez, Travis [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

2005-02-06T23:59:59.000Z

304

Tanden Mirror Reactor Systems Code (TMRSC)  

SciTech Connect

This paper describes a computer code developed to model a tandem mirror reactor. This is the first tandem mirror reactor model to couple the highly linked physics, magnetics, and neutronic analysis into a single code. Results from this code for two sensitivity studies are included in this paper. These studies are designed (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power and (2) to determine the impact of reactor power level on cost.

Reid, R.L.; Rothe, K.E.; Barrett, R.J.

1985-01-01T23:59:59.000Z

305

System reliability: An example of nuclear reactor system analysis  

Science Journals Connector (OSTI)

This paper presents an overview of the three reliability investigations of a 900 \\{MWe\\} reactor residual heat removal system. Following a description of the system and its functions, the main procedures used in operational reliability analysis, based on the analysis of occurence records, are covered. The second part of the investigations covers predicted reliability, involving the use of the Markov method. It will be noted that the two types of analysis are in good agreement, the probability of system loss after two months' operation being of the order of 10?1. Additional investigation data are also given.

R. Coudray; J.M. Mattei

1984-01-01T23:59:59.000Z

306

Developing Fully Coupled Dynamical Reactor Core Isolation System Models in RELAP-7 for Extended Station Black-Out Analysis  

SciTech Connect

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup water to the reactor vessel for core cooling when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. It was one of the very few safety systems still available during the Fukushima Daiichi accidents after the tsunamis hit the plants and the system successfully delayed the core meltdown for a few days for unit 2 & 3. Therefore, detailed models for RCIC system components are indispensable to understand extended station black-out accidents (SBO) for BWRs. As part of the effort to develop the new generation reactor system safety analysis code RELAP-7, major components to simulate the RCIC system have been developed. This paper describes the models for those components such as turbine, pump, and wet well. Selected individual component test simulations and a simplified SBO simulation up to but before core damage is presented. The successful implementation of the simplified RCIC and wet well models paves the way to further improve the models for safety analysis by including more detailed physical processes in the near future.

Haihua Zhao; Ling Zou; Hongbin Zhang; David Andrs; Richard Martineau

2014-04-01T23:59:59.000Z

307

Mechanical characterization of metallic materials for high-temperature gas-cooled reactors in air and in helium environments  

SciTech Connect

In the French R and D program for high-temperature gas-cooled reactors (HTGRs), three metallic alloys were studied: steel Chromesco-3 with 2.25% chromium, alloy 800H, and Hastelloy-X. The Chromesco-3 and alloy 800H creep behavior is the same in air and in HTGR atmosphere (helium). The tensile tests of Hastelloy-X specimens reveal that aging has embrittlement and hardening effects up to 700/sup 0/C, but the creep tests at 800/sup 0/C show opposite effects. This particular behavior could be due to induced precipitation by aging and the depletion of hardening elements from the matrix. Tests show a low influence of cobalt content on mechanical properties of Hastelloy-X.

Sainfort, G.; Cappelaere, M.; Gregoire, J.; Sannier, J.

1984-07-01T23:59:59.000Z

308

Changes in the mechanical properties of Hastelloy X when exposed to a typical gas-cooled reactor environment  

SciTech Connect

The helium used in a gas-cooled reactor will contain small amounts of H/sub 2/, CO, CH/sub 4/, H/sub 2/O, and N/sub 2/ which can lead to oxidation and carburization/decarburization of structural materials. Long-term creep tests were run on Hastelloy X to 30,000 h at 649 to 871/sup 0/C. It was found that extensive carburization occurred, the minimum creep rate and time to rupture were equal in air and impure helium environments, and the fracture strain was less in helium than in air. Thermal exposure in the temperature range of 538 to 871/sup 0/C resulted in the reduction of ductility in impact and tensile tests at ambient temperature, and this reduction was greater when the exposure was in impure helium rather than in air. A modified alloy with lower chromium and 2% titanium resisted carburization.

McCoy, H.E. Jr.

1981-01-01T23:59:59.000Z

309

Cleaning residual NaK in the fast flux test facility fuel storage cooling system  

SciTech Connect

The Fast Flux Test Facility (FFTF), located on the U.S. Department of Energy's Hanford Reservation, is a liquid metal-cooled test reactor. The FFTF was constructed to support the U.S. Liquid Metal Fast Breeder Reactor Program. The bulk of the alkali metal (sodium and NaK) has been drained and will be stored onsite prior to final disposition. Residual NaK needed to be removed from the pipes, pumps, heat exchangers, tanks, and vessels in the Fuel Storage Facility (FSF) cooling system. The cooling system was drained in 2004 leaving residual NaK in the pipes and equipment. The estimated residual NaK volume was 76 liters in the storage tank, 1.9 liters in the expansion tank, and 19-39 liters in the heat transfer loop. The residual NaK volume in the remainder of the system was expected to be very small, consisting of films, droplets, and very small pools. The NaK in the FSF Cooling System was not radiologically contaminated. The portions of the cooling system to be cleaned were divided into four groups: 1. The storage tank, filter, pump, and associated piping; 2. The heat exchanger, expansion tank, and associated piping; 3. Argon supply piping; 4. In-vessel heat transfer loop. The cleaning was contracted to Creative Engineers, Inc. (CEI) and they used their superheated steam process to clean the cooling system. It has been concluded that during the modification activities (prior to CEI coming onsite) to prepare the NaK Cooling System for cleaning, tank T-914 was pressurized relative to the In-Vessel NaK Cooler and NaK was pushed from the tank back into the Cooler and that on November 6, 2005, when the gas purge through the In-Vessel NaK Cooler was increased from 141.6 slm to 283.2 slm, NaK was forced from the In-Vessel NaK Cooler and it contacted water in the vent line and/or scrubber. The gases from the reaction then traveled back through the vent line coating the internal surface of the vent line with NaK and NaK reaction products. The hot gases also exited the scrubber through the stack and due to the temperature of the gas, the hydrogen auto ignited when it mixed with the oxygen in the air. There was no damage to equipment, no injuries, and no significant release of hazardous material. Even though the FSF Cooling System is the only system at FFTF that contains residual NaK, there are lessons to be learned from this event that can be applied to future residual sodium removal activities. The lessons learned are: - Before cleaning equipment containing residual alkali metal the volume of alkali metal in the equipment should be minimized to the extent practical. As much as possible, reconfirm the amount and location of the alkali metal immediately prior to cleaning, especially if additional evolutions have been performed or significant time has passed. This is especially true for small diameter pipe (<20.3 centimeters diameter) that is being cleaned in place since gas flow is more likely to move the alkali metal. Potential confirmation methods could include visual inspection (difficult in all-metal systems), nondestructive examination (e.g., ultrasonic measurements) and repeating previous evolutions used to drain the system. Also, expect to find alkali metal in places it would not reasonably be expected to be. - Staff with an intimate knowledge of the plant equipment and the bulk alkali metal draining activities is critical to being able to confirm the amount and locations of the alkali metal residuals and to safely clean the residuals. - Minimize the potential for movement of alkali metal during cleaning or limit the distance and locations into which alkali metal can move. - Recognize that when working with alkali metal reactions, occasional pops and bangs are to be anticipated. - Pre-plan emergency responses to unplanned events to assure responses planned for an operating reactor are appropriate for the deactivation phase.

Burke, T.M.; Church, W.R. [Fluor Hanford, PO Box 1000, Richland, Washington, 99352 (United States); Hodgson, K.M. [Fluor Government Group, PO Box 1050, Richland, Washington, 99352 (United States)

2008-01-15T23:59:59.000Z

310

IEEE TRANSACTIONS ON APPLIED SUPERCONDUCTIVITY, VOL. 14, NO. 2, JUNE 2004 883 Design and Cooling Characteristic Results  

E-Print Network (OSTI)

Characteristic Results of Cryogenic System for 6.6 kV/200 A Inductive Fault Current Limiter Hyoungku Kang, Min-cooled cryogenic cooling system for 1.2 kV/80 A inductive Superconducting Fault Current Limiter (SFCL.830309 TABLE I CHARACTERISTICS OF HTS TAPES tics of the sub-cooled nitrogen cooling system for DC reactor of 6

Chang, Ho-Myung

311

Investigation of plant control strategies for the supercritical C0{sub 2}Brayton cycle for a sodium-cooled fast reactor using the plant dynamics code.  

SciTech Connect

The development of a control strategy for the supercritical CO{sub 2} (S-CO{sub 2}) Brayton cycle has been extended to the investigation of alternate control strategies for a Sodium-Cooled Fast Reactor (SFR) nuclear power plant incorporating a S-CO{sub 2} Brayton cycle power converter. The SFR assumed is the 400 MWe (1000 MWt) ABR-1000 preconceptual design incorporating metallic fuel. Three alternative idealized schemes for controlling the reactor side of the plant in combination with the existing automatic control strategy for the S-CO{sub 2} Brayton cycle are explored using the ANL Plant Dynamics Code together with the SAS4A/SASSYS-1 Liquid Metal Reactor (LMR) Analysis Code System coupled together using the iterative coupling formulation previously developed and implemented into the Plant Dynamics Code. The first option assumes that the reactor side can be ideally controlled through movement of control rods and changing the speeds of both the primary and intermediate coolant system sodium pumps such that the intermediate sodium flow rate and inlet temperature to the sodium-to-CO{sub 2} heat exchanger (RHX) remain unvarying while the intermediate sodium outlet temperature changes as the load demand from the electric grid changes and the S-CO{sub 2} cycle conditions adjust according to the S-CO{sub 2} cycle control strategy. For this option, the reactor plant follows an assumed change in load demand from 100 to 0 % nominal at 5 % reduction per minute in a suitable fashion. The second option allows the reactor core power and primary and intermediate coolant system sodium pump flow rates to change autonomously in response to the strong reactivity feedbacks of the metallic fueled core and assumed constant pump torques representing unchanging output from the pump electric motors. The plant behavior to the assumed load demand reduction is surprising close to that calculated for the first option. The only negative result observed is a slight increase in the intermediate inlet sodium temperatures by about 10 C. This temperature rise could presumably be precluded or significantly reduced through fine adjustment of the control rods and pump motors. The third option assumes that the reactor core power and primary and intermediate system flow rates are ideally reduced linearly in a programmed fashion that instantaneously matches the prescribed load demand. The calculated behavior of this idealized case reveals a number of difficulties because the control strategy for the S-CO{sub 2} cycle overcools the reactor potentially resulting in the calculation of sodium bulk freezing and the onset of sodium boiling. The results show that autonomous SFR operation may be viable for the particular assumed load change transient and deserves further investigation for other transients and postulated accidents.

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

2011-04-12T23:59:59.000Z

312

Testing of Passive Safety System Performance for Higher Power Advanced Reactors  

SciTech Connect

This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

2004-12-31T23:59:59.000Z

313

Safety margins in zircaloy oxidation and embrittlement criteria for emergency core cooling system acceptance  

SciTech Connect

Current emergency core cooling system acceptance criteria for light water reactors specify that, under loss-of-coolant accident (LOCA) conditions, the Baker-Just (BJ) correlation must be used to calculate Zircaloy-steam oxidation, calculated peak cladding temperatures (PCT) must not exceed 1204/sup 0/C, and calculated oxidation must not exceed 17% equivalent cladding reacted (ECR). An appropriately defined minimum margin of safety was estimated for each of these criteria. The currently required BJ oxidation correlation provides margins only over the 1100 to 1500/sup 0/C temperature range at the 95% confidence level. The PCT margins for thermal shock and handling failures are adequate at oxidation temperatures above 1204/sup 0/C for up to 210 and 160 s, respectively, at the 95% confidence level. The ECR thermal shock and handling margins at the 50 and 95% confidence levels, respectively, range between 2 and 7% ECR for the BJ correlation, but vanish at temperatures above 1100 to 1160/sup 0/C for the best-estimate Cathcart-Pawel correlation. However, use of the Cathcart Pawel correlation for ''design basis'' LOCA calculations can be justified at the 85 to 88% confidence level if cooling rate effects can be neglected.

Williford, R.E.

1986-09-01T23:59:59.000Z

314

Graphite-moderated, gas-cooled, and water-moderated, water-cooled reactors as power units in nuclearelectric power stations  

Science Journals Connector (OSTI)

The present article reviews a number of papers submitted at the Second International Conference on the Peaceful Uses of Atomic Energy bearing on water-cooled, water-moderated, graphite-moderated, and gas-coole...

Yu. I. Koryakin

1960-11-01T23:59:59.000Z

315

Thermochemically recuperated and steam cooled gas turbine system  

DOE Patents (OSTI)

A gas turbine system in which the expanded gas from the turbine section is used to generate the steam in a heat recovery steam generator and to heat a mixture of gaseous hydrocarbon fuel and the steam in a reformer. The reformer converts the hydrocarbon gas to hydrogen and carbon monoxide for combustion in a combustor. A portion of the steam from the heat recovery steam generator is used to cool components, such as the stationary vanes, in the turbine section, thereby superheating the steam. The superheated steam is mixed into the hydrocarbon gas upstream of the reformer, thereby eliminating the need to raise the temperature of the expanded gas discharged from the turbine section in order to achieve effective conversion of the hydrocarbon gas.

Viscovich, Paul W. (Longwood, FL); Bannister, Ronald L. (Winter Springs, FL)

1995-01-01T23:59:59.000Z

316

Thermochemically recuperated and steam cooled gas turbine system  

DOE Patents (OSTI)

A gas turbine system is described in which the expanded gas from the turbine section is used to generate the steam in a heat recovery steam generator and to heat a mixture of gaseous hydrocarbon fuel and the steam in a reformer. The reformer converts the hydrocarbon gas to hydrogen and carbon monoxide for combustion in a combustor. A portion of the steam from the heat recovery steam generator is used to cool components, such as the stationary vanes, in the turbine section, thereby superheating the steam. The superheated steam is mixed into the hydrocarbon gas upstream of the reformer, thereby eliminating the need to raise the temperature of the expanded gas discharged from the turbine section in order to achieve effective conversion of the hydrocarbon gas. 4 figs.

Viscovich, P.W.; Bannister, R.L.

1995-07-11T23:59:59.000Z

317

[Gas cooled fuel cell systems technology development program  

SciTech Connect

Objective is the development of a gas-cooled phosphoric acid fuel cell for electric utility power plant application. Primary objectives are to: demonstrate performance endurance in 10-cell stacks at 70 psia, 190 C, and 267 mA/cm[sup 2]; improve cell degradation rate to less than 8 mV/1000 hours; develop cost effective criteria, processes, and design configurations for stack components; design multiple stack unit and a single 100 kW fuel cell stack; design a 375 kW fuel cell module and demonstrate average cell beginning-of-use performance; manufacture four 375-kW fuel cell modules and establish characteristics of 1.5 MW pilot power plant. The work is broken into program management, systems engineering, fuel cell development and test, facilities development.

Not Available

1988-03-01T23:59:59.000Z

318

Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors  

SciTech Connect

The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

Katoh, Yutai [ORNL; Wilson, Dane F [ORNL; Forsberg, Charles W [ORNL

2007-09-01T23:59:59.000Z

319

Generation -IV Reactor Concepts  

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

Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),

320

Cooling system for a bearing of a turbine rotor  

DOE Patents (OSTI)

In a gas turbine, a bore tube assembly radially inwardly of an aft bearing conveys cooling steam to the buckets of the turbine and returns the cooling steam to a return. To cool the bearing and thermally insulate the bearing from the cooling steam paths, a radiation shield is spaced from the bore tube assembly by a dead air gap. Additionally, an air passageway is provided between the radiation shield and the inner surface of an aft shaft forming part of the rotor. Air is supplied from an inlet for flow along the passage and radially outwardly through bores in the aft shaft disk to cool the bearing and insulate it from transfer of heat from the cooling steam.

Schmidt, Mark Christopher (Niskayuna, NY)

2002-01-01T23:59:59.000Z

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


321

Metrology/viewing system for next generation fusion reactors  

SciTech Connect

Next generation fusion reactors require accurate measuring systems to verify sub-millimeter alignment of plasma-facing components in the reactor vessel. A metrology system capable of achieving such accuracy must be compatible with the vessel environment of high gamma radiation, high vacuum, elevated temperature, and magnetic field. This environment requires that the system must be remotely deployed. A coherent, frequency modulated laser radar system is being integrated with a remotely operated deployment system to meet these requirements. The metrology/viewing system consists of a compact laser transceiver optics module which is linked through fiber optics to the laser source and imaging units that are located outside of the harsh environment. The deployment mechanism is a telescopic-mast positioning system. This paper identifies the requirements for the International Thermonuclear Experimental Reactor metrology and viewing system, and describes a remotely operated precision ranging and surface mapping system.

Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Menon, M.M. [Oak Ridge National Lab., TN (United States); Dagher, M.A. [Boeing Rocketdyne Div., Canoga Park, CA (United States)

1997-02-01T23:59:59.000Z

322

Micro- & Nano-Technologies Enabling More Compact, Lightweight Thermoelectric Power Generation & Cooling Systems  

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

Advanced thermoelectric energy recovery and cooling system weight and volume improvements with low-cost microtechnology heat and mass transfer devices are presented

323

Cooling system having reduced mass pin fins for components in a gas turbine engine  

DOE Patents (OSTI)

A cooling system having one or more pin fins with reduced mass for a gas turbine engine is disclosed. The cooling system may include one or more first surfaces defining at least a portion of the cooling system. The pin fin may extend from the surface defining the cooling system and may have a noncircular cross-section taken generally parallel to the surface and at least part of an outer surface of the cross-section forms at least a quartercircle. A downstream side of the pin fin may have a cavity to reduce mass, thereby creating a more efficient turbine airfoil.

Lee, Ching-Pang; Jiang, Nan; Marra, John J

2014-03-11T23:59:59.000Z

324

Modeling the transient operation of an endothermic fuel cooling system for high Mach number vehicle missions.  

E-Print Network (OSTI)

??A computer model was developed to simulate the transient operation of a hypothetical endothermic fuel cooling system. The model simulated the performance of a cross-flow,… (more)

Williams, Mark Robert

2012-01-01T23:59:59.000Z

325

Air-cooled Condensers in Next-generation Conversion Systems  

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

DOE Geothermal Program Peer Review 2010 - Presentation. Project objective: to reduce the costs associated with the generation of electrical power from air-cooled binary plants.

326

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

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

10: Cooling Towers (Revised) (Fact Sheet), Federal Energy Management Program (FEMP) nasa-msfcwatercs2.pdf More Documents & Publications NASA's Marshall Space Flight Center...

327

An economic comparison between two district cooling systems in Halmstad.  

E-Print Network (OSTI)

?? The supply of cooling has increased significantly in recent years, the trend shows that the increase will continue one reason is that the standard… (more)

Le, Alex

2014-01-01T23:59:59.000Z

328

Improving Cooling performance of the mechanical resonator with the two-level-system defects  

E-Print Network (OSTI)

We study cooling performance of a realistic mechanical resonator containing defects. The normal cooling method through an optomechanical system does not work efficiently due to those defects. We show by employing periodical $\\sigma_z$ pulses, we can eliminate the interaction between defects and their surrounded heat baths up to the first order of time. Compared with the cooling performance of no $\\sigma_z$ pulses case, much better cooling results are obtained. Moreover, this pulse sequence has an ability to improve the cooling performance of the resonator with different defects energy gaps and different defects damping rates.

Tian Chen; Xiang-Bin Wang

2014-06-03T23:59:59.000Z

329

Stochastic Cooling  

SciTech Connect

Stochastic Cooling was invented by Simon van der Meer and was demonstrated at the CERN ISR and ICE (Initial Cooling Experiment). Operational systems were developed at Fermilab and CERN. A complete theory of cooling of unbunched beams was developed, and was applied at CERN and Fermilab. Several new and existing rings employ coasting beam cooling. Bunched beam cooling was demonstrated in ICE and has been observed in several rings designed for coasting beam cooling. High energy bunched beams have proven more difficult. Signal suppression was achieved in the Tevatron, though operational cooling was not pursued at Fermilab. Longitudinal cooling was achieved in the RHIC collider. More recently a vertical cooling system in RHIC cooled both transverse dimensions via betatron coupling.

Blaskiewicz, M.

2011-01-01T23:59:59.000Z

330

Safety Issues and Approach to Meet the Safety Requirements in Tokamak Cooling Water System of ITER  

SciTech Connect

The ITER (Latin for 'the way') tokamak cooling water system (TCWS) consists of several separate systems to cool the major ITER components - the divertor/limiter, the first wall blanket, the neutral beam injector and the vacuum vessel. The ex-vessel part of the TCWS systems provides a confinement function for tritium and activated corrosion products in the cooling water. The Vacuum Vessel System also has a functional safety requirement regarding the residual heat removal from in-vessel components. A preliminary hazards assessment (PHA) was performed for a better understanding of the hazards, initiating events, and defense in depth mechanisms associated with the TCWS. The PHA was completed using the following steps. (1) Hazard Identification. Hazards associated with the TCWS were identified including radiological/chemical/electromagnetic hazards and physical hazards (e.g., high voltage, high pressure, high temperature, falling objects). (2) Hazard Categorization. Hazards identified in step (1) were categorized as to their potential for harm to the workers, the public, and/or the environment. (3) Hazard Evaluation. The design was examined to determine initiating events that might occur and that could expose the public, environment, or workers to the hazard. In addition the system was examined to identify barriers that prevent exposure. Finally, consequences to the public or workers were qualitatively assessed, should the initiating event occur and one or more of the barriers fail. Frequency of occurrence of the initiating event and subsequent barrier failure was qualitatively estimated. (4) Accident Analysis. A preliminary hazards analysis was performed on the conceptual design of the TCWS. As the design progresses, a detailed accident analysis will be performed in the form of a failure modes and effects analysis. The results of the PHA indicated that the principal hazards associated with the TCWS were those associated with radiation. These were low compared to hazards associated with nuclear fission reactors and were limited to potential exposure to the on-site workers if appropriate protective actions were not used. However, the risk to the general public off-site was found to be negligible even under worst case accident conditions.

Flanagan, George F [ORNL] [ORNL; Reyes, Susana [ITER Organization, Saint Paul Lez Durance, France] [ITER Organization, Saint Paul Lez Durance, France; Chang, Keun Pack [ITER Organization, Saint Paul Lez Durance, France] [ITER Organization, Saint Paul Lez Durance, France; Berry, Jan [ORNL] [ORNL; Kim, Seokho H [ORNL] [ORNL

2010-01-01T23:59:59.000Z

331

Vertical Pretreatment Reactor System (Poster), NREL (National Renewable Energy Laboratory)  

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

Vertical Pretreatment Reactor System Vertical Pretreatment Reactor System Two-vessel system for primary and secondary pretreatment at diff erent temperatures * Biomass is heated by steam injection to temperatures of 120°C to 210°C in the pressurized mixing tube * Preheated, premixed biomass is retained for specified residence time in vertical holding vessel; material continuously moves by gravity from top to bottom of reactor in plug-fl ow fashion * Residence time is adjusted by changing amount of material held in vertical vessel relative to continuous fl ow of material entering and exiting vessel * Optional additional reactor vessel allows for secondary pretreatment at lower temperatures-120°C to 180°C-with potential to add other chemical catalysts * First vessel can operate at residence

332

The CANDU Reactor System: An Appropriate Technology  

Science Journals Connector (OSTI)

...apparent. The devel-opment of Zircaloy-2 by the U.S. Bettis Laboratory in Pittsburgh, made this con-cept practicable...necessary to base a reactor de-sign on Zircaloy-2 had not Bettis work-ers been testing the material in NRX as part of a collaborative...

J. A. L. Robertson

1978-02-10T23:59:59.000Z

333

Radiant heating and cooling, displacement ventilation with heat recovery and storm water cooling: An environmentally responsible HVAC system  

SciTech Connect

This paper describes the design, operation, and performance of an HVAC system installed as part of a project to demonstrate energy efficiency and environmental responsibility in commercial buildings. The systems installed in the 2180 m{sup 2} office building provide superior air quality and thermal comfort while requiring only half the electrical energy of conventional systems primarily because of the hydronic heating and cooling system. Gas use for the building is higher than expected because of longer operating hours and poor performance of the boiler/absorption chiller.

Carpenter, S.C.; Kokko, J.P. [Enermodal Engineering Ltd., Kitchener, Ontario (Canada)

1998-12-31T23:59:59.000Z

334

Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1980-June 30, 1980  

SciTech Connect

Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described; this includes: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850 and 950/sup 0/C. The initiation of air creep-rupture testing in the intensive screening test program is discussed. In addition, the status of the data management system is described.

Not Available

1980-11-14T23:59:59.000Z

335

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

E-Print Network (OSTI)

8.4.3 Safety . . . . . . . . . . . . .and inherent safety . . . . . . . . . . . . . . . . . . . .II Engineered safety systems 5 Approaches for improving the

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

336

Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, October 1, 1979-December 31, 1979  

SciTech Connect

This report presents the results of work performed from October 1, 1979 through December 31, 1979. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described. This includes: screening creep results, weight gain and post-exposure mechanical properties for materials thermally exposed at 750/sup 0/ and 850/sup 0/C (1382/sup 0/ and 1562/sup 0/F). In addition, the status of the data management system is described.

Not Available

1980-04-18T23:59:59.000Z

337

System aspects of a Space Nuclear Reactor Power System  

SciTech Connect

Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, altitude control, reliability, and survivability. Among the findings are that the stowed length of the power system is important to mission design and that orbital storage for months to years may be needed for missions involving orbital assembly. The power system design evolved during the study and has continued to evolve; the current design differs somewhat from that examined in this paper.

Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Kia, T.; Nesmith, B.

1988-01-01T23:59:59.000Z

338

Cryogenic cooling system of HTS transformers by natural convection of subcooled liquid nitrogen  

E-Print Network (OSTI)

Cryogenic cooling system of HTS transformers by natural convection of subcooled liquid nitrogen Ho-793, South Korea Abstract Heat transfer analysis on a newly proposed cryogenic cooling system is performed, and over-load operation. One of the key techniques to realize these advantages in practice is the cryogenic

Chang, Ho-Myung

339

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 and others. The method could help designers to examine and choose a thermoelectric module from catalogues

340

Solar Cooling Using Variable Geometry Ejectors Centre for Sustainable Energy Systems  

E-Print Network (OSTI)

electrically driven heat pumps to be deployed to locations such as houses where a steady source of heat may pumps were other notable benefits. Heat pump cooling systems dominate the air-conditioning market to consider solar heat driven cooling systems, most prominently since the 1990s. The Ejector Heat Pump

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


341

E-Print Network 3.0 - aerospace system test reactor Sample Search...  

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

Engineering 15 www.eprg.group.cam.ac.uk EPRGWORKINGPAPER Summary: Accelerator-Driven Subcritical Reactor Park Michel-Alexandre Cardin1 Engineering Systems Division... and reactor...

342

Dual annular rotating "windowed" nuclear reflector reactor control system  

DOE Patents (OSTI)

A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

Jacox, Michael G. (Idaho Falls, ID); Drexler, Robert L. (Idaho Falls, ID); Hunt, Robert N. M. (Idaho Falls, ID); Lake, James A. (Idaho Falls, ID)

1994-01-01T23:59:59.000Z

343

Integrated intelligent systems in advanced reactor control rooms  

SciTech Connect

An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs., 5 figs.

Beckmeyer, R.R.

1989-01-01T23:59:59.000Z

344

Control of reactor coolant flow path during reactor decay heat removal  

DOE Patents (OSTI)

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Hunsbedt, Anstein N. (Los Gatos, CA)

1988-01-01T23:59:59.000Z

345

Cooling season study and economic analysis of a desiccant cooling system  

E-Print Network (OSTI)

10 20 30 40 50 60 70 80 Gas Cost (3/GJ) Figure 4. 4 Gas Price vs DINC Cycle Payback Period at Various Electricity Prices SEER = 12 35 20 18 16 ~ 14 ~ 12 D o 10 8 6 o 4 $0. 06/Kwh $0. 09/Kwh $0. 12/Kwh $0. 15/Kwh $0. 'I 8/Kwh 10 20... IV ECONOMIC ANALYSIS V CONCLUSIONS 28 36 NOMENCLATURE 39 REFERENCES 46 APPENDIX A - HOUSE CONSTRUCTION DATA . . APPENDIX B - SECOND LAW COMPARISON 48 53 APPENDIX C - COOLING SEASON AND DINC CYCLE PROGRAM LISTING 72 APPENDIX D - ECONOMIC...

Lee, James Howard

2012-06-07T23:59:59.000Z

346

Decompression cooling system operation for HTS power cable in the KEPCO power grid  

Science Journals Connector (OSTI)

A 3-phase 22.9 kV/50 MVA 410 m HTS power cable system was installed at power grid of KEPCO and had been operated for 20 months. In the HTS cable system an open type cooling system was constructed for cooling LN2 using as coolant for superconducting cable. The cooling capacity of the cooling system was 6 kW at 69 K. Subcooled LN2 flew thorough 410 m HTS cable maintaining 69 K of operating temperature for HTS cable. The electric load had fluctuated continuously with the load status so that the cooling state was also controlled to keep stable operating condition. The consumed LN2 used for making subcooled state was refilled periodically and the amount was 3 tons in average. During all the operating period the HTS cable system supplied electric power stably without any problem.

2014-01-01T23:59:59.000Z

347

Performance Assessment of a Desiccant Cooling System in a CHP Application with an IC Engine  

SciTech Connect

Performance of a desiccant cooling system was evaluated in the context of combined heat and power (CHP). The baseline system incorporated a desiccant dehumidifier, a heat exchanger, an indirect evaporative cooler, and a direct evaporative cooler. The desiccant unit was regenerated through heat recovery from a gas-fired reciprocating internal combustion engine. The system offered sufficient sensible and latent cooling capacities for a wide range of climatic conditions, while allowing influx of outside air in excess of what is typically required for commercial buildings. Energy and water efficiencies of the desiccant cooling system were also evaluated and compared with those of a conventional system. The results of parametric assessments revealed the importance of using a heat exchanger for concurrent desiccant post cooling and regeneration air preheating. These functions resulted in enhancement of both the cooling performance and the thermal efficiency, which are essential for fuel utilization improvement. Two approaches for mixing of the return air and outside air were examined, and their impact on the system cooling performance and thermal efficiency was demonstrated. The scope of the parametric analyses also encompassed the impact of improving the indirect evaporative cooling effectiveness on the overall cooling system performance.

Jalalzadeh-Azar, A. A.; Slayzak, S.; Judkoff, R.; Schaffhauser, T.; DeBlasio, R.

2005-04-01T23:59:59.000Z

348

Potential of Evaporative Cooling Systems for Buildings in India  

E-Print Network (OSTI)

Evaporative cooling potential for building in various climatic zones in India is investigated. Maintainable indoor conditions are obtained from the load - capacity analysis for the prevailing ambient conditions. For the assumed activity level...

Maiya, M. P.; Vijay, S.

2010-01-01T23:59:59.000Z

349

Understanding and reducing energy and costs in industrial cooling systems  

E-Print Network (OSTI)

Industrial cooling remains one of the largest potential areas for electrical energy savings in industrial plants today. This is in spite of a relatively small amount of attention paid to it by energy auditors and rebate program designers. US DOE...

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

2012-01-01T23:59:59.000Z

350

Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors  

SciTech Connect

This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel [1]. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as PuFP. The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the PuFP composition. The FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

Hikaru Hiruta; Gilles Youinou

2013-09-01T23:59:59.000Z

351

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems  

SciTech Connect

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

D. E. Shropshire

2009-01-01T23:59:59.000Z

352

Pressurized reactor system and a method of operating the same  

DOE Patents (OSTI)

A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

Isaksson, Juhani M. (Karhula, FI)

1996-01-01T23:59:59.000Z

353

Pressurized reactor system and a method of operating the same  

DOE Patents (OSTI)

A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Super-atmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gasification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor. 2 figs.

Isaksson, J.M.

1996-06-18T23:59:59.000Z

354

System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor  

SciTech Connect

In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses {approximately}80 W(electric).

Lee, H.H.; Abdul-Hamid, S.; Klein, A.C. [Oregon State Univ., Corvallis, OR (United States). Dept. of Nuclear Engineering Radiation Center] [Oregon State Univ., Corvallis, OR (United States). Dept. of Nuclear Engineering Radiation Center

1996-07-01T23:59:59.000Z

355

DEVELOPMENT OF A SIMULATION CODE FOR A COOL-DOWN PROCESS OF THE CRYOGENIC HYDROGEN SYSTEM  

SciTech Connect

Supercritical hydrogen with a pressure of 1.5 MPa and a temperature of 20 K has been selected as a moderator material in an intense spallation neutron source (JSNS), which is one of main experimental facilities in J-PARC. The cryogenic hydrogen system, in which a hydrogen circulation system is cooled by a helium refrigerator with the refrigeration power of 6.45 kW at 15.5 K, has been designed to provide the supercritical hydrogen to the moderator and to remove the nuclear heating generated there. In this study, we have developed a simulation code that predicts temperature behaviors in the hydrogen circulation system during its cool-down process. Cool-down process analyses have been performed, and an operational method for the cool-down process has been studied. The analytical results indicate that the hydrogen circulation system would be able to be cooled down to 18 K within 19 hours.

Tatsumoto, H.; Aso, T.; Ohtsu, K.; Kato, T.; Futakawa, M. [J-PARC Center, Japan Atomic Energy Agency, Tokai, Ibaraki, 319-1195 (Japan)

2010-04-09T23:59:59.000Z

356

Performance Modeling of a Solar Driven Absorption Cooling System for Carnegie Mellon University's Intelligent Workplace  

E-Print Network (OSTI)

system for space heating and cooling. The proposed energy supply system configuration includes integrated compound parabolic concentrator (ICPC), a hot storage tank, a gas fired auxiliary heater, a steam generator, a steam driven absorption chiller...

Masson, S. V.; Qu, M.; Archer, D. H.

2006-01-01T23:59:59.000Z

357

Conservation of Energy Through The Use of a Predictive Performance Simulator of Operating Cooling Water Systems  

E-Print Network (OSTI)

chemical treatment program for the prevention of corrosion, scale and deposit accumulations. Calgon has made available a computerized performance simulator of operating cooling water systems which reliably predicts system corrosion rates, percent scale...

Schell, C. J.

1981-01-01T23:59:59.000Z

358

Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape  

SciTech Connect

Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed.

DeVault, G.P.; Bell, C.R.

1985-01-01T23:59:59.000Z

359

IMPACTS OF REFRIGERANTLINE LENGTH ON SYSTEM EFFICIENCY IN RESIDENTIAL HEATING AND COOLING SYSTEMS USING REFRIGERANT DISTRIBUTION.  

SciTech Connect

The effects on system efficiency of excess refrigerant line length are calculated for an idealized residential heating and cooling system. By excess line length is meant refrigerant tubing in excess of the 25 R provided for in standard equipment efficiency test methods. The purpose of the calculation is to provide input for a proposed method for evaluating refrigerant distribution system efficiency. A refrigerant distribution system uses refrigerant (instead of ducts or pipes) to carry heat and/or cooling effect from the equipment to the spaces in the building in which it is used. Such systems would include so-called mini-splits as well as more conventional split systems that for one reason or another have the indoor and outdoor coils separated by more than 25 ft. This report performs first-order calculations of the effects on system efficiency, in both the heating and cooling modes, of pressure drops within the refrigerant lines and of heat transfer between the refrigerant lines and the space surrounding them.

ANDREWS, J.W.

2001-04-01T23:59:59.000Z

360

Reactor technology assessment and selection utilizing systems engineering approach  

SciTech Connect

The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

Zolkaffly, Muhammed Zulfakar; Han, Ki-In [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-02-12T23:59:59.000Z

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


361

Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems  

SciTech Connect

Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

2011-04-06T23:59:59.000Z

362

Method of, and apparatus for, killing marine life in and about the cooling system of a marine vehicle  

SciTech Connect

This patent describes a method of killing marine life in the cooling system of a marine vehicle having a water-cooled internal combustion engine. It comprises: providing an enclosure about the inlet and outlet ports such that water circulated through the cooling system will be drawn from, and discharged back into, the enclosure; and operating the engine so as to cause the temperature of the water within the enclosure and the cooling system to increase to a predetermined minimum temperature; thereby to cause marine life in the enclosure and cooling system to be killed by the increased temperature of water within the cooling system and enclosure.

Brockhurst, J.V.

1992-09-22T23:59:59.000Z

363

Reactor protection system with automatic self-testing and diagnostic  

DOE Patents (OSTI)

A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ``identical`` values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs.

Gaubatz, D.C.

1996-12-17T23:59:59.000Z

364

Effect of a Radiant Panel Cooling System on Indoor Air Quality of a Conditioned Space  

E-Print Network (OSTI)

This paper discusses the effect of a radiant cooling panel system on an indoor air quality (IAQ) of a conditioned space. In this study, ceiling radiant cooling panel, mechanical ventilation with fan coil unit (FCU) and 100% fresh air are used...

Mohamed, E.; Abdalla, K. N.

2010-01-01T23:59:59.000Z

365

System and method of active vibration control for an electro-mechanically cooled device  

DOE Patents (OSTI)

A system and method of active vibration control of an electro-mechanically cooled device is disclosed. A cryogenic cooling system is located within an environment. The cooling system is characterized by a vibration transfer function, which requires vibration transfer function coefficients. A vibration controller generates the vibration transfer function coefficients in response to various triggering events. The environments may differ by mounting apparatus, by proximity to vibration generating devices, or by temperature. The triggering event may be powering on the cooling system, reaching an operating temperature, or a reset action. A counterbalance responds to a drive signal generated by the vibration controller, based on the vibration signal and the vibration transfer function, which adjusts vibrations. The method first places a cryogenic cooling system within a first environment and then generates a first set of vibration transfer function coefficients, for a vibration transfer function of the cooling system. Next, the cryogenic cooling system is placed within a second environment and a second set of vibration transfer function coefficients are generated. Then, a counterbalance is driven, based on the vibration transfer function, to reduce vibrations received by a vibration sensitive element.

Lavietes, Anthony D. (Hayward, CA); Mauger, Joseph (Livermore, CA); Anderson, Eric H. (Mountain View, CA)

2000-01-01T23:59:59.000Z

366

Survey and evaluation of available thermal insulation materials for use on solar heating and cooling systems  

SciTech Connect

This is the final report of a survey and evaluation of insulation materials for use with components of solar heating and cooling systems. The survey was performed by mailing questionnaires to manufacturers of insulation materials and by conducting an extensive literature search to obtain data on relevant properties of various types of insulation materials. The study evaluated insulation materials for active and passive solar heating and cooling systems and for multifunction applications. Primary and secondary considerations for selecting insulation materials for various components of solar heating and cooling systems are presented.

Not Available

1980-03-01T23:59:59.000Z

367

Solar assisted cooling with sorption systems: status of the research in Mexico and Latin America  

Science Journals Connector (OSTI)

Solar refrigeration projects both national and international with sorption and other refrigeration systems have been developed in Mexico and other Latin American countries in the last 15 years. A review of the main projects, both for solar cooling and refrigeration and the results obtained are presented in this paper. A methodology where 19 solar technologies for cooling were identified is also presented. Although solar cooling is still not an economically viable technology, the advances made and the experience gained in the projects described and the improved systems envisaged, will make solar refrigeration systems play an important role in the future.

Roberto Best; Isaac Pilatowsky

1998-01-01T23:59:59.000Z

368

Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation  

SciTech Connect

The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

2005-09-27T23:59:59.000Z

369

Proposal for the Award of a Contract for Maintenance, Operation and Minor Installation Work for CERN Cooling and HVAC Systems  

E-Print Network (OSTI)

Proposal for the Award of a Contract for Maintenance, Operation and Minor Installation Work for CERN Cooling and HVAC Systems

1996-01-01T23:59:59.000Z

370

Numerical assessment of liquid cooling system for power electronics in fuel cell electric vehicles  

Science Journals Connector (OSTI)

Abstract Electrical power from the fuel cells is converted and controlled by power electronics that are composed of control units, converters and switching devices. During the power management, the inevitable power losses induce heat generation in the power electronics. In this, effective design for the cooling system is essential in terms of safety, reliability, and durability. A liquid cooling system for the power electronics is applied to chill the electrical components below the thermal specifications. Nonetheless, the layout of cooling components is usually designed after the completion of the chassis and power electronics in the automotive applications, thus, only a little freedom is allowed to change the layout. Thus, it is significant and urgent to investigate the cooling performance before finalizing the layout design. In this paper, one dimensional and computerized fluid dynamics code is employed to simulate the performance of the cooling system at the early stage of conceptual design. Three different layouts of cooling systems are chosen to compare the ensuing systematic cooling performances. The liquid flow rates, pressure drops, and maximum temperatures are computed by the numerical simulations of the cooling system which comprises the cold plates, liquid pump, radiator, and plumbing network. It is demonstrated that for a fuel cell electric vehicle of 100 kW, the dual cooling loops with a specified array control the maximum temperatures below thermal specification by inducing the higher liquid flow rate of rate of 33.4 L/min through radiator than 20.0 L/min in a single loop. The proposed systematic numerical simulation provides significant information to determine the layout of the power electronics coupled with the cooling performance at the early stage of conceptual design.

Heesung Park

2014-01-01T23:59:59.000Z

371

A Free Cooling Based Chilled Water System at Kingston  

E-Print Network (OSTI)

-04-13 Proceedings from the Sixth Annual Industrial Energy Technology Conference Volume I, Houston, TX, April 15-18, 1984 COOLING TOWER #3 FROM EAST FROM WEST TOWER TO TO EAST WES TOWER TOW8:R ELEC. #9 2000 TON MOOE, ?YAP. CHILLING LOAD SHAVING Sl...

Jansen, P. R.

1984-01-01T23:59:59.000Z

372

Designing a 'Near Optimum' Cooling-Water System  

E-Print Network (OSTI)

Cooling water is expensive to circulate. Reducing its flow - i.e., hiking exchanger outlet temperatures - can cut tower, pump and piping investment as much as one-third and operating cost almost in half. Heat-exchanger-network optimization has been...

Crozier, R. A., Jr.

1981-01-01T23:59:59.000Z

373

Passive-solar-cooling system concepts for small office buildings. Final report  

SciTech Connect

This report summarizes the efforts of a small group of building design professionals and energy analysis experts to develop passive solar cooling concepts including first cost estimates for small office buildings. Two design teams were brought together at each of two workshops held in the fall of 1982. Each team included an architect, mechanical engineer, structural engineer, and energy analysis expert. This report presents the passive cooling system concepts resulting from the workshops. It summarizes the design problems, solutions and first-cost estimates relating to each technology considered, and documents the research needs identified by the participants in attempting to implement the various technologies in an actual building design. Each design problem presented at the workshops was based on the reference (base case) small office building analyzed as part of LBL's Cooling Assessment. Chapter II summarizes the thermal performance, physical specifications and estimated first-costs of the base case design developed for this work. Chapters III - VI describe the passive cooling system concepts developed for each technology: beam daylighting; mass with night ventilation; evaporative cooling; and integrated passive cooling systems. The final Chapters, VII and VIII present the preliminary implications for economics of passive cooling technologies (based on review of the design concepts) and recommendations of workshop participants for future research in passive cooling for commercial buildings. Appendices provide backup information on each chapter as indicated.

Whiddon, W.I.; Hart, G.K.

1983-02-01T23:59:59.000Z

374

High-Temperature Gas-Cooled Reactor Steam Cycle/Cogeneration Lead Project strategy plan  

SciTech Connect

The strategy for developing the HTGR system and introducing it into the energy marketplace is based on using the most developed technology path to establish a HTGR-Steam Cycle/Cogeneration (SC/C) Lead Project. Given the status of the HTGR-SC/C technology, a Lead Plant could be completed and operational by the mid 1990s. While there is remaining design and technology development that must be accomplished to fulfill technical and licensing requirements for a Lead Project commitment, the major barriers to the realization a HTGR-SC/C Lead Project are institutional in nature, e.g. Project organization and management, vendor/supplier development, cost/risk sharing between the public and private sector, and Project financing. These problems are further exacerbated by the overall pervading issues of economic and regulatory instability that presently confront the utility and nuclear industries. This document addresses the major institutional issues associated with the HTGR-SC/C Lead Project and provides a starting point for discussions between prospective Lead Project participants toward the realization of such a Project.

None

1982-03-01T23:59:59.000Z

375

A next-generation reactor concept: The Integral Fast Reactor (IFR)  

SciTech Connect

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

Chang, Y.I.

1992-01-01T23:59:59.000Z

376

A next-generation reactor concept: The Integral Fast Reactor (IFR)  

SciTech Connect

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

Chang, Y.I.

1992-07-01T23:59:59.000Z

377

Horizontal Pretreatment Reactor System (Poster), NREL (National Renewable Energy Laboratory)  

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

Diff Diff erent pretreatment chemistry/ residence time combinations are possible using these multiple horizontal-tube reactors * Each tube is indirectly and directly steam heated to temperatures of 150 0 C to 210 0 C * Residence time is varied by changing the speed of the auger that moves the biomass through each tube reactor * Tubes are used individually or in combination to achieve diff erent pretreatment residence times * Smaller tubes made from Hastelloy, an acid-resistant material, are used with more corrosive chemicals and residence times from 3 to 20 minutes * Larger tubes made from 316 stainless steel are used for residence times from 20 to 120 minutes Horizontal Pretreatment Reactor System Versatile pretreatment system for a wide range of pretreatment chemistries

378

Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design  

SciTech Connect

The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

Smith, C

2010-02-22T23:59:59.000Z

379

Microsoft Word - DOE-ID-11-002 DOE Direct cooling system [1].doc  

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

2 2 SECTION A. Project Title: Cooling System for Substation Bldg CPP-613 SECTION B. Project Description The scope of work includes the purchase and installation of an Energy Star compliant 208V three phase staged cooling system capable of maintaining CPP-613 at a temperature below 85 degrees F. The system shall be designed to operate at an elevation of 5000 feet with outside environmental temperatures ranging from -20°F to 100°F. The cooling system shall be pad mounted on the east side of the building between the two cable feeds. The concrete pad will be provided by DOE. The scope of work includes purchasing an appropriately sized cooling system, placing the unit on the concrete pad, making all necessary wall penetrations into the building, installing ductwork and air handlers inside the building, and installing a

380

High Temperature Water Heat Pipes Radiator for a Brayton Space Reactor Power System  

SciTech Connect

A high temperature water heat pipes radiator design is developed for a space power system with a sectored gas-cooled reactor and three Closed Brayton Cycle (CBC) engines, for avoidance of single point failures in reactor cooling and energy conversion and rejection. The CBC engines operate at turbine inlet and exit temperatures of 1144 K and 952 K. They have a net efficiency of 19.4% and each provides 30.5 kWe of net electrical power to the load. A He-Xe gas mixture serves as the turbine working fluid and cools the reactor core, entering at 904 K and exiting at 1149 K. Each CBC loop is coupled to a reactor sector, which is neutronically and thermally coupled, but hydraulically decoupled to the other two sectors, and to a NaK-78 secondary loop with two water heat pipes radiator panels. The segmented panels each consist of a forward fixed segment and two rear deployable segments, operating hydraulically in parallel. The deployed radiator has an effective surface area of 203 m2, and when the rear segments are folded, the stowed power system fits in the launch bay of the DELTA-IV Heavy launch vehicle. For enhanced reliability, the water heat pipes operate below 50% of their wicking limit; the sonic limit is not a concern because of the water, high vapor pressure at the temperatures of interest (384 - 491 K). The rejected power by the radiator peaks when the ratio of the lengths of evaporator sections of the longest and shortest heat pipes is the same as that of the major and minor widths of the segments. The shortest and hottest heat pipes in the rear segments operate at 491 K and 2.24 MPa, and each rejects 154 W. The longest heat pipes operate cooler (427 K and 0.52 MPa) and because they are 69% longer, reject more power (200 W each). The longest and hottest heat pipes in the forward segments reject the largest power (320 W each) while operating at {approx} 46% of capillary limit. The vapor temperature and pressure in these heat pipes are 485 K and 1.97 MPa. By contrast, the shortest water heat pipes in the forward segments operate much cooler (427 K and 0.52 MPa), and reject a much lower power of 45 W each. The radiator with six fixed and 12 rear deployable segments rejects a total of 324 kWth, weights 994 kg and has an average specific power of 326 Wth/kg and a specific mass of 5.88 kg/m2.

El-Genk, Mohamed S.; Tournier, Jean-Michel [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM 87131 (United States); Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM 87131 (United States)

2006-01-20T23:59:59.000Z

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


381

Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor  

Science Journals Connector (OSTI)

The 500 MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam–Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1–0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than ?1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using ‘CRAFT’ software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by ? factor method. The reliability target for SGDHRS arrived from DHR reliability requirement and the ultimate number of demands per year (7/y) on SGDHRS is that the failure frequency should be ?1.4E-8/de. Since it is found from the analysis that the unreliability of SGDHRS with identical loops is 5.2E-6/de and dominated by leak rates of components like AHX, DHX and sodium dump and isolation valves, options with diversity measures in important components were studied. The failure probability of SGDHRS for a design consisting of 2 types of diverse loops (Diverse AHX, DHX and sodium dump and isolation valves) is 2.1E-8/de, which practically meets the reliability requirement.

A. John Arul; C. Senthil Kumar; S. Athmalingam; Om Pal Singh; K. Suryaprakasa Rao

2006-01-01T23:59:59.000Z

382

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

SciTech Connect

An appraisal of the potential performance of different Low Energy Cooling (LEC) systems in nonresidential buildings in California is being conducted using computer simulation. The paper presents results from the first phase of the study, which addressed the systems that can be modeled, with the DOE-2.1E simulation program. The following LEC technologies were simulated as variants of a conventional variable-air-volume system with vapor compression cooling and mixing ventilation in the occupied spaces: Air-side indirect and indirect/direct evaporative pre-cooling. Cool beams. Displacement ventilation. Results are presented for four populous climates, represented by Oakland, Sacramento, Pasadena and San Diego. The greatest energy savings are obtained from a combination of displacement ventilation and air-side indirect/direct evaporative pre-cooling. Cool beam systems have the lowest peak demand but do not reduce energy consumption significantly because the reduction in fan energy is offse t by a reduction in air-side free cooling. Overall, the results indicate significant opportunities for LEC technologies to reduce energy consumption and demand in nonresidential new construction and retrofit.

Bourassa, Norman; Haves, Philip; Huang, Joe

2002-05-17T23:59:59.000Z

383

Numerical simulation of an innovated building cooling system with combination of solar chimney and water spraying system  

Science Journals Connector (OSTI)

In this study, passive cooling of a room using a solar chimney and water spraying system in the room ... a hot and arid city with very high solar radiation). The performance of this system ... some parameters suc...

Ramin Rabani; Ahmadreza K. Faghih; Mehrdad Rabani; Mehran Rabani

2014-11-01T23:59:59.000Z

384

Retro-Commissioning and Improvement for District Heating and Cooling System Using Simulation  

E-Print Network (OSTI)

In order to improve the energy performance of a district heating and cooling (DHC) system, retro-commissioning was analyzed using visualization method and simulation based on mathematical models, and improved operation schemes were proposed...

Shingu, H.; Nakajima, R.; Yoshida, H.; Wang, F.

2006-01-01T23:59:59.000Z

385

Experimental Study of the Floor Radiant Cooling System Combined with Displacement Ventilation  

E-Print Network (OSTI)

ICEBO2006, Shenzhen, China HVAC Technologies for Energy Efficiency, Vol. IV-11-4 Experimental Study of the Floor Radiant Cooling System Combined with Displacement Ventilation Yanli Ren1, Deying Li2, Yufeng Zhang1 1...

Ren, Y.; Li, D.; Zhang, Y.

2006-01-01T23:59:59.000Z

386

Solar Energy to Drive Absorption Cooling Systems Suitable for Small Building Applications  

E-Print Network (OSTI)

results and an overview of the performance of low capacity single stage and half-effect absorption cooling systems, suitable for residential and small building applications. The primary heat source is solar energy supplied from flat plate collectors...

Gomri, R.

2010-01-01T23:59:59.000Z

387

MIT Electric Vehicle Team Porsche designing a cooling system for the AC24 electric motor  

E-Print Network (OSTI)

In this thesis I worked on the design and analysis of a cooling system for the electric motor of the MIT Electric Vehicle Team's Porsche 914 Battery Electric Vehicle. The vehicle's Azure Dynamics AC24 motor tended to ...

Meenen, Jordan N

2010-01-01T23:59:59.000Z

388

The Operation Management and Energy Consumption Analysis of the District Cooling System  

E-Print Network (OSTI)

Based on the investigation of the district cooling system of the Zhongguancun Square in Beijing, we thoroughly analyzed the process of its operation management and the main factors that affect energy consumption. The basis was provided...

Xu, Q.; Li, D.; Xu, W.

2006-01-01T23:59:59.000Z

389

Energy Efficiency Evaluation of Refrigeration Technologies in Combined Cooling, Heating and Power Systems  

E-Print Network (OSTI)

With development of absorption refrigeration technology, the cooling requirement can be met using various optional refrigeration technologies in a CCHP system, including compression refrigeration, steam double-effect absorption refrigeration, steam...

Zuo, Z.; Hu, W.

2006-01-01T23:59:59.000Z

390

Air-Cooled Condensers in Next-Generation Conversion Systems Geothermal...  

Open Energy Info (EERE)

makeup. Though they use no water, air-cooling systems have higher capital costs, reduced power output (heat is rejected at a higher temperature), lower power sales due to higher...

391

Novel Controls for Economic Dispatch of Combined Cooling, Heating and Power (CCHP) Systems  

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

The emergence of technologies that efficiently convert heat into cooling, such as absorption chillers, has opened up many new opportunities and markets for combined heat and power systems. These...

392

Effect of Cooling Flow on the Operation of a Hot Rotor-Gas Foil Bearing System  

E-Print Network (OSTI)

Gas foil bearings (GFBs) operating at high temperature rely on thermal management procedures that supply needed cooling flow streams to keep the bearing and rotor from overheating. Poor thermal management not only makes systems inefficient...

Ryu, Keun

2012-02-14T23:59:59.000Z

393

Air-Cooled Condensers in Next-Generation Conversion Systems Geothermal Lab  

Open Energy Info (EERE)

Air-Cooled Condensers in Next-Generation Conversion Systems Geothermal Lab Air-Cooled Condensers in Next-Generation Conversion Systems Geothermal Lab Call Project Jump to: navigation, search Last modified on July 22, 2011. Project Title Air-Cooled Condensers in Next-Generation Conversion Systems Project Type / Topic 1 Laboratory Call for Submission of Applications for Research, Development and Analysis of Geothermal Technologies Project Type / Topic 2 Air-Cooling Project Description As the geothermal industry moves to use geothermal resources that are more expensive to develop, there will be increased incentive to use more efficient power plants. Because of increasing demand on finite supplies of water, this next generation of more efficient plants will likely need to reject heat sensibly to the ambient (air-cooling). This will be especially true in western states having higher grade Enhanced Geothermal Systems (EGS) resources, as well as most hydrothermal resources. If one had a choice, an evaporative heat rejection system would be selected because it would provide both cost and performance advantages. The evaporative system, however, consumes a significant amount of water during heat rejection that would require makeup. Though they use no water, air-cooling systems have higher capital costs, reduced power output (heat is rejected at a higher temperature), lower power sales due to higher parasitics (fan power), and greater variability in power output (because of large variation in the dry-bulb temperature).

394

Space-reactor electric systems: subsystem technology assessment  

SciTech Connect

This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

Anderson, R.V.; Bost, D.; Determan, W.R.

1983-03-29T23:59:59.000Z

395

Design of a nuclear reactor system for lunar base applications  

E-Print Network (OSTI)

disadvantages. U02 and Pu02 fuels both have extremely poor ther mal conductivities, about 4 W/m K at 500 C, which would normally limit the maximum linear power in the reactor core to unacceptably low levels. For tunately, the ver y high melting temperatur es... conversion, however, high reactor exit temperatures are both necessary and desirable. The efficiency of the power conversion cycle is directly related to the difference between the high and low temperatur es in the system. Since the heat rejection...

Griffith, Richard Odell

2012-06-07T23:59:59.000Z

396

Solar heating and cooling system installed at RKL Controls Company, Lumberton, New Jersey. Final report  

SciTech Connect

Solar heating and cooling of a 40,000 square foot manufacturing building, sales offices and the solar computer control center/display room are described. Information on system description, test data, major problems and resolutions, performance, operation and maintenance manual, manufacturer's literature and as-built drawings are provided also. The solar system is composed of 6000 square feet of Sunworks double glazed flat plate collectors, external above ground storage subsystem, controls, ARKLA absorption chiller, heat recovery and a cooling tower.

None

1981-03-01T23:59:59.000Z

397

An experimental study of external reactor vessel cooling strategy on the critical heat flux using the graphene oxide nano-fluid  

SciTech Connect

External reactor vessel cooling (ERVC) for in-vessel retention (IVR) of corium as a key severe accident management strategy can be achieved by flooding the reactor cavity during a severe accident. In this accident mitigation strategy, the decay heat removal capability depends on whether the imposed heat flux exceeds critical heat flux (CHF). To provide sufficient cooling for high-power reactors such as APR1400, there have been some R and D efforts to use the reactor vessel with micro-porous coating and nano-fluids boiling-induced coating. The dispersion stability of graphene-oxide nano-fluid in the chemical conditions of flooding water that includes boric acid, lithium hydroxide (LiOH) and tri-sodium phosphate (TSP) was checked in terms of surface charge or zeta potential before the CHF experiments. Results showed that graphene-oxide nano-fluids were very stable under ERVC environment. The critical heat flux (CHF) on the reactor vessel external wall was measured using the small scale two-dimensional slide test section. The radius of the curvature is 0.1 m. The dimension of each part in the facility simulated the APR-1400. The heater was designed to produce the different heat flux. The magnitude of heat flux follows the one of the APR-1400 when the severe accident occurred. All tests were conducted under inlet subcooling 10 K. Graphene-oxide nano-fluids (concentration: 10 -4 V%) enhanced CHF limits up to about 20% at mass flux 50 kg/m{sup 2}s and 100 kg/m{sup 2}s in comparison with the results of the distilled water at same test condition. (authors)

Park, S. D.; Lee, S. W.; Kang, S.; Kim, S. M.; Seo, H.; Bang, I. C. [Ulsan National Inst. of Science and Technology UNIST, 100 Banyeon-ri, Eonyang-eup, Ulju-gun, Ulasn Metropolitan City 689-798 (Korea, Republic of)

2012-07-01T23:59:59.000Z

398

Colorado State University program for developing, testing, evaluating and optimizing solar heating and cooling systems  

SciTech Connect

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 as 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--92 fifteen-month period. These include: (1) a project employing isothermal operation of air and liquid solar space heating 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.

Not Available

1992-03-23T23:59:59.000Z

399

Heat exchanger and water tank arrangement for passive cooling system  

DOE Patents (OSTI)

A water storage tank in the coolant water loop of a nuclear reactor contains a tubular heat exchanger. The heat exchanger has tubesheets mounted to the tank connections so that the tubesheets and tubes may be readily inspected and repaired. Preferably, the tubes extend from the tubesheets on a square pitch and then on a rectangular pitch therebetween. Also, the heat exchanger is supported by a frame so that the tank wall is not required to support all of its weight.

Gillett, James E. (Greensburg, PA); Johnson, F. Thomas (Baldwin Boro, PA); Orr, Richard S. (Pittsburgh, PA); Schulz, Terry L. (Murrysville Boro, PA)

1993-01-01T23:59:59.000Z

400

Assessment of adsorber bed designs in waste-heat driven adsorption cooling systems for vehicle air conditioning and refrigeration  

E-Print Network (OSTI)

Assessment of adsorber bed designs in waste-heat driven adsorption cooling systems for vehicle air conditioning Finned tube adsorber bed Specific cooling power Adsorber bed to adsorbent mass ratio a b s t r a c t Adsorber bed design strongly affects the performance of waste-heat driven adsorption cooling systems (ACS

Bahrami, Majid

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


401

Operation of staged membrane oxidation reactor systems  

SciTech Connect

A method of operating a multi-stage ion transport membrane oxidation system. The method comprises providing a multi-stage ion transport membrane oxidation system with at least a first membrane oxidation stage and a second membrane oxidation stage, operating the ion transport membrane oxidation system at operating conditions including a characteristic temperature of the first membrane oxidation stage and a characteristic temperature of the second membrane oxidation stage; and controlling the production capacity and/or the product quality by changing the characteristic temperature of the first membrane oxidation stage and/or changing the characteristic temperature of the second membrane oxidation stage.

Repasky, John Michael

2012-10-16T23:59:59.000Z

402

Expert system for online surveillance of nuclear reactor coolant pumps  

DOE Patents (OSTI)

An expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

Gross, Kenny C. (Bolingbrook, IL); Singer, Ralph M. (Naperville, IL); Humenik, Keith E. (Columbia, MD)

1993-01-01T23:59:59.000Z

403

13 - Generation IV reactor designs, operation and fuel cycle  

Science Journals Connector (OSTI)

Abstract: This chapter looks at Generation IV nuclear reactors, such as the very high-temperature reactor (VHTR), the supercritical water reactor (SCWR), the molten salt reactor (MSR), the sodium-cooled fast reactor (SFR), the lead-cooled fast reactor (LFR) and the gas-cooled fast reactor (GFR). Reactor designs and fuel cycles are also described.

N. Cerullo; G. Lomonaco

2012-01-01T23:59:59.000Z

404

Emulsion polymerization of ethylene-vinyl acetate-branched vinyl ester using a pressure reactor system.  

E-Print Network (OSTI)

??A new pressure reactor system was designed to synthesize a novel branched ester-ethylene-vinyl acetate (BEEVA) emulsion polymer. The reactor system was capable of handling pressure… (more)

Tan, Chee Boon

2008-01-01T23:59:59.000Z

405

Propellant feed system of a regeneratively cooled scramjet  

SciTech Connect

An expander cycle for an airframe-integrated hydrogen-fueled scramjet is analyzed to study regenerative cooling characteristics and overall specific impulse. Below Mach 10, the specific impulse and thrust coincide with the reference values. At Mach numbers above 10, a reduction of the specific impulse occurs due to the coolant flow rate requirement, which is accompanied by an increase of thrust. It is shown that the thrust may be increased by injecting excess fuel into the combustor to compensate for the decrease of the specific impulse. 9 refs.

Kanda, Takeshi; Masuya, Goro; Wakamatsu, Yoshio (National Aerospace Laboratory, Kakuda (Japan))

1991-04-01T23:59:59.000Z

406

Heat exchanger and water tank arrangement for passive cooling system  

DOE Patents (OSTI)

A water storage tank in the coolant water loop of a nuclear reactor contains a tubular heat exchanger. The heat exchanger has tube sheets mounted to the tank connections so that the tube sheets and tubes may be readily inspected and repaired. Preferably, the tubes extend from the tube sheets on a square pitch and then on a rectangular pitch there between. Also, the heat exchanger is supported by a frame so that the tank wall is not required to support all of its weight. 6 figures.

Gillett, J.E.; Johnson, F.T.; Orr, R.S.; Schulz, T.L.

1993-11-30T23:59:59.000Z

407

Theoretical study on a novel ammonia–water cogeneration system with adjustable cooling to power ratios  

Science Journals Connector (OSTI)

Abstract A novel ammonia–water cogeneration system with adjustable cooling to power ratios is proposed and investigated. In the combined system, a modified Kalina subcycle and an ammonia absorption cooling subcycle are interconnected by mixers, splitters, absorbers and heat exchangers. The proposed system can adjust its cooling to power ratios from the separate mode without splitting/mixing processes in the two subcycles to the combined operation modes which can produce different ratios of cooling and power. Simulation analysis is conducted to investigate the effects of operation parameter on system performance. The results indicate that the combined system efficiency can reach the maximum values of 37.79% as SR1 (split ratio 1) is equal to 1. Compared with the separate system, the combined efficiency and COP values of the proposed system can increase by 6.6% and 100% with the same heat input, respectively. In addition, the cooling to power ratios of the proposed system can be adjusted in the range of 1.8–3.6 under the given operating conditions.

Zeting Yu; Jitian Han; Hai Liu; Hongxia Zhao

2014-01-01T23:59:59.000Z

408

Asymmetric crystallization during cooling and heating in model glass-forming systems  

E-Print Network (OSTI)

We perform molecular dynamics (MD) simulations of the crystallization process in binary Lennard-Jones systems during heating and cooling to investigate atomic-scale crystallization kinetics in glass-forming materials. For the cooling protocol, we prepared equilibrated liquids above the liquidus temperature $T_l$ and cooled each sample to zero temperature at rate $R_c$. For the heating protocol, we first cooled equilibrated liquids to zero temperature at rate $R_p$ and then heated the samples to temperature $T > T_l$ at rate $R_h$. We measured the critical heating and cooling rates $R_h^*$ and $R_c^*$, below which the systems begin to form a substantial fraction of crystalline clusters during the heating and cooling protocols. We show that $R_h^* > R_c^*$, and that the asymmetry ratio $R_h^*/R_c^*$ includes an intrinsic contribution that increases with the glass-forming ability (GFA) of the system and a preparation-rate dependent contribution that increases strongly as $R_p \\rightarrow R_c^*$ from above. We also show that the predictions from classical nucleation theory (CNT) can qualitatively describe the dependence of the asymmetry ratio on the GFA and preparation rate $R_p$ from the MD simulations and results for the asymmetry ratio measured in Zr- and Au-based bulk metallic glasses (BMG). This work emphasizes the need for and benefits of an improved understanding of crystallization processes in BMGs and other glass-forming systems.

Minglei Wang; Kai Zhang; Zhusong Li; Yanhui Liu; Jan Schroers; Mark D. Shattuck; Corey S. O'Hern

2015-01-09T23:59:59.000Z

409

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

E-Print Network (OSTI)

MSE): ratio of total cooling power to cooling provided, inGenerally, total modular cooling power demand was somewhathigher server loads. The cooling power demand decreased when

Adams, Barbara J

2009-01-01T23:59:59.000Z

410

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

E-Print Network (OSTI)

MSE): ratio of total cooling power to cooling transported,Generally, total modular cooling power demand stabilized atrack) in this study. The cooling power demand decreased when

Xu, TengFang

2009-01-01T23:59:59.000Z

411

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

E-Print Network (OSTI)

of total hydraulic power for cooling to cooling delivered,temperatures, and cooling output power. 6 Test proceduresefficiency defined as power demand per cooling transferred.

Xu, TengFang

2009-01-01T23:59:59.000Z

412

Mechanically Cooled Large-Volume Germanium Detector Systems for Nuclear Explosion Monitoring  

SciTech Connect

Compact maintenance free mechanical cooling systems are being developed to operate large volume (~570 cm3, ~3 kg, 140% or larger) germanium detectors for field applications. We are using a new generation of Stirling-cycle mechanical coolers for operating the very largest volume germanium detectors with absolutely no maintenance or liquid nitrogen requirements. The user will be able to leave these systems unplugged on the shelf until needed. The flip of a switch will bring a system to life in ~1 hour for measurements. The maintenance-free operating lifetime of these detector systems will exceed five years. These features are necessary for remote long-duration liquid-nitrogen free deployment of large-volume germanium gamma-ray detector systems for Nuclear Explosion Monitoring (NEM). The Radionuclide Aerosol Sampler/Analyzer (RASA) will greatly benefit from the availability of such detectors by eliminating the need for liquid nitrogen at RASA sites while still allowing the very largest available germanium detectors to be utilized. These mechanically cooled germanium detector systems being developed here will provide the largest, most sensitive detectors possible for use with the RASA. To provide such systems, the appropriate technical fundamentals are being researched. Mechanical cooling of germanium detectors has historically been a difficult endeavor. The success or failure of mechanically cooled germanium detectors stems from three main technical issues: temperature, vacuum, and vibration. These factors affect one another. There is a particularly crucial relationship between vacuum and temperature. These factors will be experimentally studied both separately and together to insure a solid understanding of the physical limitations each factor places on a practical mechanically cooled germanium detector system for field use. Using this knowledge, a series of mechanically cooled germanium detector prototype systems are being designed and fabricated. Our collaborators at Pacific Northwest National Laboratory (PNNL) will evaluate these detector systems on the bench top and eventually in RASA systems to insure reliable and practical operation.

Hull, Ethan L.; Pehl, Richard H.; Lathrop, James R.; Martin, Gregory N.; Mashburn, R. B.; Miley, Harry S.; Aalseth, Craig E.; Hossbach, Todd W.; Bowyer, Ted W.

2006-09-21T23:59:59.000Z

413

Structural Oil Pan With Integrated Oil Filtration And Cooling System  

DOE Patents (OSTI)

An oil pan for an internal combustion engine includes a body defining a reservoir for collecting engine coolant. The reservoir has a bottom and side walls extending upwardly from the bottom to present a flanged lip through which the oil pan may be mounted to the engine. An oil cooler assembly is housed within the body of the oil pan for cooling lubricant received from the engine. The body includes an oil inlet passage formed integrally therewith for receiving lubricant from the engine and delivering lubricant to the oil cooler. In addition, the body also includes an oil pick up passage formed integrally therewith for providing fluid communication between the reservoir and the engine through the flanged lip.

Freese, V, Charles Edwin (Westland, MI)

2000-05-09T23:59:59.000Z

414

Small Reactor for Deep Space Exploration  

ScienceCinema (OSTI)

This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

None

2014-05-30T23:59:59.000Z

415

Small Reactor for Deep Space Exploration  

SciTech Connect

This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

None

2012-11-29T23:59:59.000Z

416

Systems and methods for dismantling a nuclear reactor  

DOE Patents (OSTI)

Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

2014-10-28T23:59:59.000Z

417

Analysis of simultaneous cooling and heating in supermarket refrigeration systems.  

E-Print Network (OSTI)

?? In this master thesis project, conventional supermarket refrigeration systems using R404A are compared with refrigeration system solutions using natural refrigerants such as carbon dioxide… (more)

Marigny, Johan

2011-01-01T23:59:59.000Z

418

Method for reliability analysis of complex reactor systems. [LMFBR  

SciTech Connect

A method and a computer code for efficient and accurate reliability analyses of complex reactor systems are described and illustrated through an example. The method permits realistic analyses through its ability to accurately model and evaluate instantaneous and average unavailabilities for large systems with dependencies. The component models can include continuously monitored, non-repairable, and periodically tested components which are subject to failures resulting from components which are subject to failures resulting from component demands, stand-by conditions, human errors associated with testing and repair, as well as failures during actual operation. The numerical process used is efficient and allows analysis of general system configurations with arbitrary scheduling of maintenance operations.

Elerath, J.G.; Vaurio, J.K.; Wood, A.P.

1982-01-01T23:59:59.000Z

419

Cryogenic system component development for the fusion experimental reactor at JAERI  

SciTech Connect

The major objective of fusion R and D at the Japan Atomic Energy Research Institute (JAERI) is to construct the Fusion Experimental Reactor (FER) to follow JT-60. The construction of FER inevitably requires development of a large, reliable, and efficient helium liquefier/refrigerator and the more advanced cryogenic technology for cooling superconducting toroidal and poloidal coils. Typical characteristics required for the cryogenic system of FER are 10 to 20 kW at 4 K as one unit, reliability for > 8000 h, a stable pulsed heat load, and high-energy efficiency of > 1/500. In this cryogenic system, the major components such as the helium compressor, turbo-expander, cold circulation pump for supercritical helium, and cold compressor to reduce operating temperature below 4 K should be scaled up to a mass flow rate of > 1000 g/s. For this purpose, JAERI has developed cryogenics since 1980 in accordance with the development program in which the scaling up of the major components mentioned above are involved as well as cooling technology development.

Kato, T.; Kamiya, S.; Tada, E.; Hiyama, T.; Kawano, K.; Shimamoto, S.

1986-01-01T23:59:59.000Z

420

SIMULATION OF A SOLAR ABSORPTION COOLING SYSTEM J.P. Praene*, D. Morau, F. Lucas, F. Garde, H. Boyer  

E-Print Network (OSTI)

collector. A field of these collectors feed a single-effect absorption chiller of 35 kW nominal coolingSIMULATION OF A SOLAR ABSORPTION COOLING SYSTEM J.P. Praene*, D. Morau, F. Lucas, F. Garde, H; accepted: 15 Oct 2007 This paper describes the dynamic modeling of a solar absorption cooling plant

Paris-Sud XI, Université de

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


421

"Self Cooled Recirculating Liquid Metal Plasma Facing Wall System"  

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

Self Cooled Recirculating Liquid Metal Plasma Facing Wall System" Self Cooled Recirculating Liquid Metal Plasma Facing Wall System" Inventor ..--.. Richard P. Majeski Disclosed is a design for a fully axisymmetric, fast flowing liquid lithium plasma facing "wall" or surface which, in its present form, is intended for implementation in a tokamak. The design employs JxB forces to form a free-surface flow along a guide wall at the outer boundary of the plasma. The implementation of the disclosure design includes a system for recirculating the liquid metal within the volume of the toroidal field coils using inductive pumping, an approach wich allows independent energizing of the wall-forming and recirculating pumping systems, cooling of the recirculating liquid using fluid heat exchange with a molten salt,

422

Analysis of combined cooling, heating, and power systems based on source primary energy consumption  

Science Journals Connector (OSTI)

Combined cooling, heating, and power (CCHP) is a cogeneration technology that integrates an absorption chiller to produce cooling, which is sometimes referred to as trigeneration. For building applications, CCHP systems have the advantage to maintain high overall energy efficiency throughout the year. Design and operation of CCHP systems must consider the type and quality of the energy being consumed. Type and magnitude of the on-site energy consumed by a building having separated heating and cooling systems is different than a building having CCHP. Therefore, building energy consumption must be compared using the same reference which is usually the primary energy measured at the source. Site-to-source energy conversion factors can be used to estimate the equivalent source energy from site energy consumption. However, building energy consumption depends on multiple parameters. In this study, mathematical relations are derived to define conditions a CCHP system should operate in order to guarantee primary energy savings.

Nelson Fumo; Louay M. Chamra

2010-01-01T23:59:59.000Z

423

Heating and cooling performance analysis of a ground source heat pump system in Southern Germany  

Science Journals Connector (OSTI)

Abstract This paper examines thermal performance of a ground source heat pump (GSHP) system. The GSHP system was installed in an office building in Nuremberg city of Germany. In order to evaluate system performance the GSHP system has been continuously monitored for 4 years. Heating and cooling performance of the GSHP system is analyzed based on the accumulated data. Major findings of this work include: (1) coefficient of performance (COP) is estimated to be 3.9 for a typical winter day and energy efficiency ratio (EER) is assessed to be 8.0 for a typical summer day. These results indicate that the GSHP system has a higher efficiency for building cooling than building heating. (2) For a long-term period, the seasonal energy efficiency ratio (SEER) of the GSHP system is observed to increase by 8.7% annually, whereas the seasonal COP is decreased by 4.0% over a 4-year period. The heating and cooling performance of the GSHP system migrates in opposite trend is caused by the unevenly distributed heating and cooling load of the building. This phenomenon deserves serious attention in the design of future GSHP systems in order to avoid the reducing of energy efficiency over long-term operation.

Jin Luo; Joachim Rohn; Manfred Bayer; Anna Priess; Lucas Wilkmann; Wei Xiang

2015-01-01T23:59:59.000Z

424

Audit Report - Naval Reactors Information Technology System Development Efforts, IG-0879  

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

Naval Reactors Information Naval Reactors Information Technology System Development Efforts DOE/IG-0879 December 2012 U.S. Department of Energy Office of Inspector General Office of Audits & Inspections Department of Energy Washington, DC 20585 December 21, 2012 MEMORANDUM FOR THE ADMINISTRATOR, NATIONAL NUCLEAR SECURITY ADMINISTRATION FROM: Gregory H. Friedman Inspector General SUBJECT: INFORMATION: Audit Report on the "Naval Reactors Information Technology System Development Efforts" INTRODUCTION AND OBJECTIVE The Naval Reactors Program (Naval Reactors), an organization within the National Nuclear Security Administration, was established to provide the military with safe and reliable nuclear propulsion plants to power warships and submarines. Naval Reactors maintains responsibility

425

Performance evaluation of an active solar cooling system utilizing low cost plastic collectors and an evaporatively-cooled absorption chiller. Final report  

SciTech Connect

During the summer of 1982, air conditioning in Solar House III at Colorado State University was provided by an evaporatively-cooled absorption chiller. The single-effect lithium bromide chiller provided by Arkla Industries is an experimental three-ton unit from which heat is rejected by direct evaporative cooling of the condenser and absorber walls, thereby eliminating the need for a separate cooling tower. Domestic hot water was also provided by use of a double-walled heat exchanger and 300-l (80-gal) hot water tank. For solar heat supply to the cooling system, plastic thin film collectors developed by Brookhaven National Laboratory were installed on the roof of Solar House III. Failure to withstand stagnation temperatures forced replacement of solar energy with an electric heat source. Objectives of the project were: (1) evaluation of system performance over the course of one cooling season in Fort collins, Colorado; (2) optimization of system operation and control; (3) development of a TRNSYS compatible model of the chiller; and (4) determination of cooling system performance in several US climates by use of the model.

Lof, G.O.G.; Westhoff, M.A.; Karaki, S.

1984-02-01T23:59:59.000Z

426

The Gas Reactor Makes a Comeback  

Science Journals Connector (OSTI)

...while the operators of a gas reactor can leave the...ofhis high 699 temperature gas-cooled reactor (HTGR...from the highly pressured turbine drive system might get...would produce combustible gas-es, creating the potential...too much to complete the remaining contracts. So General...

ELIOT MARSHALL

1984-05-18T23:59:59.000Z

427

Efficiency of producing additional power in units of nuclear power stations containing water-cooled-water-moderated reactors  

Science Journals Connector (OSTI)

There is a basic possibility to raise the maximum power of a unit containing the VVÉR-1000 reactor in the course of the fuel charge burn-up and with lowering the coefficient of the energy-release nonuniformity...

R. Z. Aminov; V. A. Khrustalev; A. A. Serdobintsev…

1986-12-01T23:59:59.000Z

428

CFD Analysis of Core Bypass Flow and Crossflow in the Prismatic Very High Temperature Gas-cooled Nuclear Reactor  

E-Print Network (OSTI)

Very High Temperature Rector (VHTR) had been designated as one of those promising reactors for the Next Generation (IV) Nuclear Plant (NGNP). For a prismatic core VHTR, one of the most crucial design considerations is the bypass flow and crossflow...

Wang, Huhu 1985-

2012-12-13T23:59:59.000Z

429

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report  

SciTech Connect

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 of 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.

Mac Donald, Philip Elsworth

2002-06-01T23:59:59.000Z

430

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

E-Print Network (OSTI)

Total Power for Server Power Cooling Module Power (kW) (Cooling is the amount of cooling power removed from the dataratio of total cooling power to the cooling transported by

Xu, TengFang T.

2009-01-01T23:59:59.000Z

431

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, Progress Report for Work Through September 2002, 4th Quarterly Report  

SciTech Connect

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 of 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. The Generation IV Roadmap effort has identified the thermal spectrum SCWR (followed by the fast spectrum SCWR) as one of the advanced concepts that should be developed for future use. Therefore, the work in this NERI project is addressing both types of SCWRs.

Mac Donald, Philip Elsworth

2002-09-01T23:59:59.000Z

432

Adaptive Thermal Management for Portable System Batteries by Forced Convection Cooling  

E-Print Network (OSTI)

Adaptive Thermal Management for Portable System Batteries by Forced Convection Cooling Qing Xie the battery longevity increases. This is the first work that formulates the adaptive thermal management is proposed to derive the ATMB policy. Keywords-- battery system; adaptive thermal management; forced

Pedram, Massoud

433

A Project Management and Systems Engineering Structure for a Generation IV Very High Temperature Reactor  

SciTech Connect

The Very High Temperature Reactor (VHTR) will be an advanced, very high temperature (approximately 1000o C. coolant outlet temperature), gas cooled nuclear reactor and is the nearest term of six Generation IV reactor technologies for nuclear assisted hydrogen production. In 2001, the Generation IV International Forum (GIF), a ten nation international forum working together with the Department of Energy’s (DOE) Nuclear Energy Research Advisory Committee (NERAC), agreed to proceed with the development of a technology roadmap and identified the next generation of nuclear reactor systems for producing new sources of power. Since a new reactor has not been licensed in the United States since the 1970s, the risks are too large for a single utility to assume in the development of an unprecedented Generation IV reactor. The government must sponsor and invest in the research to resolve major first of a kind (FOAK) issues through a full-scale demonstration prior to industry implementation. DOE’s primary mission for the VHTR is to demonstrate nuclear reactor assisted cogeneration of electricity and hydrogen while meeting the Generation IV goals for safety, sustainability, proliferation resistance and physical security and economics. The successful deployment of the VHTR as a demonstration project will aid in restarting the now atrophied U.S. nuclear power industry infrastructure. It is envisioned that VHTR project participants will include DOE Laboratories, industry partners such as designers, constructors, manufacturers, utilities, and Generation IV international countries. To effectively mange R&D, engineering, procurement, construction, and operation for this multi-organizational and technologically complex project, systems engineering will be used extensively to ensure delivery of the final product. Although the VHTR is an unprecedented FOAK system, the R&D, when assessed using the Office of Science and Technology Gate Model, falls primarily in the 3rd - Exploratory Development, 4th – Advanced Development, and 5th- Engineering Development stages of maturity rather than in the basic and viability stages. Therefore the R&D must be controlled and project driven from the top down to address specific issues of feasibility, proof of design or support of engineering. The design evolution must be through the systems approach including an iterative process of high-level requirements definition, engineering to focus R&D to verify feasibility, requirements development and conceptual design, R&D to verify design and refine detailed requirements for final detailed design. This paper will define a framework for project management and application of systems engineering at the Idaho National Engineering and Environmental Laboratory (INEEL). The VHTR Project includes an overall reactor design and construction activity and four major supporting activities: fuel development and qualification, materials selection and qualification, NRC licensing and regulatory support, and the hydrogen production plant.

Ed Gorski; Dennis Harrell; Finis Southworth

2004-09-01T23:59:59.000Z

434

Passive heat transfer means for nuclear reactors  

DOE Patents (OSTI)

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

Burelbach, James P. (Glen Ellyn, IL)

1984-01-01T23:59:59.000Z

435

System for fuel rod removal from a reactor module  

DOE Patents (OSTI)

A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

1988-07-28T23:59:59.000Z

436

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors  

SciTech Connect

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01T23:59:59.000Z

437

Ignition reactor and pump pulse parameters in a reactor–laser system  

Science Journals Connector (OSTI)

The experience gained in operating a demonstration nuclear-pumped laser in stand B (Physics and Power- Engineering Institute (FEI)) with a pulsed ignition reactor based on the 235U BARS-6 reactor is analyzed. It ...

P. P. D’yachenko; G. N. Fokin

2012-09-01T23:59:59.000Z

438

Thermal performance of oil spray cooling system for in-wheel motor in electric vehicles  

Science Journals Connector (OSTI)

Abstract The cooling of the motor in an in-wheel system is critical to its performance and durability. In the present study, the shape of the channel in the hollow shaft for the oil spray cooling of a high-capacity 35 kW in-wheel motor was optimized, and the thermal performance of the motor was evaluated by numerical analysis and experiments. The thermal flow was analyzed by evaluating the thermal performance of two conventional cooling models of in-wheel motors under conditions of continuous rating base speed. For conventional model #1, in which the cooling oil is stagnant in the lower end of the motor, the maximum temperature of the coil was 221.7 °C. For conventional model #2, in which the cooling oil circulates through the exit and entrance of the housing and jig, the maximum temperature of the coil was 155.4 °C. Both models thus proved to be unsuitable for in-wheel motors because the motor control specifications limit the maximum temperature to 150 °C. We designed and manufactured an enhanced model for in-wheel motors, which we equipped with an optimized channel for the oil spray cooling mode, and evaluated its thermal performance under continuous rating conditions. The maximum temperatures of the coil at the base and maximum speeds, which were set as the design points, were below the motor temperature limit, being 138.1 and 137.8 °C, respectively.

Dong Hyun Lim; Sung Chul Kim

2014-01-01T23:59:59.000Z

439

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

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

demand and prices are at their lowest. An energy savings calculator from Oklahoma Gas & Electric suggests the system could save nearly 42,000 a year over conventional...

440

Simulation of solar lithium bromide–water absorption cooling system with parabolic trough collector  

Science Journals Connector (OSTI)

Ahwaz is one of the sweltering cities in Iran where an enormous amount of energy is being consumed to cool residential places in a year. The aim of this research is to simulate a solar single effect lithium bromide–water absorption cooling system in Ahwaz. The solar energy is absorbed by a horizontal N–S parabolic trough collector and stored in an insulated thermal storage tank. The system has been designed to supply the cooling load of a typical house where the cooling load peak is about 17.5 kW (5 tons of refrigeration), which occurs in July. A thermodynamic model has been used to simulate the absorption cycle. The working fluid is water, which is pumped directly to the collector. The results showed that the collector mass flow rate has a negligible effect on the minimum required collector area, but it has a significant effect on the optimum capacity of the storage tank. The minimum required collector area was about 57.6 m2, which could supply the cooling loads for the sunshine hours of the design day for July. The operation of the system has also been considered after sunset by saving solar energy.

M. Mazloumi; M. Naghashzadegan; K. Javaherdeh

2008-01-01T23:59:59.000Z

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441

A comparative assessment of alternative combustion turbine inlet air cooling system  

SciTech Connect

Interest in combustion turbine inlet air cooling (CTAC) has increased during the last few years as electric utilities face increasing demand for peak power. Inlet air cooling increases the generating capacity and decreases the heat rate of a combustion turbine during hot weather when the demand for electricity is generally the greatest. Several CTAC systems have been installed, but the general applicability of the concept and the preference for specific concep