Dynamic modeling of temperature change in outdoor operated tubular photobioreactors.

PubMed

Androga, Dominic Deo; Uyar, Basar; Koku, Harun; Eroglu, Inci

2017-07-01

In this study, a one-dimensional transient model was developed to analyze the temperature variation of tubular photobioreactors operated outdoors and the validity of the model was tested by comparing the predictions of the model with the experimental data. The model included the effects of convection and radiative heat exchange on the reactor temperature throughout the day. The temperatures in the reactors increased with increasing solar radiation and air temperatures, and the predicted reactor temperatures corresponded well to the measured experimental values. The heat transferred to the reactor was mainly through radiation: the radiative heat absorbed by the reactor medium, ground radiation, air radiation, and solar (direct and diffuse) radiation, while heat loss was mainly through the heat transfer to the cooling water and forced convection. The amount of heat transferred by reflected radiation and metabolic activities of the bacteria and pump work was negligible. Counter-current cooling was more effective in controlling reactor temperature than co-current cooling. The model developed identifies major heat transfer mechanisms in outdoor operated tubular photobioreactors, and accurately predicts temperature changes in these systems. This is useful in determining cooling duty under transient conditions and scaling up photobioreactors. The photobioreactor design and the thermal modeling were carried out and experimental results obtained for the case study of photofermentative hydrogen production by Rhodobacter capsulatus, but the approach is applicable to photobiological systems that are to be operated under outdoor conditions with significant cooling demands.

Liquid metal reactor air cooling baffle

DOEpatents

Hunsbedt, Anstein

1994-01-01

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

Liquid metal reactor air cooling baffle

DOEpatents

Hunsbedt, A.

1994-08-16

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

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, R.B.; Fero, A.H.; Sejvar, J.

1997-12-16

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

DOEpatents

Schreiber, Roger B.; Fero, Arnold H.; Sejvar, James

1997-01-01

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

A COOLED NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.; Creutz, E.C.

1960-03-15

A nuclear reactor comprising a pair of graphite blocks separated by an air gap is described. Each of the blocks contains a plurality of channels extending from the gap through the block with a plurality of fuel elements being located in the channels. Means are provided for introducing air into the gap between the graphite blocks and for exhausting the air from the ends of the channels opposite the gap.

Passive heat-transfer means for nuclear reactors. [LMFBR

DOEpatents

Burelbach, J.P.

1982-06-10

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. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

Passive heat transfer means for nuclear reactors

DOEpatents

Burelbach, James P.

1984-01-01

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.

Temperature Swing Adsorption Compressor Development

NASA Technical Reports Server (NTRS)

Finn, John E.; Mulloth, Lila M.; Affleck, Dave L.

2001-01-01

Closing the oxygen loop in an air revitalization system based on four-bed molecular sieve and Sabatier reactor technology requires a vacuum pump-compressor that can take the low-pressure CO, from the 4BMS and compress and store for use by a Sabatier reactor. NASA Ames Research Center proposed a solid-state temperature-swing adsorption (TSA) compressor that appears to meet performance requirements, be quiet and reliable, and consume less power than a comparable mechanical compressor/accumulator combination. Under this task, TSA compressor technology is being advanced through development of a complete prototype system. A liquid-cooled TSA compressor has been partially tested, and the rest of the system is being fabricated. An air-cooled TSA compressor is also being designed.

New reactor cavity cooling system having passive safety features using novel shape for HTGRs and VHTRs

DOE PAGES

Takamatsu, Kuniyoshi; Hu, Rui

2014-11-27

A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features without a requirement for electricity and mechanical drive is proposed for high temperature gas cooled reactors (HTGRs) and very high temperature reactors (VHTRs). The RCCS design consists of continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 (°C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV with radiation and natural convection. Employing the air as the working fluid and the ambient airmore » as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting due to overheating the fuels. The simulation results from a commercial CFD code, STAR-CCM+, show that the temperature distribution of the RCCS is within the temperature limits of the structures, such as the maximum operating temperature of the RPV, 713.15 (K) = 440 (°C), and the heat released from the RPV could be removed safely, even during a loss of coolant accident (LOCA). Finally, when the RCCS can remove 600 (kW) of the rated nominal state even during LOCA, the safety review for building the HTTR could confirm that the temperature distribution of the HTTR is within the temperature limits of the structures to secure structures and fuels after the shutdown because the large heat capacity of the graphite core can absorb heat from the fuel in a short period. Therefore, the capacity of the new RCCS design would be sufficient for decay heat removal.« less

STEAM PLANT, TRA609. SECTION A SHOWS FEATURES OF NORTH/SOUTH AXIS: ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

STEAM PLANT, TRA-609. SECTION A SHOWS FEATURES OF NORTH/SOUTH AXIS: STEAM GENERATOR AND CATWALK, STACK, DEGREASER FEED WATER HEATER IN PENTHOUSE, MEZZANINE, SURGE TANK PIT (BELOW GROUND LEVEL). UTILITY ROOM SHOWS DIESEL ENGINE GENERATORS, AIR TANKS, STARTING AIR COMPRESSORS. OUTSIDE SOUTH END ARE EXHAUST MUFFLER, AIR INTAKE OIL FILTER, RADIATOR COOLING UNIT, AIR SURGE TANK. SECTION B CROSSES WEST TO EAST NEAR SOUTH END OF BUILDING TO SHOW ARRANGEMENT OF DIESEL ENGINE GENERATOR, AIR DRIER, AFTER COOLER, AIR COMPRESSOR, AND BLOWDOWN TANK. BLAW-KNOX 3150-9-2, 6/1950. INL INDEX NO. 431-0609-00-098-100018, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2014 CFR

2014-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2012 CFR

2012-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2010 CFR

2010-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2013 CFR

2013-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

Gas phase oxidation downstream of a catalytic combustor

NASA Technical Reports Server (NTRS)

Tien, J. S.; Anderson, D. N.

1979-01-01

Effect of the length available for gas-phase reactions downstream of the catalytic reactor on the emission of CO and unburned hydrocarbons was investigated. A premixed, prevaporized propane/air feed to a 12/cm/diameter catalytic/reactor test section was used. The catalytic reactor was made of four 2.5 cm long monolithic catalyst elements. Four water cooled gas sampling probes were located at positions between 0 and 22 cm downstream of the catalytic reactor. Measurements of unburned hydrocarbon, CO, and CO2 were made. Tests were performed with an inlet air temperature of 800 K, a reference velocity of 10 m/s, pressures of 3 and 600,000 Pa, and fuel air equivalence ratios of 0.14 to 0.24. For very lean mixtures, hydrocarbon emissions were high and CO continued to be formed downstream of the catalytic reactor. At the highest equivalence ratios tested, hydrocarbon levels were much lower and CO was oxidized to CO2 in the gas phase downstream. To achieve acceptable emissions, a downstream region several times longer than the catalytic reactor could be required.

Model validation using CFD-grade experimental database for NGNP Reactor Cavity Cooling Systems with water and air

DOE Office of Scientific and Technical Information (OSTI.GOV)

Manera, Annalisa; Corradini, Michael; Petrov, Victor

This project has been focused on the experimental and numerical investigations of the water-cooled and air-cooled Reactor Cavity Cooling System (RCCS) designs. At this aim, we have leveraged an existing experimental facility at the University of Wisconsin-Madison (UW), and we have designed and built a separate effect test facility at the University of Michigan. The experimental facility at UW has underwent several upgrades, including the installation of advanced instrumentation (i.e. wire-mesh sensors) built at the University of Michigan. These provides highresolution time-resolved measurements of the void-fraction distribution in the risers of the water-cooled RCCS facility. A phenomenological model has beenmore » developed to assess the water cooled RCCS system stability and determine the root cause behind the oscillatory behavior that occurs under normal two-phase operation. Testing under various perturbations to the water-cooled RCCS facility have resulted in changes in the stability of the integral system. In particular, the effects on stability of inlet orifices, water tank volume have and system pressure been investigated. MELCOR was used as a predictive tool when performing inlet orificing tests and was able to capture the Density Wave Oscillations (DWOs) that occurred upon reaching saturation in the risers. The experimental and numerical results have then been used to provide RCCS design recommendations. The experimental facility built at the University of Michigan was aimed at the investigation of mixing in the upper plenum of the air-cooled RCCS design. The facility has been equipped with state-of-theart high-resolution instrumentation to achieve so-called CFD grade experiments, that can be used for the validation of Computational Fluid Dynanmics (CFD) models, both RANS (Reynold-Averaged) and LES (Large Eddy Simulations). The effect of risers penetration in the upper plenum has been investigated as well.« less

Design Report for the ½ Scale Air-Cooled RCCS Tests in the Natural convection Shutdown heat removal Test Facility (NSTF)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lisowski, D. D.; Farmer, M. T.; Lomperski, S.

The Natural convection Shutdown heat removal Test Facility (NSTF) is a large scale thermal hydraulics test facility that has been built at Argonne National Laboratory (ANL). The facility was constructed in order to carry out highly instrumented experiments that can be used to validate the performance of passive safety systems for advanced reactor designs. The facility has principally been designed for testing of Reactor Cavity Cooling System (RCCS) concepts that rely on natural convection cooling for either air or water-based systems. Standing 25-m in height, the facility is able to supply up to 220 kW at 21 kW/m 2 tomore » accurately simulate the heat fluxes at the walls of a reactor pressure vessel. A suite of nearly 400 data acquisition channels, including a sophisticated fiber optic system for high density temperature measurements, guides test operations and provides data to support scaling analysis and modeling efforts. Measurements of system mass flow rate, air and surface temperatures, heat flux, humidity, and pressure differentials, among others; are part of this total generated data set. The following report provides an introduction to the top level-objectives of the program related to passively safe decay heat removal, a detailed description of the engineering specifications, design features, and dimensions of the test facility at Argonne. Specifications of the sensors and their placement on the test facility will be provided, along with a complete channel listing of the data acquisition system.« less

Research of heat transfer of staggered horizontal bundles of finned tubes at free air convection

NASA Astrophysics Data System (ADS)

Novozhilova, A. V.; Maryna, Z. G.; Samorodov, A. V.; Lvov, E. A.

2017-11-01

The study of free-convective processes is important because of the cooling problem in many machines and systems, where other ways of cooling are impossible or impractical. Natural convective processes are common in the steam turbine air condensers of electric power plants located within the city limits, in dry cooling towers of circulating water systems, in condensers cooled by air and water, in radiators cooling oil of power electric transformers, in emergency cooling systems of nuclear reactors, in solar power, as well as in air-cooling of power semiconductor energy converters. All this makes actual the synthesis of the results of theoretical and experimental research of free convection for heat exchangers with finned tube bundles. The results of the study of free-convection heat transfer for two-, three- and four-row staggered horizontal bundles of industrial bimetallic finned tubes with finning factor of 16.8 and equilateral tubes arrangement are presented. Cross and diagonal steps in the bundles are the same: 58; 61; 64; 70; 76; 86; 100 mm, which corresponds to the relative steps: 1.042; 1.096; 1.152; 1.258; 1.366; 1.545; 1.797. These steps are standardized for air coolers. An equation for calculating the free-convection heat transfer, taking into account the influence of geometrical parameters in the range of Rayleigh number from 30,000 to 350,000 with an average deviation of ± 4.8%, has been obtained. The relationship presented in the article allows designing a wide range of air coolers for various applications, working in the free convection modes.

Core assembly storage structure

DOEpatents

Jones, Jr., Charles E.; Brunings, Jay E.

1988-01-01

A structure for the storage of core assemblies from a liquid metal-cooled nuclear reactor. The structure comprises an enclosed housing having a substantially flat horizontal top plate, a bottom plate and substantially vertical wall members extending therebetween. A plurality of thimble members extend downwardly through the top plate. Each thimble member is closed at its bottom end and has an open end adjacent said top plate. Each thimble member has a length and diameter greater than that of the core assembly to be stored therein. The housing is provided with an inlet duct for the admission of cooling air and an exhaust duct for the discharge of air therefrom, such that when hot core assemblies are placed in the thimbles, the heat generated will by convection cause air to flow from the inlet duct around the thimbles and out the exhaust duct maintaining the core assemblies at a safe temperature without the necessity of auxiliary powered cooling equipment.

Numerical study of air ingress transition to natural circulation in a high temperature helium loop

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franken, Daniel; Gould, Daniel; Jain, Prashant K.

Here, the generation-IV high temperature gas cooled reactors (HTGRs) are designed with many passive safety features, one of which is the ability to passively remove heat under a loss of coolant accident (LOCA). However, several common reactor designs do not prevent against a large break in the coolant system and may therefore experience a depressurized LOCA. This would lead to air entering into the reactor system via several potential modes of ingress: diffusion, gravity currents, and natural circulation. At the onset of a LOCA, the initial rate of air ingress is expected to be very slow because it is governedmore » by molecular diffusion. However, after several hours, natural circulation would commence, thus, bringing the air into the reactor system at a much higher rate. As a consequence, air ingress would cause the high temperature graphite matrix to oxidize, leading to its thermal degradation and decreased passive heat (decay) removal capability. Therefore, it is essential to understand the transition of air ingress from molecular diffusion to natural circulation in an HTGR system. This paper presents results from a computational fluid dynamics (CFD) model to study the air ingress transition behavior. These results are validated against an h-shaped high temperature helium loop experiment. Details are provided to quantitatively predict the transition time from molecular diffusion to natural circulation.« less

Numerical study of air ingress transition to natural circulation in a high temperature helium loop

DOE PAGES

Franken, Daniel; Gould, Daniel; Jain, Prashant K.; ...

2017-09-21

Here, the generation-IV high temperature gas cooled reactors (HTGRs) are designed with many passive safety features, one of which is the ability to passively remove heat under a loss of coolant accident (LOCA). However, several common reactor designs do not prevent against a large break in the coolant system and may therefore experience a depressurized LOCA. This would lead to air entering into the reactor system via several potential modes of ingress: diffusion, gravity currents, and natural circulation. At the onset of a LOCA, the initial rate of air ingress is expected to be very slow because it is governedmore » by molecular diffusion. However, after several hours, natural circulation would commence, thus, bringing the air into the reactor system at a much higher rate. As a consequence, air ingress would cause the high temperature graphite matrix to oxidize, leading to its thermal degradation and decreased passive heat (decay) removal capability. Therefore, it is essential to understand the transition of air ingress from molecular diffusion to natural circulation in an HTGR system. This paper presents results from a computational fluid dynamics (CFD) model to study the air ingress transition behavior. These results are validated against an h-shaped high temperature helium loop experiment. Details are provided to quantitatively predict the transition time from molecular diffusion to natural circulation.« less

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

DOEpatents

Hunsbedt, Anstein; Busboom, Herbert J.

1991-01-01

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

Passive cooling safety system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

1991-01-01

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

Indirect passive cooling system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1990-01-01

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

Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arcilesi, David J.; Ham, Tae Kyu; Kim, In Hun

2015-07-01

A critical event in the safety analysis of the very high-temperature gas-cooled reactor (VHTR) is an air-ingress accident. This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the coaxial cross vessel, which leads to a rapid reactor vessel depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. Following the vessel depressurization, the dominant mode of ingress of an air–helium mixture into the reactor vessel will either be molecular diffusion or density-driven stratified flow. The mode of ingress is hypothesized to dependmore » largely on the break conditions of the cross vessel. Since the time scales of these two ingress phenomena differ by orders of magnitude, it is imperative to understand under which conditions each of these mechanisms will dominate in the air ingress process. Computer models have been developed to analyze this type of accident scenario. There are, however, limited experimental data available to understand the phenomenology of the air-ingress accident and to validate these models. Therefore, there is a need to design and construct a scaled-down experimental test facility to simulate the air-ingress accident scenarios and to collect experimental data. The current paper focuses on the analyses performed for the design and operation of a 1/8th geometric scale (by height and diameter), high-temperature test facility. A geometric scaling analysis for the VHTR, a time scale analysis of the air-ingress phenomenon, a transient depressurization analysis of the reactor vessel, a hydraulic similarity analysis of the test facility, a heat transfer characterization of the hot plenum, a power scaling analysis for the reactor system, and a design analysis of the containment vessel are discussed.« less

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

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curry, J J; Gallagher, D W; Modarres, M

Appendices are presented concerning isolation condenser makeup; vapor suppression system; station air system; reactor building closed cooling water system; turbine building secondary closed water system; service water system; emergency service water system; fire protection system; emergency ac power; dc power system; event probability estimation; methodology of accident sequence quantification; and assignment of dominant sequences to release categories.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoder Jr, Graydon L; Aaron, Adam M; Cunningham, Richard Burns

2014-01-01

The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during themore » 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.« less

Liquid metal cooled nuclear reactors with passive cooling system

DOEpatents

Hunsbedt, Anstein; Fanning, Alan W.

1991-01-01

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

Advanced Low-Emissions Catalytic-Combustor Program, phase 1. [aircraft gas turbine engines

NASA Technical Reports Server (NTRS)

Sturgess, G. J.

1981-01-01

Six catalytic combustor concepts were defined, analyzed, and evaluated. Major design considerations included low emissions, performance, safety, durability, installations, operations and development. On the basis of these considerations the two most promising concepts were selected. Refined analysis and preliminary design work was conducted on these two concepts. The selected concepts were required to fit within the combustor chamber dimensions of the reference engine. This is achieved by using a dump diffuser discharging into a plenum chamber between the compressor discharge and the turbine inlet, with the combustors overlaying the prediffuser and the rear of the compressor. To enhance maintainability, the outer combustor case for each concept is designed to translate forward for accessibility to the catalytic reactor, liners and high pressure turbine area. The catalytic reactor is self-contained with air-cooled canning on a resilient mounting. Both selected concepts employed integrated engine-starting approaches to raise the catalytic reactor up to operating conditions. Advanced liner schemes are used to minimize required cooling air. The two selected concepts respectively employ fuel-rich initial thermal reaction followed by rapid quench and subsequent fuel-lean catalytic reaction of carbon monoxide, and, fuel-lean thermal reaction of some fuel in a continuously operating pilot combustor with fuel-lean catalytic reaction of remaining fuel in a radially-staged main combustor.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, C.W.

1985-02-19

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

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, Charles W.

1987-01-01

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

Design Construction and Operation of a Supercritical Carbon Dioxide (sCO 2) Loop for Investigation of Dry Cooling and Natural Circulation Potential for Use in Advanced Small Modular Reactors Utilizing sCO 2 Power Conversion Cycles.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Middleton, Bobby D.; Rodriguez, Salvador B.; Carlson, Matthew David

This report outlines the work completed for a Laboratory Directed Research and Development project at Sandia National Laboratories from October 2012 through September 2015. An experimental supercritical carbon dioxide (sCO 2 ) loop was designed, built, and o perated. The experimental work demonstrated that sCO 2 can be uti lized as the working fluid in an air - cooled, natural circulation configuration to transfer heat from a source to the ultimate heat sink, which is the surrounding ambient environment in most ca ses. The loop was also operated in an induction - heated, water - cooled configuration that allows formore » measurements of physical parameters that are difficult to isolate in the air - cooled configuration. Analysis included the development of two computational flu id dynamics models. Future work is anticipated to answer questions that were not covered in this project.« less

Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

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

Solvent refined coal reactor quench system

DOEpatents

Thorogood, Robert M.

1983-01-01

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

Solvent refined coal reactor quench system

DOEpatents

Thorogood, R.M.

1983-11-08

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

Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

NASA Technical Reports Server (NTRS)

Larson, V. R.; Gunn, S. V.; Lee, J. C.

1975-01-01

The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

Comparisons of RELAP5-3D Analyses to Experimental Data from the Natural Convection Shutdown Heat Removal Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew; Hu, Rui; Lisowski, Darius

2016-04-17

The Reactor Cavity Cooling System (RCCS) is an important passive safety system being incorporated into the overall safety strategy for high temperature advanced reactor concepts such as the High Temperature Gas- Cooled Reactors (HTGR). The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory (Argonne) reflects a 1/2-scale model of the primary features of one conceptual air-cooled RCCS design. The project conducts ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program, while also generating data for code validation purposes. While experiments are being conducted at themore » NSTF to evaluate the feasibility of the passive RCCS, parallel modeling and simulation efforts are ongoing to support the design, fabrication, and operation of these natural convection systems. Both system-level and high fidelity computational fluid dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. This paper provides a summary of the RELAP5-3D NSTF model development efforts and provides comparisons between simulation results and experimental data from the NSTF. Overall, the simulation results compared favorably to the experimental data, however, further analyses need to be conducted to investigate any identified differences.« less

Cooling system for a nuclear reactor

DOEpatents

Amtmann, Hans H.

1982-01-01

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

Neutron fluxes in test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 1: Objectives, summary results and introduction

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

The objective was to determine which reactor, conversion, and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. Specifically, the requirement was 10 megawatts for 5 years of full power operation and 10 years systems life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study. The concepts are: a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heat pipe and pumped tube-fin heat rejection; a lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator; a lithium cooled reactor with potassium Rankine turbine-alternator and heat pipe radiator; and a lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the lithium cooled incore thermionic reactor with heat pipe radiator.

THE COOLING REQUIREMENTS AND PROCESS SYSTEMS OF THE SOUTH AFRICAN RESEARCH REACTOR, SAFARI 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Colley, J.R.

1962-12-01

The SAFARI 1 research reactor is cooled and moderated by light water. There are three process systems, a primary water system which cools the reactor core and surroundings, a pool water system, and a secondary water system which removes the heat from the primary and pool systems. The cooling requirements for the reactor core and experimental facilities are outlined, and the cooling and purification functions of the three process systems are described. (auth)

Reactor water cleanup system

DOEpatents

Gluntz, Douglas M.; Taft, William E.

1994-01-01

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

Thermal storage/discharge performances of Cu-Si alloy for solar thermochemical process

NASA Astrophysics Data System (ADS)

Gokon, Nobuyuki; Yamaguchi, Tomoya; Cho, Hyun-seok; Bellan, Selvan; Hatamachi, Tsuyoshi; Kodama, Tatsuya

2017-06-01

The present authors (Niigata University, Japan) have developed a tubular reactor system using novel "double-walled" reactor/receiver tubes with carbonate molten-salt thermal storage as a phase change material (PCM) for solar reforming of natural gas and with Al-Si alloy thermal storage as a PCM for solar air receiver to produce high-temperature air. For both of the cases, the high heat capacity and large latent heat (heat of solidification) of the PCM phase circumvents the rapid temperature change of the reactor/receiver tubes at high temperatures under variable and uncontinuous characteristics of solar radiation. In this study, we examined cyclic properties of thermal storage/discharge for Cu-Si alloy in air stream in order to evaluate a potentiality of Cu-Si alloy as a PCM thermal storage material. Temperature-increasing performances of Cu-Si alloy are measured during thermal storage (or heat-charge) mode and during cooling (or heat-discharge) mode. A oxidation state of the Cu-Si alloy after the cyclic reaction was evaluated by using electron probe micro analyzer (EPMA).

The Concentration Of Tritium In Urine And Internal Radiation Dose Estimation Of PTNBR Radiation Workers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tjahaja, Poppy Intan; Sukmabuana, Putu; Aisyah, Neneng Nur

2010-12-23

The operation of Triga 2000 reactor in Nuclear Technology Center for Materials and Radiometry (PTNBR BATAN) normally produce tritium radionuclide which is the activation product of deuterium atom in reactor primary cooling water. According to previous monitoring, tritium was detected with the concentration of 8.236{+-}0.677 kBq/L and 1.704{+-}0.046 Bq/L in the primary cooling water and in reactor hall air, respectively. The tritium in reactor hall air chronically can be inhaled by the workers. In this research, tritium content in radiation workers' urine was determined to estimate the internal radiation doses received by the workers. About 50-100 mL of urine samplesmore » were collected from 48 PTNBR workers that is classified as 24 radiation workers and 24 administration staffs as a control. Urine samples of 25 mL were then prepared by active charcoal and KMnO{sub 4} addition and followed with complete distillation. The 2 mL of distillate was added with 13 mL scintillator, shaked vigorously and remained in cool and dark condition for about 24 hours. The tritium in the samples was then measured using liquid scintillation counter (LSC) for 1 hour. From the measurement results it was obtained that the tritium concentration in the urine of radiation workers were in the range of not detected and 5.191 Bq/mL, whereas in the administration staffs the concentration were between not detected and 4.607 Bq/mL. Internally radiation doses were calculated using the tritium concentration data, and it was found the averages about 0.602 {mu}Sv/year and 0.532 {mu}Sv/year for radiation workers and administration staffs, respectively. The doses received by the workers were lower than that of the permissible doses from tritium, i.e. 40 {mu}Sv/year.« less

Passive cooling system for top entry liquid metal cooled nuclear reactors

DOEpatents

Boardman, Charles E.; Hunsbedt, Anstein; Hui, Marvin M.

1992-01-01

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

Thermal device and method for production of carbon monoxide and hydrogen by thermal dissociation of hydrocarbon gases

DOEpatents

Detering, Brent A.; Kong, Peter C.

2001-01-01

Carbon monoxide is produced in a fast quench reactor. The production of carbon monoxide includes injecting carbon dioxide and some air into a reactor chamber having a high temperature at its inlet and a rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Carbon dioxide and other reactants such as methane and other low molecular weight hydrocarbons are injected into the reactor chamber. Other gas may be added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

Fluidized bed coal combustion reactor

NASA Technical Reports Server (NTRS)

Moynihan, P. I.; Young, D. L. (Inventor)

1981-01-01

A fluidized bed coal reactor includes a combination nozzle-injector ash-removal unit formed by a grid of closely spaced open channels, each containing a worm screw conveyor, which function as continuous ash removal troughs. A pressurized air-coal mixture is introduced below the unit and is injected through the elongated nozzles formed by the spaces between the channels. The ash build-up in the troughs protects the worm screw conveyors as does the cooling action of the injected mixture. The ash layer and the pressure from the injectors support a fluidized flame combustion zone above the grid which heats water in boiler tubes disposed within and/or above the combustion zone and/or within the walls of the reactor.

THE SM-1 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM, NOVEMBER 1954- DECEMBER 1960

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pressman, M; Pruett, P B

1961-08-31

BS>An environmental radiological monitoring program was conducted. All data obtained during a period extending from l 1/2 years prior to SM-1 reactor start-up through more than 3 years of reactor operation are summarized. The period extended from November 1954 through December 1960. Samples assayed for radioactivity include river water and bottom silt, SM-1 condenser cooling water, subsurface ground water, rain and snow, atmospheric particles obtained by air filtration and fallout, and biota. The report concludes that after more than 3 years of SM-1 reactor operation, no significant increase has been noted in the radiological background level in the Fort Belvoirmore » area.« less

Evaluation of some candidate materials for automobile thermal reactors in engine-dynamometer screening tests

NASA Technical Reports Server (NTRS)

Oldrieve, R. E.

1971-01-01

Fourteen materials were evaluated in engine screening tests on full-size thermal reactors for automobile engine pollution control systems. Cyclic test-stand engine operation provided 2 hours at 1040 C and a 20-minute air-cool to 70 C each test cycle. Each reactor material was exposed to 83 cycles in 200 hours of engine testing. On the basis of resistance to oxidation and distortion, the best materials included two ferritic iron alloys (Ge 1541 and Armco 18S/R), several commercial oxidation-resistant coatings on AlSl 651 (19-9 DL), and possibly uncoated AISI 310. The best commercial coatings were Cr-Al, Ni-Cr, and a glass ceramic.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 4: Concepts selection, conceptual designs, recommendations

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

A study was conducted by NASA Lewis Research Center for the Triagency SP-100 program office. The objective was to determine which reactor, conversion and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. The requirement was 10 megawatts for 5 years of full power operation and 10 years system life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study: (1) a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heatpipe and pumped tube fin rejection, (2) a Lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator,(3) a Lithium cooled reactor with a Potassium Rankine turbine-alternator and heat pipe radiator, and (4) a Lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the Lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the Lithium cooled incore thermionic reactor with heat pipe radiator.

Reactor water cleanup system

DOEpatents

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

1994-12-20

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

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

NASA Astrophysics Data System (ADS)

Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

I-NERI Quarterly Technical Report (April 1 to June 30, 2005)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang Oh; Prof. Hee Cheon NO; Prof. John Lee

2005-06-01

The objective of this Korean/United States/laboratory/university collaboration is to develop new advanced computational methods for safety analysis codes for very-high-temperature gas-cooled reactors (VHTGRs) and numerical and experimental validation of these computer codes. This study consists of five tasks for FY-03: (1) development of computational methods for the VHTGR, (2) theoretical modification of aforementioned computer codes for molecular diffusion (RELAP5/ATHENA) and modeling CO and CO2 equilibrium (MELCOR), (3) development of a state-of-the-art methodology for VHTGR neutronic analysis and calculation of accurate power distributions and decay heat deposition rates, (4) reactor cavity cooling system experiment, and (5) graphite oxidation experiment. Second quartermore » of Year 3: (A) Prof. NO and Kim continued Task 1. As a further plant application of GAMMA code, we conducted two analyses: IAEA GT-MHR benchmark calculation for LPCC and air ingress analysis for PMR 600MWt. The GAMMA code shows comparable peak fuel temperature trend to those of other country codes. The analysis results for air ingress show much different trend from that of previous PBR analysis: later onset of natural circulation and less significant rise in graphite temperature. (B) Prof. Park continued Task 2. We have designed new separate effect test device having same heat transfer area and different diameter and total number of U-bands of air cooling pipe. New design has smaller pressure drop in the air cooling pipe than the previous one as designed with larger diameter and less number of U-bands. With the device, additional experiments have been performed to obtain temperature distributions of the water tank, the surface and the center of cooling pipe on axis. The results will be used to optimize the design of SNU-RCCS. (C) Prof. NO continued Task 3. The experimental work of air ingress is going on without any concern: With nuclear graphite IG-110, various kinetic parameters and reaction rates for the C/CO2 reaction were measured. Then, the rates of C/CO2 reaction were compared to the ones of C/O2 reaction. The rate equation for C/CO2 has been developed. (D) INL added models to RELAP5/ATHENA to cacilate the chemical reactions in a VHTR during an air ingress accident. Limited testing of the models indicate that they are calculating a correct special distribution in gas compositions. (E) INL benchmarked NACOK natural circulation data. (F) Professor Lee et al at the University of Michigan (UM) Task 5. The funding was received from the DOE Richland Office at the end of May and the subcontract paperwork was delivered to the UM on the sixth of June. The objective of this task is to develop a state of the art neutronics model for determining power distributions and decay heat deposition rates in a VHTGR core. Our effort during the reporting period covered reactor physics analysis of coated particles and coupled nuclear-thermal-hydraulic (TH) calculations, together with initial calculations for decay heat deposition rates in the core.« less

High efficiency Brayton cycles using LNG

DOEpatents

Morrow, Charles W [Albuquerque, NM

2006-04-18

A modified, closed-loop Brayton cycle power conversion system that uses liquefied natural gas as the cold heat sink media. When combined with a helium gas cooled nuclear reactor, achievable efficiency can approach 68 76% (as compared to 35% for conventional steam cycle power cooled by air or water). A superheater heat exchanger can be used to exchange heat from a side-stream of hot helium gas split-off from the primary helium coolant loop to post-heat vaporized natural gas exiting from low and high-pressure coolers. The superheater raises the exit temperature of the natural gas to close to room temperature, which makes the gas more attractive to sell on the open market. An additional benefit is significantly reduced costs of a LNG revaporization plant, since the nuclear reactor provides the heat for vaporization instead of burning a portion of the LNG to provide the heat.

HVAC [Heating, Ventilation and Air Conditioning] subsystem design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1986-06-01

The HVAC system is a subsystem within the Mechanical Services Group (MSG). The HVAC system for the 4 x 350 MW(t) Modular HTGR Plant presently consists of ten, nonsafety-related subsystems located in the Nuclear Island (NI) and Energy Conversion Area (ECA) of the plant.

Process for remediation of plastic waste

DOEpatents

Pol, Vilas G [Westmont, IL; Thiyagarajan, Pappannan [Germantown, MD

2012-04-10

A single step process for degrading plastic waste by converting the plastic waste into carbonaceous products via thermal decomposition of the plastic waste by placing the plastic waste into a reactor, heating the plastic waste under an inert or air atmosphere until the temperature of 700.degree. C. is achieved, allowing the reactor to cool down, and recovering the resulting decomposition products therefrom. The decomposition products that this process yields are carbonaceous materials, and more specifically egg-shaped and spherical-shaped solid carbons. Additionally, in the presence of a transition metal compound, this thermal decomposition process produces multi-walled carbon nanotubes.

An intrinsically safe facility for forefront research and training on nuclear technologies — General description of the system

NASA Astrophysics Data System (ADS)

Mansani, L.; Bruzzone, M.; Frambati, S.; Reale, M.

2014-04-01

In the framework of research on generation-IV reactors, it is very important to have infrastructures specifically dedicated to the study of fundamental parameters in dynamics and kinetics of future fast-neutron reactors. Among various options pursued by international groups, Italy focused on lead-cooled reactors, which guarantee minimal neutron slowdown and capture and efficient cooling. In this paper it is described the design of a the low-power prototype generator, LEADS, that could be used within research facilities such as the National Laboratory of Legnaro of the INFN. The LEADS has a high safety standard in order to be used as a training facility, but it has also a good flexibility so as to allow a wide range of measurements and experiments. A high safety standard is achieved by limiting the reactor power to less than few hundred kW and the neutron multiplication factor k eff to less than 0.95 (a limiting value for spent fuel pool), by using a pure-uranium fuel (no plutonium) and by using solid lead as a diffuser. The proposed core is therefore intrinsically subcritical and has to be driven by an external neutron source generated by a proton beam impinging in a target. Preliminary simulations, performed with the MCNPX code indicated, for a 0.75mA continuous proton beam current at 70MeV proton energy, a reactor power of about 190kW when using a beryllium converter. The enriched-uranium fuel elements are immersed in a solid-lead matrix and contained within a steel vessel. The system is cooled by helium gas, which is transparent to neutrons and does not undergo activation. The gas is pumped by a compressor through specific holes at the entrance of the active volume with a temperature which varies according to the operating conditions and a pressure of about 1.1MPa. The hot gas coming out of the vessel is cooled by an external helium-water heat exchanger. The beryllium converter is cooled by its dedicated helium gas cooling system. After shutdown, the decay is completely dissipated by conduction through the lead reflector and steel vessel, and then evacuated by irradiation from the vessel surface to the external ambient air.

Evaluation of earthquake and tsunami on JSFR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chikazawa, Y.; Enuma, Y.; Kisohara, N.

2012-07-01

Evaluation of earthquake and tsunami on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. As for Tsunami, some parts of reactor building might be submerged including component cooling water system whose final heat sink is sea water. However, in the JSFR design, safety grade components are independent from component cooling water system (CCWS). The JSFR emergency power supply adopts a gas turbine system with air cooling, since JSFR does not basically require quick start-up of the emergency power supply thanks to the natural convection DHRS. Even in casemore » of long station blackout, the DHRS could be activated by emergency batteries or manually and be operated continuously by natural convection. (authors)« less

Investigation of a para-ortho hydrogen reactor for application to spacecraft sensor cooling

NASA Technical Reports Server (NTRS)

Nast, T. C.

1983-01-01

The utilization of solid hydrogen in space for sensor and instrument cooling is a very efficient technique for long term cooling or for cooling at high heat rates. The solid hydrogen can provide temperatures as low as 7 to 8 K to instruments. Vapor cooling is utilized to reduce parasitic heat inputs to the 7 to 8 K stage and is effective in providing intermediate cooling for instrument components operating at higher temperatures. The use of solid hydrogen in place of helium may lead to weight reductions as large as a factor of ten and an attendent reduction in system volume. The results of an investigation of a catalytic reactor for use with a solid hydrogen cooling system is presented. Trade studies were performed on several configurations of reactor to meet the requirements of high reactor efficiency with low pressure drop. Results for the selected reactor design are presented for both liquid hydrogen systems operating at near atmospheric pressure and the solid hydrogen cooler operating as low as 1 torr.

Portable thermo-photovoltaic power source

DOEpatents

Zuppero, Anthony C.; Krawetz, Barton; Barklund, C. Rodger; Seifert, Gary D.

1997-01-14

A miniature thermo-photovoltaic (TPV) device for generation of electrical power for use in portable electronic devices. A TPV power source is constructed to provide a heat source chemical reactor capable of using various fuels, such as liquid hydrocarbons, including but not limited to propane, LPG, butane, alcohols, oils and diesel fuels to generate a source of photons. A reflector dish guides misdirected photon energy from the photon source toward a photovoltaic array. A thin transparent protector sheet is disposed between the photon source and the array to reflect back thermal energy that cannot be converted to electricity, and protect the array from thermal damage. A microlens disposed between the protector sheet and the array further focuses the tailored band of photon energy from the photon source onto an array of photovoltaic cells, whereby the photon energy is converted to electrical power. A heat recuperator removes thermal energy from reactor chamber exhaust gases, preferably using mini- or micro-bellows to force air and fuel past the exhaust gases, and uses the energy to preheat the fuel and oxidant before it reaches the reactor, increasing system efficiency. Mini- or micro-bellows force ambient air through the system both to supply oxidant and to provide cooling. Finally, an insulator, which is preferably a super insulator, is disposed around the TPV power source to reduce fuel consumption, and to keep the TPV power source cool to the touch so it can be used in hand-held devices.

Fluidized bed combustor and tube construction therefor

DOEpatents

De Feo, Angelo; Hosek, William

1981-01-01

A fluidized bed combustor comprises a reactor or a housing which has a windbox distributor plate adjacent the lower end thereof which contains a multiplicity of hole and air discharge nozzles for discharging air and coal into a fluidized bed which is maintained above the distributor plate and below a take-off connection or flue to a cyclone separator in which some of the products of combustion are treated to remove the dust which is returned into the fluidized bed. A windbox is spaced below the fluidized bed and it has a plurality of tubes passing therethrough with the passage of combustion air and fluidizing air which passes through an air space so that fluidizing air is discharged into the reaction chamber fluidized bed at the bottom thereof to maintain the bed in a fluidized condition. A fluid, such as air, is passed through the tubes which extend through the windbox and provide a preheating of the combustion air and into an annular space between telescoped inner and outer tubes which comprise heat exchanger tubes or cooling tubes which extend upwardly through the distributor plate into the fluidized bed. The heat exchanger tubes are advantageously arranged so that they may be exposed in groups within the reactor in a cluster which is arranged within holding rings.

Tube construction for fluidized bed combustor

DOEpatents

De Feo, Angelo; Hosek, William

1984-01-01

A fluidized bed combustor comprises a reactor or a housing which has a windbox distributor plate adjacent the lower end thereof which contains a multiplicity of hole and air discharge nozzles for discharging air and coal into a fluidized bed which is maintained above the distributor plate and below a take-off connection or flue to a cyclone separator in which some of the products of combustion are treated to remove the dust which is returned into the fluidized bed. A windbox is spaced below the fluidized bed and it has a plurality of tubes passing therethrough with the passage of combustion air and fluidizing air which passes through an air space so that fluidizing air is discharged into the reaction chamber fluidized bed at the bottom thereof to maintain the bed in a fluidized condition. A fluid, such as air, is passed through the tubes which extend through the windbox and provide a preheating of the combustion air and into an annular space between telescoped inner and outer tubes which comprise heat exchanger tubes or cooling tubes which extend upwardly through the distributor plate into the fluidized bed. The heat exchanger tubes are advantageously arranged so that they may be exposed in groups within the reactor in a cluster which is arranged within holding rings.

Design of conduction cooling system for a high current HTS DC reactor

NASA Astrophysics Data System (ADS)

Dao, Van Quan; Kim, Taekue; Le Tat, Thang; Sung, Haejin; Choi, Jongho; Kim, Kwangmin; Hwang, Chul-Sang; Park, Minwon; Yu, In-Keun

2017-07-01

A DC reactor using a high temperature superconducting (HTS) magnet reduces the reactor’s size, weight, flux leakage, and electrical losses. An HTS magnet needs cryogenic cooling to achieve and maintain its superconducting state. There are two methods for doing this: one is pool boiling and the other is conduction cooling. The conduction cooling method is more effective than the pool boiling method in terms of smaller size and lighter weight. This paper discusses a design of conduction cooling system for a high current, high temperature superconducting DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

Testimony of Fred R. Mynatt before the Energy Research and Development Subcommittee of the Committee on Science, Space, and Technology, US House of Representatives. [Advanced fuel technology, gas-cooled reactor technology, and liquid metal-cooled reactor technology programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mynatt, F.R.

1987-03-18

This report provides a description of the statements submitted for the record to the committee on Science, Space, and Technology of the United States House of Representatives. These statements describe three principal areas of activity of the Advanced Reactor Technology Program of the Department of Energy (DOE). These areas are advanced fuel cycle technology, modular high-temperature gas-cooled reactor technology, and liquid metal-cooled reactor. The areas of automated reactor control systems, robotics, materials and structural design shielding and international cooperation were included in these statements describing the Oak Ridge National Laboratory's efforts in these areas. (FI)

The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

NASA Astrophysics Data System (ADS)

Dautray, Robert

2011-06-01

The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation, etc., radioprotection measures and treatment for the "transuranic" elements. For a long period of time, France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts and this could per se constitute one of the national priorities. Nous sommes naturellement bien conscients de l'énorme problème qui se pose au Japon actuellement comme suite au tremblement de terre et au tsunami de mars 2011 et leurs conséquences, notamment sur des installations électronucléaires. Le texte que nous présentons concerne des conditions totalement générales, indépendantes des problèmes spécifiques de sûreté qu'il faudra, de toute façon, traiter dans le cadre d'un développement éventuel de l'énergie nucléaire.We are aware, of course, of the huge problem that Japan has to deal with the aftermath of the quake and tsunami of March 2011 and their consequences on electronuclear power plants. The text that we present here concerns general physical topics independent of the specific safety problems, general physical topics which will have to be solved in the case of a contingent development of electronuclear power plants.

Computer model of catalytic combustion/Stirling engine heater head

NASA Technical Reports Server (NTRS)

Chu, E. K.; Chang, R. L.; Tong, H.

1981-01-01

The basic Acurex HET code was modified to analyze specific problems for Stirling engine heater head applications. Specifically, the code can model: an adiabatic catalytic monolith reactor, an externally cooled catalytic cylindrical reactor/flat plate reactor, a coannular tube radiatively cooled reactor, and a monolithic reactor radiating to upstream and downstream heat exchangers.

RADON LEVELS AND ЕQUIVALENT DOSE RATES AT THE IRT-SOFIA RESEARCH REACTOR SITE.

PubMed

Krezhov, Kiril; Mladenov, Aleksander; Dimitrov, Dobromir

2018-06-11

Results from radon measurements by active sampling of indoor air in the buildings within the Nuclear Scientific Experimental and Educational Centre (NSEEC) protected site at the Institute for Nuclear Research and Nuclear Energy (INRNE) are presented. The inspected buildings included in this report are the IRT research reactor structure and several auxiliary formations wherein the laundry facilities and the gamma irradiator GOU-1 (60Co source) are installed as well as the Central Alarm Station (CAS) premises. Besides the reactor hall and the primary cooling loop area, special attention was given to the premises of the First Class Radiochemical Laboratory in the IRT reactor basement. Determination of radon concentration distribution in the premises of the constructions within the site is an important part of radiation surveillance during the operation and maintenance of the NSEEC facilities as well as for their involvement in the educational activities at INRNE.

System Analysis for Decay Heat Removal in Lead-Bismuth Cooled Natural Circulated Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takaaki Sakai; Yasuhiro Enuma; Takashi Iwasaki

2002-07-01

Decay heat removal analyses for lead-bismuth cooled natural circulation reactors are described in this paper. A combined multi-dimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural circulation reactors. For the preliminary study, transient analysis has been performed for a 100 MWe lead-bismuth-cooled reactor designed by Argonne National Laboratory (ANL). In addition, decay heat removal characteristics of a 400 MWe lead-bismuth-cooled natural circulation reactor designed by Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. PRACS (Primary Reactor Auxiliary Cooling System) is prepared for the JNC's concept to get sufficient heatmore » removal capacity. During 2000 sec after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 Centigrade, because the buoyancy force in a primary circulation path is temporary reduced. However, the natural circulation is recovered by the PRACS system and the out let temperature decreases successfully. (authors)« less

System Analysis for Decay Heat Removal in Lead-Bismuth-Cooled Natural-Circulation Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sakai, Takaaki; Enuma, Yasuhiro; Iwasaki, Takashi

2004-03-15

Decay heat removal analyses for lead-bismuth-cooled natural-circulation reactors are described in this paper. A combined multidimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural-circulation reactors. For the preliminary study, transient analysis has been performed for a 300-MW(thermal) lead-bismuth-cooled reactor designed by Argonne National Laboratory. In addition, decay heat removal characteristics of a 400-MW(electric) lead-bismuth-cooled natural-circulation reactor designed by the Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. The primary reactor auxiliary cooling system (PRACS) is prepared for the JNC concept to get sufficient heat removal capacity. During 2000 smore » after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 deg. C because the buoyancy force in a primary circulation path is temporarily reduced. However, the natural circulation is recovered by the PRACS system, and the outlet temperature decreases successfully.« less

Process for remediation of plastic waste

DOEpatents

Pol, Vilas G; Thiyagarajan, Pappannan

2013-11-12

A single step process for degrading plastic waste by converting the plastic waste into carbonaceous products via thermal decomposition of the plastic waste by placing the plastic waste into a reactor, heating the plastic waste under an inert or air atmosphere until the temperature of about 700.degree. C. is achieved, allowing the reactor to cool down, and recovering the resulting decomposition products therefrom. The decomposition products that this process yields are carbonaceous materials, and more specifically carbon nanotubes having a partially filled core (encapsulated) adjacent to one end of the nanotube. Additionally, in the presence of a transition metal compound, this thermal decomposition process produces multi-walled carbon nanotubes.

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

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

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

AP1000{sup R} severe accident features and post-Fukushima considerations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scobel, J. H.; Schulz, T. L.; Williams, M. G.

2012-07-01

The AP1000{sup R} passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at Fukushima without compromising core and containment integrity. The AP1000 plant shuts down the reactor, cools the core, containment and spent fuel pool for more than 3 days using passive systems that do not require AC or DC power or operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. To provide defense-in-depth for design extension conditions, themore » AP1000 plant has engineered features that mitigate the effects of core damage. Engineered features retain damaged core debris within the reactor vessel as a key feature. Other aspects of the design protect containment integrity during severe accidents, including unique features of the AP1000 design relative to passive containment cooling with water and air, and hydrogen management. (authors)« less

Heat stress control in the TMI-2 (Three Mile Island Unit 2) defueling and decontamination activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schork, J.S.; Parfitt, B.A.

During the initial stages of the Three Mile Island Unit 2 (TMI-2) defueling and decontamination activities for the reactor building, it was realized that the high levels of loose radioactive contamination would require the use of extensive protective clothing by entry personnel. While there was no doubt that layered protective clothing protects workers from becoming contaminated, it was recognized that these same layers of clothing would impose a very significant heat stress burden. To prevent the potentially serious consequences of a severe reaction to heat stress by workers in the hostile environment of the TMI-2 reactor building and yet maintainmore » the reasonable work productivity necessary to perform the recovery adequately, an effective program of controlling worker exposure to heat stress had to be developed. Body-cooling devices produce a flow of cool air, which is introduced close to the skin to remove body heat through convection and increased sweat evaporation. The cooling effect produced by the Vortex tube successfully protected the workers from heat stress, however, there were several logistical and operational problems that hindered extensive use of these devices. The last type of cooling garment examined was the frozen water garment (FWG) developed by Elizier Kamon at the Pennsylvania State University as part of an Electric Power Research Institute research grant. Personal protection, i.e., body cooling, engineering controls, and administrative controls, have been implemented successfully.« less

Maximal design basis accident of fusion neutron source DEMO-TIN

NASA Astrophysics Data System (ADS)

Kolbasov, B. N.

2015-12-01

When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission-fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

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

DOEpatents

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

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

Supporting the Air Mobility Needs of Geographic Combatant Commanders: An Evaluation Using the Principal-Agent Construct

DTIC Science & Technology

2014-06-01

had reached over 500,000. Another important aspect of this disaster was the damage sustained by several Fukushima Daiichi Nuclear plant reactors.3...The damage, resulting from the constant battering of tsunami waves, affected the cooling systems of the nuclear plant and resulted in several ... Nuclear Regulatory Commission & DoE nuclear expertise to help with the emerging Fukushima crisis. All branches of the US armed forces actively

Loss of control air at Browns Ferry Unit One: accident sequence analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrington, R.M.; Hodge, S.A.

1986-04-01

This study describes the predicted response of the Browns Ferry Nuclear Plant to a postulated complete failure of plant control air. The failure of plant control air cascades to include the loss of drywell control air at Units 1 and 2. Nevertheless, this is a benign accident unless compounded by simultaneous failures in the turbine-driven high pressure injection systems. Accident sequence calculations are presented for Loss of Control Air sequences with assumed failure upon demand of the Reactor Core Isolation Cooling (RCIC) and the High Pressure Coolant Injection (HPCI) at Unit 1. Sequences with and without operator action are considered.more » Results show that the operators can prevent core uncovery if they take action to utilize the Control Rod Drive Hydraulic System as a backup high pressure injection system.« less

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE PAGES

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; ...

2016-12-21

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Nuclear reactor vessel fuel thermal insulating barrier

DOEpatents

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki; Miura, Ryosuke

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design.more » The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.« less

NEUTRONIC REACTORS

DOEpatents

Vernon, H.C.

1959-01-13

A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

Emergency Cooling of Nuclear Power Plant Reactors With Heat Removal By a Forced-Draft Cooling Tower

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murav’ev, V. P., E-mail: murval1@mail.ru

The feasibility of heat removal during emergency cooling of a reactor by a forced-draft cooling tower with accumulation of the peak heat release in a volume of precooled water is evaluated. The advantages of a cooling tower over a spray cooling pond are demonstrated: it requires less space, consumes less material, employs shorter lines in the heat removal system, and provides considerably better protection of the environment from wetting by entrained moisture.

158. ARAIII Reactor building (ARA608) Secondary cooling loop and piping ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

158. ARA-III Reactor building (ARA-608) Secondary cooling loop and piping plan. This drawing was selected as a typical example of piping arrangements within reactor building. Aerojet/general 880-area/GCRE-608-P-16. Date: February 1958. INeel index code no. 063-0608-50-013-102641. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Method and apparatus for a catalytic firebox reactor

DOEpatents

Smith, Lance L.; Etemad, Shahrokh; Ulkarim, Hasan; Castaldi, Marco J.; Pfefferle, William C.

2001-01-01

A catalytic firebox reactor employing an exothermic catalytic reaction channel and multiple cooling conduits for creating a partially reacted fuel/oxidant mixture. An oxidation catalyst is deposited on the walls forming the boundary between the multiple cooling conduits and the exothermic catalytic reaction channel, on the side of the walls facing the exothermic catalytic reaction channel. This configuration allows the oxidation catalyst to be backside cooled by any fluid passing through the cooling conduits. The heat of reaction is added to both the fluid in the exothermic catalytic reaction channel and the fluid passing through the cooling conduits. After discharge of the fluids from the exothermic catalytic reaction channel, the fluids mix to create a single combined flow. A further innovation in the reactor incorporates geometric changes in the exothermic catalytic reaction channel to provide streamwise variation of the velocity of the fluids in the reactor.

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

DOEpatents

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

1993-12-14

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

Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

NASA Astrophysics Data System (ADS)

Cisneros, Anselmo Tomas, Jr.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of economic assumptions about the electricity market to evaluate the economic implications of design decisions. The optimal PB-FHR design---Mark 1 PB-FHR---is described along with a detailed summary of its performance characteristics including: the burnup, the burnup evolution, temperature reactivity coefficients, the power distribution, radiation damage distributions, control element worths, decay heat curves and tritium production rates. The Mk1 PB-FHR satisfies the PB-FHR safety criteria. The fuel, moderator (pebble core, pebble shell, graphite matrix, TRISO layers) and coolant have global negative temperature reactivity coefficients and the fuel temperatures are well within their limits.

Benchmark Simulation of Natural Circulation Cooling System with Salt Working Fluid Using SAM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahmed, K. K.; Scarlat, R. O.; Hu, R.

Liquid salt-cooled reactors, such as the Fluoride Salt-Cooled High-Temperature Reactor (FHR), offer passive decay heat removal through natural circulation using Direct Reactor Auxiliary Cooling System (DRACS) loops. The behavior of such systems should be well-understood through performance analysis. The advanced system thermal-hydraulics tool System Analysis Module (SAM) from Argonne National Laboratory has been selected for this purpose. The work presented here is part of a larger study in which SAM modeling capabilities are being enhanced for the system analyses of FHR or Molten Salt Reactors (MSR). Liquid salt thermophysical properties have been implemented in SAM, as well as properties ofmore » Dowtherm A, which is used as a simulant fluid for scaled experiments, for future code validation studies. Additional physics modules to represent phenomena specific to salt-cooled reactors, such as freezing of coolant, are being implemented in SAM. This study presents a useful first benchmark for the applicability of SAM to liquid salt-cooled reactors: it provides steady-state and transient comparisons for a salt reactor system. A RELAP5-3D model of the Mark-1 Pebble-Bed FHR (Mk1 PB-FHR), and in particular its DRACS loop for emergency heat removal, provides steady state and transient results for flow rates and temperatures in the system that are used here for code-to-code comparison with SAM. The transient studied is a loss of forced circulation with SCRAM event. To the knowledge of the authors, this is the first application of SAM to FHR or any other molten salt reactors. While building these models in SAM, any gaps in the code’s capability to simulate such systems are identified and addressed immediately, or listed as future improvements to the code.« less

COOLING TOWER PUMP HOUSE, TRA606. THREE OF SIX SECTIONS OF ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

COOLING TOWER PUMP HOUSE, TRA-606. THREE OF SIX SECTIONS OF COOLING TOWER ARE VISIBLE ABOVE RAILING. PUMP HOUSE IN FOREGROUND IS ON SOUTH SIDE OF COOLING TOWER. NOTE THREE PIPES TAKING WATER FROM PUMP HOUSE TO HOT DECK OF COOLING TOWER. EMERGENCY WATER SUPPLY TOWER IS ALSO IN VIEW. INL NEGATIVE NO. 6197. Unknown Photographer, 6/27/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Maximal design basis accident of fusion neutron source DEMO-TIN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kolbasov, B. N., E-mail: Kolbasov-BN@nrcki.ru

2015-12-15

When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission–fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or themore » first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.« less

Modeling and Comparison of Options for the Disposal of Excess Weapons Plutonium in Russia

DTIC Science & Technology

2002-04-01

fuel LWR cooling time LWR Pu load rate LWR net destruction frac ~ LWR reactors op life mox core frac Excess Separated Pu HTGR Cycle Pu in Waste LWR MOX...reflecting the cycle used in this type of reactor. For the HTGR , the entire core consists of plutonium fuel , therefore a core fraction is not specified...cooling time Time spent fuel unloaded from HTGR reactor must cool before permanently stored 3 years Mox core fraction Fraction of

INERT GAS SHIELD FOR WELDING

DOEpatents

Jones, S.O.; Daly, F.V.

1958-10-14

S>An inert gas shield is presented for arc-welding materials such as zirconium that tend to oxidize rapidly in air. The device comprises a rectangular metal box into which the welding electrode is introduced through a rubber diaphragm to provide flexibility. The front of the box is provided with a wlndow having a small hole through which flller metal is introduced. The box is supplied with an inert gas to exclude the atmosphere, and with cooling water to promote the solidification of the weld while in tbe inert atmosphere. A separate water-cooled copper backing bar is provided underneath the joint to be welded to contain the melt-through at the root of the joint, shielding the root of the joint with its own supply of inert gas and cooling the deposited weld metal. This device facilitates the welding of large workpieces of zirconium frequently encountered in reactor construction.

Heat exchanger with auxiliary cooling system

DOEpatents

Coleman, John H.

1980-01-01

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.

Material Issues of Blanket Systems for Fusion Reactors - Compatibility with Cooling Water -

NASA Astrophysics Data System (ADS)

Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

Environmental assisted cracking (EAC) is one of the material issues for the reactor core components of light water power reactors(LWRs). Much experience and knowledge have been obtained about the EAC in the LWR field. They will be useful to prevent the EAC of water-cooled blanket systems of fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC in a water-cooled blanket does not seem to be acritical issue. However, some uncertainties about influences on water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations elucidating the uncertainties are discussed.

78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0237] Cost-Benefit Analysis for Radwaste Systems for Light... (RG) 1.110, ``Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors... components for light water nuclear power reactors. ADDRESSES: Please refer to Docket ID NRC-2013-0237 when...

Application of a Self-Actuating Shutdown System (SASS) to a Gas-Cooled Fast Reactor (GCFR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Germer, J.H.; Peterson, L.F.; Kluck, A.L.

1980-09-01

The application of a SASS (Self-Actuated Shutdown System) to a GCFR (Gas-Cooled Fast Reactor) is compared with similar systems designed for an LMFBR (Liquid Metal Fast Breeder Reactor). A comparison of three basic SASS concepts is given: hydrostatic holdup, fluidic control, and magnetic holdup.

PROCESS FOR COOLING A NUCLEAR REACTOR

DOEpatents

Borst, L.B.

1962-12-11

This patent relates to the operation of a reactor cooled by liquid sulfur dioxide. According to the invention the pressure on the sulfur dioxide in the reactor is maintained at least at the critical pressure of the sulfur dioxide. Heating the sulfur dioxide to its critical temperature results in vaporization of the sulfur dioxide without boiling. (AEC)

78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

...The U.S. Nuclear Regulatory Commission (NRC) is issuing a revision to regulatory guide (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance for preoperational testing of new pressurized water reactor (PWR) designs.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

LIQUID METAL REACTOR COOLING SYSTEMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aberdam, M.; Gros, G.

1965-02-01

This report is part of a series of bibliographies. The specific purpose of this report is to describe the various elements of the cooling systems in the principal liquid-metal-cooled reactors now operating, being contsructed, or in the design stage. The information given is drawn from reports or publicatios received during or before September 1964.

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2015-03-01

The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoidmore » overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety system components such as the safety relief valve (SRV), the RCIC system, the wet well, and the dry well. The results show reasonable system behaviors while exhibiting rich dynamics such as variable flow rates through RCIC turbine and pump during the SBO transient. The model has the potential to resolve the Fukushima RCIC mystery after adding the off-design two-phase turbine operation model and other additional improvements.« less

Overall Effectiveness Measurement at Engine Temperatures with Reactive Film Cooling and Surface Curvature

DTIC Science & Technology

2015-03-26

supplied with hot mainstream gas from a well-stirred reactor operating on a propane/air mixture capable of multiple equivalence ratios. The...consequently the Pr’s were around 2.3 which is more representative of liquids rather than 0.7 of most gases. Case 5 vastly overpredicted η showing the...performance engines, an afterburner may be added for additional power which exposes aft sections and the nozzle of the engine to a high temperature flame

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-01-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-03-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

Control rod system useable for fuel handling in a gas-cooled nuclear reactor

DOEpatents

Spurrier, Francis R.

1976-11-30

A control rod and its associated drive are used to elevate a complete stack of fuel blocks to a position above the core of a gas-cooled nuclear reactor. A fuel-handling machine grasps the control rod and the drive is unlatched from the rod. The stack and rod are transferred out of the reactor, or to a new location in the reactor, by the fuel-handling machine.

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

NASA Astrophysics Data System (ADS)

Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

2014-08-01

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

Transient Response to Rapid Cooling of a Stainless Steel Sodium Heat Pipe

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Houts, Michael G.

2011-01-01

Compact fission power systems are under consideration for use in long duration space exploration missions. Power demands on the order of 500 W, to 5 kW, will be required for up to 15 years of continuous service. One such small reactor design consists of a fast spectrum reactor cooled with an array of in-core alkali metal heat pipes coupled to thermoelectric or Stirling power conversion systems. Heat pipes advantageous attributes include a simplistic design, lack of moving parts, and well understood behavior. Concerns over reactor transients induced by heat pipe instability as a function of extreme thermal transients require experimental investigations. One particular concern is rapid cooling of the heat pipe condenser that would propagate to cool the evaporator. Rapid cooling of the reactor core beyond acceptable design limits could possibly induce unintended reactor control issues. This paper discusses a series of experimental demonstrations where a heat pipe operating at near prototypic conditions experienced rapid cooling of the condenser. The condenser section of a stainless steel sodium heat pipe was enclosed within a heat exchanger. The heat pipe - heat exchanger assembly was housed within a vacuum chamber held at a pressure of 50 Torr of helium. The heat pipe was brought to steady state operating conditions using graphite resistance heaters then cooled by a high flow of gaseous nitrogen through the heat exchanger. Subsequent thermal transient behavior was characterized by performing an energy balance using temperature, pressure and flow rate data obtained throughout the tests. Results indicate the degree of temperature change that results from a rapid cooling scenario will not significantly influence thermal stability of an operating heat pipe, even under extreme condenser cooling conditions.

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gougar, Hans David

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each ofmore » the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.« less

Passive cooling system for nuclear reactor containment structure

DOEpatents

Gou, Perng-Fei; Wade, Gentry E.

1989-01-01

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fathi, Nima; McDaniel, Patrick; Vorobieff, Peter

The aim of this paper is evaluating the efficiency of a novel combined solar-nuclear cycle. CFD-Thermal analysis is performed to apply the available surplus heat from the nuclear cycle and measure the available kinetic energy of air for the turbine of a solar chimney power plant system (SCPPS). The presented idea helps to decrease the thermal pollution and handle the water shortage supply for water plant by replacing the cooling tower by solar chimney power plant to get the surplus heat from the available warm air in the secondary loop of the reactor. By applying this idea to a typicalmore » 1000 MW nuclear power plant with a 0.33 thermal efficiency, we can increase it to 0.39.« less

Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wilson, Dane F.

2015-09-01

Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity],more » a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O 2 cannot be ignored, especially for the FHR, in which environment the product, SiO 2, can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.« less

ADVANCED COURSE ON FUEL ELEMENTS FOR WATER COOLED POWER REACTORS, ORGANIZED BY THE NETHERLANDS'-NORWEGIAN REACTOR SCHOOL AT INSTITUTT FOR ATOMENERGI, KJELLER, NORWAY, 22nd AUGUST-3rd SEPTEMBER,1960. VOLUME III

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aas, S.; Barendregt, T.J.; Chesne, A.

1960-07-01

A series of lectures on fuel elements for water-cooled power reactors are presented. Topics covered include fabrication, properties, cladding, radiation damage, design, cycling, storage and transpont, and reprocessing. Separate records have been prepared for each section.

Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

NASA Astrophysics Data System (ADS)

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y.

2012-05-01

There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

Natural circulating passive cooling system for nuclear reactor containment structure

DOEpatents

Gou, Perng-Fei; Wade, Gentry E.

1990-01-01

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

Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

NASA Technical Reports Server (NTRS)

Lantz, E.; Breitwieser, R.; Niederauer, G. F.

1974-01-01

There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1983-06-01

During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

77 FR 64148 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-18

... Regulatory Guides (RG) RG 1.79, ````Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors,'' Revision 2 and RG 1.79.1, ``Initial Test Program of Emergency Core Cooling Systems for...

DETECTION OF COATING FAILURES IN A NEUTRONIC REACTOR

DOEpatents

Snell, A.H.; Allison, S.K.

1958-02-11

This patent relates to water-cooled reactor systems and discloses a means to detect leaks in the jackets of jacketed fuel elements comprising a neutron detector located in the cooling water discharge pipe,the pipe being provided with an enlarged portion for housing the detector so that the latter is completely surrounded by the water in its passage through the pipe, said enlarged portion and detector being shielded from the reactor for the purpose of detecting only those delayed neutrons emitted in the cooling water and due to the latter picking up fission fragments from the defective fuel elements.

High-Temperature Gas-Cooled Test Reactor Point Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-04-01

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2013 CFR

2013-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2012 CFR

2012-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2014 CFR

2014-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

A nuclear powered air cushion freighter for the 1980's.

NASA Technical Reports Server (NTRS)

Anderson, J. L.

1971-01-01

A design for a transoceanic, dry cargo-carrying freighter is suggested; its use and operation in port are discussed. With a gross weight of 4500 metric tons (5000 tons), more than 50 percent of which is cargo, it will cruise at 50 meters per second (100 knots) in waves 2.4 meters (8 ft) high. Its peripheral jet-flexible skirt air cushion concept and air thrustors will let the freighter go over waves 8 meters high at reduced velocity. Power comes from a 1280 megawatt, helium-cooled thermal reactor. It could dock at any major port in the world, but because it needs no surface contact, it could also travel inland to land-locked ports. A modular terminal design and methods of cargo transfer are suggested. The concept of cargo containerization influences both the freighter and terminal design.

Preliminary design of high temperature ultrasonic transducers for liquid sodium environments

NASA Astrophysics Data System (ADS)

Prowant, M. S.; Dib, G.; Qiao, H.; Good, M. S.; Larche, M. R.; Sexton, S. S.; Ramuhalli, P.

2018-04-01

Advanced reactor concepts include fast reactors (including sodium-cooled fast reactors), gas-cooled reactors, and molten-salt reactors. Common to these concepts is a higher operating temperature (when compared to light-water-cooled reactors), and the proposed use of new alloys with which there is limited operational experience. Concerns about new degradation mechanisms, such as high-temperature creep and creep fatigue, that are not encountered in the light-water fleet and longer operating cycles between refueling intervals indicate the need for condition monitoring technology. Specific needs in this context include periodic in-service inspection technology for the detection and sizing of cracking, as well as technologies for continuous monitoring of components using in situ probes. This paper will discuss research on the development and evaluation of high temperature (>550°C; >1022°F) ultrasonic probes that can be used for continuous monitoring of components. The focus of this work is on probes that are compatible with a liquid sodium-cooled reactor environment, where the core outlet temperatures can reach 550°C (1022°F). Modeling to assess sensitivity of various sensor configurations and experimental evaluation have pointed to a preferred design and concept of operations for these probes. This paper will describe these studies and ongoing work to fabricate and fully evaluate survivability and sensor performance over extended periods at operational temperatures.

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

NASA Astrophysics Data System (ADS)

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

2005-02-01

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

Turbomachine rotor with improved cooling

DOEpatents

Hultgren, Kent Goran; McLaurin, Leroy Dixon; Bertsch, Oran Leroy; Lowe, Perry Eugene

1998-01-01

A gas turbine rotor has an essentially closed loop cooling air scheme in which cooling air drawn from the compressor discharge air that is supplied to the combustion chamber is further compressed, cooled, and then directed to the aft end of the turbine rotor. Downstream seal rings attached to the downstream face of each rotor disc direct the cooling air over the downstream disc face, thereby cooling it, and then to cooling air passages formed in the rotating blades. Upstream seal rings attached to the upstream face of each disc direct the heated cooling air away from the blade root while keeping the disc thermally isolated from the heated cooling air. From each upstream seal ring, the heated cooling air flows through passages in the upstream discs and is then combined and returned to the combustion chamber from which it was drawn.

Turbomachine rotor with improved cooling

DOEpatents

Hultgren, K.G.; McLaurin, L.D.; Bertsch, O.L.; Lowe, P.E.

1998-05-26

A gas turbine rotor has an essentially closed loop cooling air scheme in which cooling air drawn from the compressor discharge air that is supplied to the combustion chamber is further compressed, cooled, and then directed to the aft end of the turbine rotor. Downstream seal rings attached to the downstream face of each rotor disc direct the cooling air over the downstream disc face, thereby cooling it, and then to cooling air passages formed in the rotating blades. Upstream seal rings attached to the upstream face of each disc direct the heated cooling air away from the blade root while keeping the disc thermally isolated from the heated cooling air. From each upstream seal ring, the heated cooling air flows through passages in the upstream discs and is then combined and returned to the combustion chamber from which it was drawn. 5 figs.

Neutronics and Transient Calculations for the Conversion of the Transient Reactor Rest Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Papadias, Dionissios D.

2015-01-01

The Transient Reactor Test Facility (TREAT) is a graphite-reflected, graphitemoderated, and air-cooled reactor fueled with 93.1% enriched UO2 particles dispersed in graphite, with a carbon-to-235U ratio of ~10000:1. TREAT was used to simulate accident conditions by subjecting fuel test samples placed at the center of the core to high energy transient pulses. The transient pulse production is based on the core’s selflimiting nature due to the negative reactivity feedback provided by the fuel graphite as the core temperature rises. The analysis of the conversion of TREAT to low enriched uranium (LEU) is currently underway. This paper presents the analytical methodsmore » used to calculate the transient performance of TREAT in terms of power pulse production and resulting peak core temperatures. The validation of the HEU neutronics TREAT model, the calculation of the temperature distribution and the temperature reactivity feedback as well as the number of fissions generated inside fuel test samples are discussed.« less

A thermodynamic approach for advanced fuels of gas-cooled reactors

NASA Astrophysics Data System (ADS)

Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

2005-09-01

For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

Method for fabricating a seal between a ceramic and a metal alloy

DOEpatents

Kelsey, Jr., Paul V.; Siegel, William T.

1983-01-01

A method of fabricating a seal between a ceramic and an alloy comprising the steps of prefiring the alloy in an atmosphere with a very low partial pressure of oxygen, firing the assembled alloy and ceramic in air, and gradually cooling the fired assembly to avoid the formation of thermal stress in the ceramic. The method forms a bond between the alloy and the ceramic capable of withstanding the environment of a pressurized water reactor and suitable for use in an electrical conductivity sensitive liquid level transducer.

Low and medium heating value coal gas catalytic combustor characterization

NASA Technical Reports Server (NTRS)

Schwab, J. A.

1982-01-01

Catalytic combustion with both low and medium heating value coal gases obtained from an operating gasifier was demonstrated. A practical operating range for efficient operation was determined, and also to identify potential problem areas were identified for consideration during stationary gas turbine engine design. The test rig consists of fuel injectors, a fuel-air premixing section, a catalytic reactor with thermocouple instrumentation and a single point, water cooled sample probe. The test rig included inlet and outlet transition pieces and was designed for installation into an existing test loop.

Method for fabricating a seal between a ceramic and a metal alloy

DOEpatents

Kelsey, P.V. Jr.; Siegel, W.T.

1983-08-16

A method of fabricating a seal between a ceramic and an alloy comprising the steps of prefiring the alloy in an atmosphere with a very low partial pressure of oxygen, firing the assembled alloy and ceramic in air, and gradually cooling the fired assembly to avoid the formation of thermal stress in the ceramic. The method forms a bond between the alloy and the ceramic capable of withstanding the environment of a pressurized water reactor and suitable for use in an electrical conductivity sensitive liquid level transducer.

Power and Thermal Technologies for Air and Space-Scientific Research Program. Delivery Order 0012: High-Temperature Superconductor Performance Enhancement

DTIC Science & Technology

2010-06-01

house to grow CNTs. Initially the CNTs were grown at atmospheric pressure using C2H2/Ar mixtures. Prior to deposition, the quartz tube of the reactor...imaged clearly. It appears that there could be some amorphous carbon present on the surface of the tubes with the present set of conditions used and...chip cooling with CNT microfin architectures have been recently proposed by Kordas et al. [5]. CNT films as thermal interface materials were also

Clearing of ventilating emissions in low temperature environment of plasma

NASA Astrophysics Data System (ADS)

Mansurov, R. Sh; Rafalskaya, T. A.

2017-11-01

The method of high-temperature processing of streams of the ventilating air which is a subject clearing from organic pollutions is developed. Data about its efficiency, including on a number of economic parameters are obtained. Results of work are recommended for use, first of all, by development clearing plasma-thermal reactors (CPTR) for clearing air, especially from toxic substances, and also for large technological clearing installations, containing organic ventilating emissions (OVE). It is created experimental CPTR. Laws of the expiration of a plasma jet in stream of OVE limited by cylindrical walls, water-cooled channel are experimentally investigated. Dependences of a trajectory and long-range the plasma jet blown radially in stream of OVE are received. Heat exchange of stream of OVE with walls of CPTR after blowing a plasma jet is experimentally investigated; dependences of distribution of temperatures on length of a reactor and a thermal stream in a wall of channel of CPTR are received. Are investigated chemical compound of OVE after plasma-thermal clearing, some experimental data by formation of oxides of nitrogen and mono-oxide of carbon during clearing are received.

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

DOEpatents

DeVolpi, A.

1984-07-20

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

Turbine airfoil cooling system with cooling systems using high and low pressure cooling fluids

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marsh, Jan H.; Messmann, Stephen John; Scribner, Carmen Andrew

A turbine airfoil cooling system including a low pressure cooling system and a high pressure cooling system for a turbine airfoil of a gas turbine engine is disclosed. In at least one embodiment, the low pressure cooling system may be an ambient air cooling system, and the high pressure cooling system may be a compressor bleed air cooling system. In at least one embodiment, the compressor bleed air cooling system in communication with a high pressure subsystem that may be a snubber cooling system positioned within a snubber. A delivery system including a movable air supply tube may be usedmore » to separate the low and high pressure cooling subsystems. The delivery system may enable high pressure cooling air to be passed to the snubber cooling system separate from low pressure cooling fluid supplied by the low pressure cooling system to other portions of the turbine airfoil cooling system.« less

Development work for a borax internal core-catcher for a gas-cooled fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Donne, M.D.; Dorner, S.; Schumacher, G.

1978-07-01

Preliminary thermal calculations show that a corecatcher, which is able to cope with the complete meltdown of the core and blankets of a 1000-MW(electric) gas-cooled fast reactor, appears to be feasible. This core-catcher is based on borax (Na/sub 2/B/sub 4/O/sub 7/) dissolving the oxide fuel and the fission products occurring in oxide form. The borax is contained in steel boxes forming a 2.2-m-thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel (PCRV), just underneath the reactor core. After a complete meltdown accident, the fission products, in oxide form, are dispersed in the pool formedmore » by the liquid borax. The metallic fission products are contained in the steel lying below the borax pool and in contact with the water-cooled PCRV liner. The volumetric power density of the molten core is conveniently reduced as it is dissolved in the borax, and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system.« less

Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho; Lee, Sangjin; Jin, Yoon-Su; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2014-09-01

This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

A lightweight ambient air-cooling unit for use in hazardous environments.

PubMed

Chen, Y T; Constable, S H; Bomalaski, S H

1997-01-01

Recent research demonstrated (a) the effectiveness of intermittent conditioned air cooling during rest breaks to significantly reduce cumulative heat storage and (b) that longer work sessions were possible for individuals wearing chemical defense ensembles. To further advance this concept, a strategy for implementing continuous air cooling was conceived; ambient air cooling was added during work cycles and conditioned air cooling was delivered during rest periods. A compact battery-powered beltpack cooling unit (3.9 kg) designed and made at the U.S. Air Force Armstrong Laboratory was used to deliver 5.7 L/sec filtered ambient air during work cycles: 4.7 L/sec to the body and 1 L/sec to the face. Five experimental cycles were conducted in a thermally controlled chamber under warm conditions (32 degrees C, 40% relative humidity) with (1) no cooling-intermittent work, (2) intermittent cooling, (3) continuous cooling during intermittent exercise, and (4) no cooling-continuous work and (5) ambient air cooling during continuous exercise. Intermittent, conditioned, and continuous air cooling resulted in significant reductions in rectal temperature, mean skin temperature, and heart rate as compared with the no-cooling trials. The continuous air-cooling trial significantly improved thermal comfort and sweat evaporation. Results suggest that ambient air delivered during work cycles by a lightweight portable unit (in conjunction with conditioned air delivered during rest periods), can definitely improve personal comfort, reduce skin temperature, and decrease the cumulative fatigue common to repeated work/rest cycles in selected military and industrial applications in which individuals work in chemical defense ensembles.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scheele, Randall D.; Casella, Andrew M.

2010-09-28

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

NEUTRONIC REACTOR

DOEpatents

Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

1959-10-27

BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carbajo, Juan J; Qualls, A L

2016-01-01

INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a smallmore » version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.« less

Emergency core cooling system

DOEpatents

Schenewerk, William E.; Glasgow, Lyle E.

1983-01-01

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

Core cooling under accident conditions at the high-flux beam reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tichler, P.; Cheng, L.; Fauske, H.

The High-Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is cooled and moderated by heavy water and contains {sup 235}U in the form of narrow-channel, parallel-plate-type fuel elements. During normal operation, the flow direction is downward through the core. This flow direction is maintained at a reduced flow rate during routine shutdown and on loss of commercial power by means of redundant pumps and power supplies. However, in certain accident scenarios, e.g. loss-of-coolant accidents (LOCAs), all forced-flow cooling is lost. Although there was experimental evidence during the reactor design period (1958-1963) that the heat removal capacity in the fullymore » developed natural circulation cooling mode was relatively high, it was not possible to make a confident prediction of the heat removal capacity during the transition from downflow to natural circulation. Accordingly, a test program was initiated using an electrically heated section to simulate the fuel channel and a cooling loop to simulate the balance of the primary cooling system.« less

Correction of Temperatures of Air-Cooled Engine Cylinders for Variation in Engine and Cooling Conditions

NASA Technical Reports Server (NTRS)

Schey, Oscar W; Pinkel, Benjamin; Ellerbrock, Herman H , Jr

1939-01-01

Factors are obtained from semiempirical equations for correcting engine-cylinder temperatures for variation in important engine and cooling conditions. The variation of engine temperatures with atmospheric temperature is treated in detail, and correction factors are obtained for various flight and test conditions, such as climb at constant indicated air speed, level flight, ground running, take-off, constant speed of cooling air, and constant mass flow of cooling air. Seven conventional air-cooled engine cylinders enclosed in jackets and cooled by a blower were tested to determine the effect of cooling-air temperature and carburetor-air temperature on cylinder temperatures. The cooling air temperature was varied from approximately 80 degrees F. to 230 degrees F. and the carburetor-air temperature from approximately 40 degrees F. to 160 degrees F. Tests were made over a large range of engine speeds, brake mean effective pressures, and pressure drops across the cylinder. The correction factors obtained experimentally are compared with those obtained from the semiempirical equations and a fair agreement is noted.

Cyclone reactor with internal separation and axial recirculation

DOEpatents

Becker, F.E.; Smolensky, L.A.

1988-07-19

A cyclone combustor apparatus contains a circular partition plate containing a central circular aperture is described. The partition plate divides the apparatus into a cylindrical precombustor chamber and a combustor chamber. A coal-water slurry is passed axially into the inlet end of the precombustor chamber, and primary air is passed tangentially into said chamber to establish a cyclonic air flow. Combustion products pass through the partition plate aperture and into the combustor chamber. Secondary air may also be passed tangentially into the combustor chamber adjacent the partition plate to maintain the cyclonic flow. Flue gas is passed axially out of the combustor chamber at the outlet end and ash is withdrawn tangentially from the combustor chamber at the outlet end. A first mixture of flue gas and ash may be tangentially withdrawn from the combustor chamber at the outlet end and recirculated to the axial inlet of the precombustor chamber with the coal-water slurry. A second mixture may be tangentially withdrawn from the outlet end and passed to a heat exchanger for cooling. Cooled second mixture is then recirculated to the axial inlet of the precombustor chamber. In another embodiment a single cyclone combustor chamber is provided with both the recirculation streams of the first mixture and the second mixture. 10 figs.

Cyclone reactor with internal separation and axial recirculation

DOEpatents

Becker, Frederick E.; Smolensky, Leo A.

1989-01-01

A cyclone combustor apparatus contains a circular partition plate containing a central circular aperture. The partition plate divides the apparatus into a cylindrical precombustor chamber and a combustor chamber. A coal-water slurry is passed axially into the inlet end of the precombustor chamber, and primary air is passed tangentially into said chamber to establish a cyclonic air flow. Combustion products pass through the partition plate aperture and into the combustor chamber. Secondary air may also be passed tangentially into the combustor chamber adjacent the partition plate to maintain the cyclonic flow. Flue gas is passed axially out of the combustor chamber at the outlet end and ash is withdrawn tangentially from the combuston chamber at the outlet end. A first mixture of flue gas and ash may be tangentially withdrawn from the combustor chamber at the outlet end and recirculated to the axial inlet of the precombustor chamber with the coal-water slurry. A second mixture of flue gas and ash may be tangentially withdrawn from the outlet end of the combustor chamber and passed to a heat exchanger for cooling. Cooled second mixture is then recirculated to the axial inlet of the precombustor chamber. In another embodiment a single cyclone combustor chamber is provided with both the recirculation streams of the first mixture and the second mixture.

A 100-kWt NaK-Cooled Space Reactor Concept for an Early-Flight Mission

NASA Astrophysics Data System (ADS)

Poston, David I.

2003-01-01

A stainless-steel (SS) sodium-potassium (NaK) cooled reactor could potentially be the first step in utilizing fission technology in space. The sum of all system-level experience for liquid-metal-cooled space reactors has been with NaK, including the SNAP-10a, the only reactor ever launched by the US. This paper describes a 100-kWt NaK reactor, the NaK-100, which is designed to be developed with minimal technical risk. In additional to NaK technology heritage, the NaK-100 uses a proven fuel-form (SS/UO2) and is designed for simplified system integration and testing. The pins are placed within a solid SS prism, and the NaK flows in an annulus between the pins and the prism. The nuclear and thermal-hydraulic performance of the NaK-100 is presented, as well as the major differences between the NaK-100 and SNAP-10a.

PBF Cooling Tower. View from highbay roof of Reactor Building ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Cooling Tower. View from high-bay roof of Reactor Building (PER-620). Camera faces northwest. East louvered face has been installed. Inlet pipes protrude from fan deck. Two redwood vents under construction at top. Note piping, control, and power lines at sub-grade level in trench leading to Reactor Building. Photographer: Kirsh. Date: June 6, 1969. INEEL negative no. 69-3466 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Multi-Megawatt Power System Trade Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Longhurst, Glen Reed; Schnitzler, Bruce Gordon; Parks, Benjamin Travis

2001-11-01

As part of a larger task, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform a trade study comparing liquid-metal cooled reactors having Rankine power conversion systems with gas-cooled reactors having Brayton power conversion systems. This report summarizes the approach, the methodology, and the results of that trade study. Findings suggest that either approach has the possibility to approach the target specific mass of 3-5 kg/kWe for the power system, though it appears either will require improvements to achieve that. Higher reactor temperatures have the most potential for reducing the specific mass of gas-cooled reactors but domore » not necessarily have a similar effect for liquid-cooled Rankine systems. Fuels development will be the key to higher reactor operating temperatures. Higher temperature turbines will be important for Brayton systems. Both replacing lithium coolant in the primary circuit with gallium and replacing potassium with sodium in the power loop for liquid systems increase system specific mass. Changing the feed pump turbine to an electric motor in Rankine systems has little effect. Key technologies in reducing specific mass are high reactor and radiator operating temperatures, low radiator areal density, and low turbine/generator system masses. Turbine/generator mass tends to dominate overall power system mass for Rankine systems. Radiator mass was dominant for Brayton systems.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HXmore » channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.« less

Evaporative cooling enhanced cold storage system

DOEpatents

Carr, Peter

1991-01-01

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

Evaporative cooling enhanced cold storage system

DOEpatents

Carr, P.

1991-10-15

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

Nuclear propulsion apparatus with alternate reactor segments

DOEpatents

Szekely, Thomas

1979-04-03

1. Nuclear propulsion apparatus comprising: A. means for compressing incoming air; B. nuclear fission reactor means for heating said air; C. means for expanding a portion of the heated air to drive said compressing means; D. said nuclear fission reactor means being divided into a plurality of radially extending segments; E. means for directing a portion of the compressed air for heating through alternate segments of said reactor means and another portion of the compressed air for heating through the remaining segments of said reactor means; and F. means for further expanding the heated air from said drive means and the remaining heated air from said reactor means through nozzle means to effect reactive thrust on said apparatus.

Hybrid radiator cooling system

DOEpatents

France, David M.; Smith, David S.; Yu, Wenhua; Routbort, Jules L.

2016-03-15

A method and hybrid radiator-cooling apparatus for implementing enhanced radiator-cooling are provided. The hybrid radiator-cooling apparatus includes an air-side finned surface for air cooling; an elongated vertically extending surface extending outwardly from the air-side finned surface on a downstream air-side of the hybrid radiator; and a water supply for selectively providing evaporative cooling with water flow by gravity on the elongated vertically extending surface.

16 CFR Appendix H to Part 305 - Cooling Performance and Cost for Central Air Conditioners

Code of Federal Regulations, 2014 CFR

2014-01-01

... for Central Air Conditioners Manufacturer's rated cooling capacities (Btu's/hr.) Range of SEER's Low High Single Package Units Central Air Conditioners (Cooling Only): All capacities 10.6 16.5 Heat Pumps (Cooling Function): All capacities 10.6 16.0 Split System Units Central Air Conditioners (Cooling Only...

Fuel leak detection apparatus for gas cooled nuclear reactors

DOEpatents

Burnette, Richard D.

1977-01-01

Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

Physiological Responses to Microclimate Cooling Used By the Air Soldier Dressed at MOPP 4 in an Extreme Desert Condition: Effects of Six Configurations

DTIC Science & Technology

2012-12-01

USARIEM TECHNICAL REPORT PHYSIOLOGICAL RESPONSES TO MICROCLIMATE COOLING USED BY THE AIR SOLDIER DRESSED AT MOPP 4 IN AN...2012 2. REPORT TYPE 4. TITLE AND SUBTITLE 5a. CONTRACT NUMBER Physiological Responses to Microclimate Cooling Used By the Air Soldier 5b. GRANT... Microclimate Cooling System MCG HI – Air Warrior Microclimate Cooling Garment Used in High Cooling Configuration MCG LO - Air Warrior Microclimate

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, Terry L.

1993-01-01

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, T.L.

1993-10-19

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures.

System Study: Reactor Core Isolation Cooling 1998-2014

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schroeder, John Alton

2015-12-01

This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

Experimental investigation of a new method for advanced fast reactor shutdown cooling

NASA Astrophysics Data System (ADS)

Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

2017-07-01

We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

Fukushima Daiichi Nuclear Accident; based on the Final Report of Atomic Energy Society of Japan

NASA Astrophysics Data System (ADS)

Sekimura, Naoto

2014-09-01

The Atomic Energy Society of Japan (AESJ) published the Final Report of the AESJ Investigation Committee on Fukushima Daiichi NPS Accident in March 2014. The AESJ is responsible to identify the underlying root causes of the accident through technical surveys and analyses, and to offer solutions for nuclear safety. At the Fukushima Daiichi, Units 1 to 3, which were under operation, were automatically shut down at 14:46 on March 11, 2011 by the Tohoku District-off the Pacific Ocean Earthquake. About 50 minutes later, the tsunami flooded and destroyed the emergency diesel generators, the seawater cooling pumps, the electric wiring system and the DC power for Units 1, 2 and 4, resulting in loss of all power except for an air-cooled emergency diesel generator at Unit 6. Unit 3 lost all AC power, and later lost DC before dawn of March 13. Cooling the reactors and monitoring the results were heavily dependent on electricity for high-pressure water injection, depressurizing the reactor, low pressure water injection, and following continuous cooling. In Unit 3, for example, recent re-evaluation in August 2014 by TEPCO shows that no cooling water was injected into the reactor core region after 8 PM on March 12, leading to the fuel melting from 5:30 AM on March 13. Even though seawater was injected from fire engines afterwards, the rupture of pressure vessel was caused and the majority of melted fuel dropped into the containment vessel of Unit 3. The estimation of amount of radioactive materials such as Xe-133, I-131, Cs-137 and Cs-134, emitted to the environment from Units 1 to 3 is discussed in the presentation. Direct causes of the accident identified in the AESJ Report were, 1) inadequate tsunami measures, 2) inadequate severe accident management measures and 3) inadequate emergency response, post-accident management/mitigation, and recovery measures. These were caused by the following underlying factors, i.e., a) lack of awareness on the roles and responsibilities by experts, b) shortfalls in establishing safety measures and fostering safety awareness by utilities, c) lack of safety awareness by the regulatory body, d) inadequacies in attitude of learning from efforts and collaborations in the international community, and e) shortage of qualified personnel to ensure safety and inadequacies in organization and management framework.

Preliminary Design of Critical Function Monitoring System of PGSFR

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2015-07-01

A PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) is under development at Korea Atomic Energy Research Institute. A critical function monitoring system of the PGSFR is preliminarily studied. The functions of CFMS are to display critical plant variables related to the safety of the plant during normal and accident conditions and guide the operators corrective actions to keep the plant in a safe condition and mitigate the consequences of accidents. The minimal critical functions of the PGSFR are composed of reactivity control, reactor core cooling, reactor coolant system integrity, primary heat transfer system(PHTS) heat removal, sodium water reaction mitigation, radiation controlmore » and containment conditions. The variables and alarm legs of each critical function of the PGSFR are as follows; - Reactivity control: The variables of reactivity control function are power range neutron flux instrumentation, intermediate range neutron flux instrumentation, source range neutron flux instrumentation, and control rod bottom contacts. The alarm leg to display the reactivity controls consists of status of control drop malfunction, high post trip power and thermal reactivity addition. - Reactor core cooling: The variables are PHTS sodium level, hot pool temperature of PHTS, subassembly exit temperature, cold pool temperature of the PHTS, PHTS pump current, and PHTS pump breaker status. The alarm leg consists of high core delta temperature, low sodium level of the PHTS, high subassembly exit temperature, and low PHTS pump load. - Reactor coolant system integrity: The variables are PHTS sodium level, cover gas pressure, and safeguard vessel sodium level. The alarm leg is composed of low sodium level of PHTS, high cover gas pressure and high sodium level of the safety guard vessel. - PHTS heat removal: The variables are PHTS sodium level, hot pool temperature of PHTS, core exit temperature, cold pool temperature of the PHTS, flow rate of passive residual heat removal system, flow rate of active residual heat removal system, and temperatures of air heat exchanger temperature of residual heat removal systems. The alarm legs are composed of two legs of a 'passive residual heat removal system not cooling' and 'active residual heat removal system not cooling'. - Sodium water reaction mitigation: The variables are intermediate heat transfer system(IHTS) pressure, pressure and temperature and level of sodium dump tank, the status of rupture disk, hydrogen concentration in IHTS and direct variable of sodium-water-reaction measure. The alarm leg consists of high IHTS pressure, the status of sodium water reaction mitigation system and the indication of direct measure. - Radiation control: The variables are radiation of PHTS, radiation of IHTS, and radiation of containment purge. The alarm leg is composed of high radiation of PHTS and IHTS, and containment purge system. - Containment condition: The variables are containment pressure, containment isolation status, and sodium fire. The alarm leg consists of high containment pressure, status of containment isolation and status of sodium fire. (authors)« less

Development of advanced high heat flux and plasma-facing materials

NASA Astrophysics Data System (ADS)

Linsmeier, Ch.; Rieth, M.; Aktaa, J.; Chikada, T.; Hoffmann, A.; Hoffmann, J.; Houben, A.; Kurishita, H.; Jin, X.; Li, M.; Litnovsky, A.; Matsuo, S.; von Müller, A.; Nikolic, V.; Palacios, T.; Pippan, R.; Qu, D.; Reiser, J.; Riesch, J.; Shikama, T.; Stieglitz, R.; Weber, T.; Wurster, S.; You, J.-H.; Zhou, Z.

2017-09-01

Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating tungsten alloys, demonstrating favorable tungsten-like plasma-wall interaction behavior under normal operation conditions, are an intrinsic solution to otherwise catastrophic consequences of loss-of-coolant and air ingress events in a fusion reactor. Permeation barrier layers avoid the escape of tritium into structural and cooling materials, thereby minimizing the release of tritium under normal operation conditions. Finally, solutions for the unique bonding requirements of dissimilar material used in a fusion reactor are demonstrated by describing the current status and prospects of functionally graded materials.

Air-cooled, hydrogen-air fuel cell

NASA Technical Reports Server (NTRS)

Shelekhin, Alexander B. (Inventor); Bushnell, Calvin L. (Inventor); Pien, Michael S. (Inventor)

1999-01-01

An air-cooled, hydrogen-air solid polymer electrolyte (SPE) fuel cell with a membrane electrode assembly operatively associated with a fluid flow plate having at least one plate cooling channel extending through the plate and at least one air distribution hole extending from a surface of the cathode flow field into the plate cooling channel.

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

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

DOEpatents

Peterson, Per F.

2013-05-14

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

Floating Loop System For Cooling Integrated Motors And Inverters Using Hot Liquid Refrigerant

DOEpatents

Hsu, John S [Oak Ridge, TN; Ayers, Curtis W [Kingston, TN; Coomer, Chester [Knoxville, TN; Marlino, Laura D [Oak Ridge, TN

2006-02-07

A floating loop vehicle component cooling and air-conditioning system having at least one compressor for compressing cool vapor refrigerant into hot vapor refrigerant; at least one condenser for condensing the hot vapor refrigerant into hot liquid refrigerant by exchanging heat with outdoor air; at least one floating loop component cooling device for evaporating the hot liquid refrigerant into hot vapor refrigerant; at least one expansion device for expanding the hot liquid refrigerant into cool liquid refrigerant; at least one air conditioning evaporator for evaporating the cool liquid refrigerant into cool vapor refrigerant by exchanging heat with indoor air; and piping for interconnecting components of the cooling and air conditioning system.

Preparation of high temperature gas-cooled reactor fuel element

DOEpatents

Bradley, Ronnie A.; Sease, John D.

1976-01-01

This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

ETR AND MTR COMPLEXES IN CONTEXT. CAMERA FACING NORTHERLY. FROM ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR AND MTR COMPLEXES IN CONTEXT. CAMERA FACING NORTHERLY. FROM BOTTOM TO TOP: ETR COOLING TOWER, ELECTRICAL BUILDING AND LOW-BAY SECTION OF ETR BUILDING, HEAT EXCHANGER BUILDING (WITH U SHAPED YARD), COMPRESSOR BUILDING. MTR REACTOR SERVICES BUILDING IS ATTACHED TO SOUTH WALL OF MTR. WING A IS ATTACHED TO BALCONY FLOOR OF MTR. NEAR UPPER RIGHT CORNER OF VIEW IS MTR PROCESS WATER BUILDING. WING B IS AT FAR WEST END OF COMPLEX. NEAR MAIN GATE IS GAMMA FACILITY, WITH "COLD" BUILDINGS BEYOND: RAW WATER STORAGE TANKS, STEAM PLANT, MTR COOLING TOWER PUMP HOUSE AND COOLING TOWER. INL NEGATIVE NO. 56-4101. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Nuclear reactor insulation and preheat system

DOEpatents

Wampole, Nevin C.

1978-01-01

An insulation and preheat system for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the compartment. An external surface of the compartment or enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair.

A Comparison of Fission Power System Options for Lunar and Mars Surface Applications

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2006-01-01

This paper presents a comparison of reactor and power conversion design options for 50 kWe class lunar and Mars surface power applications with scaling from 25 to 200 kWe. Design concepts and integration approaches are provided for three reactor-converter combinations: gas-cooled Brayton, liquid-metal Stirling, and liquid-metal thermoelectric. The study examines the mass and performance of low temperature, stainless steel based reactors and higher temperature refractory reactors. The preferred system implementation approach uses crew-assisted assembly and in-situ radiation shielding via installation of the reactor in an excavated hole. As an alternative, self-deployable system concepts that use earth-delivered, on-board radiation shielding are evaluated. The analyses indicate that among the 50 kWe stainless steel reactor options, the liquid-metal Stirling system provides the lowest mass at about 5300 kg followed by the gas-cooled Brayton at 5700 kg and the liquid-metal thermoelectric at 8400 kg. The use of a higher temperature, refractory reactor favors the gas-cooled Brayton option with a system mass of about 4200 kg as compared to the Stirling and thermoelectric options at 4700 and 5600 kg, respectively. The self-deployed concepts with on-board shielding result in a factor of two system mass increase as compared to the in-situ shielded concepts.

Preliminary analysis of problem of determining experimental performance of air-cooled turbine II : methods for determining cooling-air-flow characteristics

NASA Technical Reports Server (NTRS)

Ellerbrock, Herman H , Jr

1950-01-01

In the determination of the performance of an air-cooled turbine, the cooling-air-flow characteristics between the root and the tip of the blades must be evaluated. The methods, which must be verified and the unknown functions evaluated, that are expected to permit the determination of pressure, temperature, and velocity through the blade cooling-air passages from specific investigation are presented.

DE-NE0008277_PROTEUS final technical report 2018

DOE Office of Scientific and Technical Information (OSTI.GOV)

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

Reactor core isolation cooling system

DOEpatents

Cooke, F.E.

1992-12-08

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

Reactor core isolation cooling system

DOEpatents

Cooke, Franklin E.

1992-01-01

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

COOLED NEUTRONIC REACTOR

DOEpatents

Binner, C.R.; Wilkie, C.B.

1958-03-18

This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

Temperature Mapping of Air Film-Cooled Thermal Barrier Coated Surfaces Using Phosphor Thermometry

NASA Technical Reports Server (NTRS)

Eldridge, Jeffrey I.

2016-01-01

While the effects of thermal barrier coating (TBC) thermal protection and air film cooling effectiveness for jet engine components are usually studied separately, their contributions to combined cooling effectiveness are interdependent and are not simply additive. Therefore, combined cooling effectiveness must be measured to achieve an optimum balance between TBC thermal protection and air film cooling. Phosphor thermometry offers several advantages for mapping temperatures of air film cooled surfaces. While infrared thermography has been typically applied to study air film cooling effectiveness, temperature accuracy depends on knowing surface emissivity (which may change) and correcting for effects of reflected radiation. Because decay time-based full-field phosphor thermometry is relatively immune to these effects, it can be applied advantageously to temperature mapping of air film-cooled TBC-coated surfaces. In this presentation, an overview will be given of efforts at NASA Glenn Research Center to perform temperature mapping of air film-cooled TBC-coated surfaces in a burner rig test environment. The effects of thermal background radiation and flame chemiluminescence on the measurements are investigated, and the strengths and limitations of this method for studying air film cooling effectiveness are discussed.

Cooling system with compressor bleed and ambient air for gas turbine engine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marsh, Jan H.; Marra, John J.

A cooling system for a turbine engine for directing cooling fluids from a compressor to a turbine blade cooling fluid supply and from an ambient air source to the turbine blade cooling fluid supply to supply cooling fluids to one or more airfoils of a rotor assembly is disclosed. The cooling system may include a compressor bleed conduit extending from a compressor to the turbine blade cooling fluid supply that provides cooling fluid to at least one turbine blade. The compressor bleed conduit may include an upstream section and a downstream section whereby the upstream section exhausts compressed bleed airmore » through an outlet into the downstream section through which ambient air passes. The outlet of the upstream section may be generally aligned with a flow of ambient air flowing in the downstream section. As such, the compressed air increases the flow of ambient air to the turbine blade cooling fluid supply.« less

In situ high temperature oxidation analysis of Zircaloy-4 using acoustic emission coupled with thermogravimetry

NASA Astrophysics Data System (ADS)

Omar, Al Haj; Véronique, Peres; Eric, Serris; François, Grosjean; Jean, Kittel; François, Ropital; Michel, Cournil

2015-06-01

Zircaloy-4 oxidation behavior at high temperature (900 °C), which can be reached in case of severe accidental situations in nuclear pressurised water reactor, was studied using acoustic emission analysis coupled with thermogravimetry. Two different atmospheres were used to study the oxidation of Zircaloy-4: (a) helium and pure oxygen, (b) helium and oxygen combined with slight addition of air. The experiments with 20% of oxygen confirm the dependence on oxygen anions diffusion in the oxide scale. Under a mixture of oxygen and air in helium, an acceleration of the corrosion was observed due to the detrimental effect of nitrogen. The kinetic rate increased significantly after a kinetic transition (breakaway). This acceleration was accompanied by an acoustic emission activity. Most of the acoustic emission bursts were recorded after the kinetic transition (post-transition) or during the cooling of the sample. The characteristic features of the acoustic emission signals appear to be correlated with the different populations of cracks and their occurrence in the ZrO2 layer or in the α-Zr(O) layer. Acoustic events were recorded during the isothermal dwell time at high temperature under air. They were associated with large cracks in the zirconia porous layer. Acoustic events were also recorded during cooling after oxidation tests both under air or oxygen. For the latter, cracks were observed in the oxygen enriched zirconium metal phase and not in the dense zirconia layer after 5 h of oxidation.

Dry-vault storage of spent fuel at the CASCAD facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baillif, L.; Guay, M.

A new modular dry storage vault concept using vertical metallic wells cooled by natural convection has been developed by the Commissariat a l'Energie Atomique and Societe Generale pour les Techniques Nouvelles to accommodate special fuels for high-level wastes. Basic specifications and design criteria have been followed to guarantee a double containment system and cooling to maintain the fuel below an acceptable temperature. The double containment is provided by two static barriers: At the reactor, fuels are placed in containers playing the role of the first barrier; the storage wells constitute the second barrier. Spent fuel placed in wells is cooledmore » by natural convection: a boundary layer is created along the outer side of the well. The heated air rises along the well leading to a thermosiphon flow that extracts the heat released. For heat transfer, studies, computations, and experimental tests have been carried out to calculate and determine the temperature of the containers and the fuel rod temperatures in various situations. The CASCAD vault storage can be applied to light water reactor (LWR) fuels without any difficulties if two requirements are satisfied: (1) Spend fuels have to be inserted in tight canisters. (2) Spent fuels have to be received only after a minimum decay time of 5 yr.« less

Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

NASA Astrophysics Data System (ADS)

Natesan, K.; Li, Meimei; Chopra, O. K.; Majumdar, S.

2009-07-01

Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Permana, Sidik; Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132; Sekimoto, Hiroshi

2010-12-23

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

Reactor safety method

DOEpatents

Vachon, Lawrence J.

1980-03-11

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

Breeder Reactors, Understanding the Atom Series.

ERIC Educational Resources Information Center

Mitchell, Walter, III; Turner, Stanley E.

The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

Performance characteristic of hybrid cooling system based on cooling pad and evaporator

NASA Astrophysics Data System (ADS)

Yoon, J. I.; Son, C. H.; Choi, K. H.; Kim, Y. B.; Sung, Y. H.; Roh, S. J.; Kim, Y. M.; Seol, S. H.

2018-01-01

In South Korea, most of domestic animals such as pigs and chickens might die due to thermal diseases if they are exposed to the high temperature consistently. In order to save them from the heat wave, numerous efforts have been carried out: installing a shade net, adjusting time of feeding, spraying mist and setting up a circulation fan. However, these methods have not shown significant improvements. Thus, this study proposes a hybrid cooling system combining evaporative cooler and air-conditioner in order to resolve the conventional problems caused by the high temperature in the livestock industry. The problem of cooling systems using evaporative cooling pads is that they are not effective for eliminating huge heat load due to their limited capacity. And, temperature of the supplied air cannot be low enough compared to conventional air-conditioning systems. On the other hand, conventional air-conditioning systems require relatively expensive installation cost, and high operating cost compared to evaporative cooling system. The hybrid cooling system makes up for the lack of cooling capacity of the evaporative cooler by employing the conventional air-conditioner. Additionally, temperature of supplied air can be lowered enough. In the hybrid cooling system, induced air by a fan is cooled by the evaporation of water in the cooling pad, and it is cooled again by an evaporator in the air-conditioner. Therefore, the more economical operation is possible due to additionally obtained cooling capacity from the cooling pads. Major results of experimental analysis of hybrid cooling system are as follows. The compressor power consumption of the hybrid cooling system is about 23% lower, and its COP is 17% higher than that of the conventional air-conditioners. Regarding the condition of changing ambient temperature, the total power consumption decreased by about 5% as the ambient temperature changed from 28.7°C to 31.7°C. Cooling capacity and COP also presented about 3% and 1% of minor difference at the same comparison condition.

Electronic Equipment Cold Plates

DTIC Science & Technology

1976-04-01

using fans or blowers to force the air through the cooling device». ^J^lJ^^i 5!°^ it ia ’«"^"o^ve. nonto«ic. nonfla—able, *nd possesses good ...Techniques GENERAL THERMAL CONTROL SYSTEMS AND THEIR REgUIREMENTS FLON DISTRIBUTION IN MANIFOLDS THE COLD PLATE IA ,! 1 3 S 12 15 33 32 32... IA .1 132 132 132 155 169 179 183 r (1) Air-Cooled Cold Plate No. (2) Air-Cooled Cold Plate No. (3) Air-cooled Cold Plate No. (4) Air-Cooled

Cooling circuit for steam and air-cooled turbine nozzle stage

DOEpatents

Itzel, Gary Michael; Yu, Yufeng

2002-01-01

The turbine vane segment includes inner and outer walls with a vane extending therebetween. The vane includes leading and trailing edge cavities and intermediate cavities. An impingement plate is spaced from the outer wall to impingement-cool the outer wall. Post-impingement cooling air flows through holes in the outer wall to form a thin air-cooling film along the outer wall. Cooling air is supplied an insert sleeve with openings in the leading edge cavity for impingement-cooling the leading edge. Holes through the leading edge afford thin-film cooling about the leading edge. Cooling air is provided the trailing edge cavity and passes through holes in the side walls of the vane for thin-film cooling of the trailing edge. Steam flows through a pair of intermediate cavities for impingement-cooling of the side walls. Post-impingement steam flows to the inner wall for impingement-cooling of the inner wall and returns the post-impingement cooling steam through inserts in other intermediate cavities for impingement-cooling the side walls of the vane.

PBF Reactor Building (PER620). Plot plan shows layout, including auxiliary ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Plot plan shows layout, including auxiliary buildings: Emergency Generator (621), Hose House (622), Cooling Tower Auxiliary (624), Maintenance and Storage Warehouse (625), Gas Cylinder Storage (627), Hose House (628), Cooling Tower (720), Substation (719), and other features. Road connections between PBF Reactor, its control building, and SPERT-I site. Note cable trenches along road to control building. Date: July 1965. Ebasco Services, PER-U-101. INEEL index no. 761-0100-00-205-123005 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Low exchange element for nuclear reactor

DOEpatents

Brogli, Rudolf H.; Shamasunder, Bangalore I.; Seth, Shivaji S.

1985-01-01

A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

Monitoring of tritium

DOEpatents

Corbett, James A.; Meacham, Sterling A.

1981-01-01

The fluid from a breeder nuclear reactor, which may be the sodium cooling fluid or the helium reactor-cover-gas, or the helium coolant of a gas-cooled reactor passes over the portion of the enclosure of a gaseous discharge device which is permeable to hydrogen and its isotopes. The tritium diffused into the discharge device is radioactive producing beta rays which ionize the gas (argon) in the discharge device. The tritium is monitored by measuring the ionization current produced when the sodium phase and the gas phase of the hydrogen isotopes within the enclosure are in equilibrium.

U.S./CIS eye joint nuclear rocket venture

NASA Technical Reports Server (NTRS)

Clark, John S.; Mcilwain, Melvin C.; Smetanikov, Vladimir; D'Yakov, Evgenij K.; Pavshuk, Vladimir A.

1993-01-01

An account is given of the significance for U.S. spacecraft development of a nuclear thermal rocket (NTR) reactor concept that has been developed in the (formerly Soviet) Commonwealth of Independent States (CIS). The CIS NTR reactor employs a hydrogen-cooled zirconium hydride moderator and ternary carbide fuels; the comparatively cool operating temperatures associated with this design promise overall robustness.

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2010 CFR

2010-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents SECTION I. Introduction. Section 50.34a provides that an application for a construction...

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2011 CFR

2011-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents SECTION I. Introduction. Section 50.34a provides that an application for a construction...

The low-power low-pressure flow resonance in a natural circulation cooled boiling water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hagen, T.H.J.J. van der; Stekelenburg, A.J.C.

1995-09-01

The last few years the possibility of flow resonances during the start-up phase of natural circulation cooled BWRs has been put forward by several authors. The present paper reports on actual oscillations observed at the Dodewaard reactor, the world`s only operating BWR cooled by natural circulation. In addition, results of a parameter study performed by means of a simple theoretical model are presented. The influence of relevant parameters on the resonance characteristics, being the decay ratio and the resonance frequency, is investigated and explained.

Apparatus for controlling nuclear core debris

DOEpatents

Jones, Robert D.

1978-01-01

Nuclear reactor apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling.

Flight Investigation of the Cooling Characteristics of a Two-Row Radial Engine Installation. 2 - Cooling-Air Pressure Recovery and Pressure Distribution

DTIC Science & Technology

1946-07-01

good distribution of cooling air, as well as minimum drag for the installation. The fact that these tests showed that the front recovery decreased...installations on engine cooling-air distribution indicates that good coin-elation of the cooling results of like engines in different installations...tests indicate that an important consider- ation in the design of cowlings and cowl flaps should be the obtaining of good distribution of cooling air

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yacout, A. M.; Billone, M. C.

2016-09-16

The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of datamore » were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.« less

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

76 FR 79229 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on January 19-20, 2012, 11545 Rockville... Cooling Systems for Light- Water Nuclear Power Reactors'' (Open)--The Committee will hear presentations by...

Radial blanket assembly orificing arrangement

DOEpatents

Patterson, J.F.

1975-07-01

A nuclear reactor core for a liquid metal cooled fast breeder reactor is described in which means are provided for increasing the coolant flow through the reactor fuel assemblies as the reactor ages by varying the coolant flow rate with the changing coolant requirements during the core operating lifetime. (auth)

Film cooling air pocket in a closed loop cooled airfoil

DOEpatents

Yu, Yufeng Phillip; Itzel, Gary Michael; Osgood, Sarah Jane; Bagepalli, Radhakrishna; Webbon, Waylon Willard; Burdgick, Steven Sebastian

2002-01-01

Turbine stator vane segments have radially inner and outer walls with vanes extending between them. The inner and outer walls are compartmentalized and have impingement plates. Steam flowing into the outer wall plenum passes through the impingement plate for impingement cooling of the outer wall upper surface. The spent impingement steam flows into cavities of the vane having inserts for impingement cooling the walls of the vane. The steam passes into the inner wall and through the impingement plate for impingement cooling of the inner wall surface and for return through return cavities having inserts for impingement cooling of the vane surfaces. To provide for air film cooing of select portions of the airfoil outer surface, at least one air pocket is defined on a wall of at least one of the cavities. Each air pocket is substantially closed with respect to the cooling medium in the cavity and cooling air pumped to the air pocket flows through outlet apertures in the wall of the airfoil to cool the same.

Hydronic rooftop cooling systems

DOEpatents

Bourne, Richard C [Davis, CA; Lee, Brian Eric [Monterey, CA; Berman, Mark J [Davis, CA

2008-01-29

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

Progressing batch hydrolysis process

DOEpatents

Wright, J.D.

1985-01-10

A progressive batch hydrolysis process is disclosed for producing sugar from a lignocellulosic feedstock. It comprises passing a stream of dilute acid serially through a plurality of percolation hydrolysis reactors charged with feed stock, at a flow rate, temperature and pressure sufficient to substantially convert all the cellulose component of the feed stock to glucose. The cooled dilute acid stream containing glucose, after exiting the last percolation hydrolysis reactor, serially fed through a plurality of pre-hydrolysis percolation reactors, charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the hemicellulose component of said feedstock to glucose. The dilute acid stream containing glucose is cooled after it exits the last prehydrolysis reactor.

Progressing batch hydrolysis process

DOEpatents

Wright, John D.

1986-01-01

A progressive batch hydrolysis process for producing sugar from a lignocellulosic feedstock, comprising passing a stream of dilute acid serially through a plurality of percolation hydrolysis reactors charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the cellulose component of the feedstock to glucose; cooling said dilute acid stream containing glucose, after exiting the last percolation hydrolysis reactor, then feeding said dilute acid stream serially through a plurality of prehydrolysis percolation reactors, charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the hemicellulose component of said feedstock to glucose; and cooling the dilute acid stream containing glucose after it exits the last prehydrolysis reactor.

Experimental feasibility study of radial injection cooling of three-pad radial air foil bearings

NASA Astrophysics Data System (ADS)

Shrestha, Suman K.

Air foil bearings use ambient air as a lubricant allowing environment-friendly operation. When they are designed, installed, and operated properly, air foil bearings are very cost effective and reliable solution to oil-free turbomachinery. Because air is used as a lubricant, there are no mechanical contacts between the rotor and bearings and when the rotor is lifted off the bearing, near frictionless quiet operation is possible. However, due to the high speed operation, thermal management is one of the very important design factors to consider. Most widely accepted practice of the cooling method is axial cooling, which uses cooling air passing through heat exchange channels formed underneath the bearing pad. Advantage is no hardware modification to implement the axial cooling because elastic foundation structure of foil bearing serves as a heat exchange channels. Disadvantage is axial temperature gradient on the journal shaft and bearing. This work presents the experimental feasibility study of alternative cooling method using radial injection of cooling air directly on the rotor shaft. The injection speeds, number of nozzles, location of nozzles, total air flow rate are important factors determining the effectiveness of the radial injection cooling method. Effectiveness of the radial injection cooling was compared with traditional axial cooling method. A previously constructed test rig was modified to accommodate a new motor with higher torque and radial injection cooling. The radial injection cooling utilizes the direct air injection to the inlet region of air film from three locations at 120° from one another with each location having three axially separated holes. In axial cooling, a certain axial pressure gradient is applied across the bearing to induce axial cooling air through bump foil channels. For the comparison of the two methods, the same amount of cooling air flow rate was used for both axial cooling and radial injection. Cooling air flow rate was referenced to the rotor surface speed for radial injection cooling. The mass flow rates for the radial injection were 0.032, 0.0432, 0.054 and 0.068 Kg/min, which result in average injection speed of 150, 200, 250 and 300% of rotor surface speed. Several thermocouples were attached at various circumferential directions of the bearing sleeve, two plenums, bearing holder and ball bearing housings to collect the temperature data of the bearing at 30krpm under 10lb of load. Both axial cooling and radial injection are effective cooling mechanism and effectiveness of both cooling methods is directly proportional to the total mass flow rates. However, axial cooling is slightly more efficient in controlling the average temperature of the bearing sleeve, but results in higher thermal gradient of the shaft along the axial direction and also higher thermal gradient of the bearing sleeve along the circumferential direction compared to the radial injection cooling. The smaller thermal gradient of the radial injection cooling is due to the direct cooling effect of the shaft by impinging jets. While the axial cooling has an effect on only the bearing, the radial injection has a cooling effect on both the bearing sleeve and shaft. It is considered the radial injection cooling needs to be further optimized in terms of number of injection holes and their locations.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yamada, K.; Aksan, S. N.

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,more » 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)« less

Draft environmental impact statement siting, construction, and operation of New Production Reactor capacity. Volume 4, Appendices D-R

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1991-04-01

This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build newmore » production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains 15 appendices.« less

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

Cooling System Design for PEM Fuel Cell Powered Air Vehicles

DTIC Science & Technology

2010-06-18

Research Laboratory (NRL) has developed a proton exchange membrane fuel cell ( PEMFC ) powered unmanned air vehicle (UAV) called the Ion Tiger. The Ion Tiger...to design a cooling system for the Ion Tiger and investigate cooling approaches that may be suitable for future PEMFC powered air vehicles. The...modifications) to other PEMFC systems utilizing a CHE for cooling. 18-06-2010 Memorandum Report Unmanned Air Vehicle UAV Fuel cell PEM Cooling Radiator January

Turbine airfoil with ambient cooling system

DOEpatents

Campbell, Jr, Christian X.; Marra, John J.; Marsh, Jan H.

2016-06-07

A turbine airfoil usable in a turbine engine and having at least one ambient air cooling system is disclosed. At least a portion of the cooling system may include one or more cooling channels configured to receive ambient air at about atmospheric pressure. The ambient air cooling system may have a tip static pressure to ambient pressure ratio of at least 0.5, and in at least one embodiment, may include a tip static pressure to ambient pressure ratio of between about 0.5 and about 3.0. The cooling system may also be configured such that an under root slot chamber in the root is large to minimize supply air velocity. One or more cooling channels of the ambient air cooling system may terminate at an outlet at the tip such that the outlet is aligned with inner surfaces forming the at least one cooling channel in the airfoil to facilitate high mass flow.

Cooling air recycling for gas turbine transition duct end frame and related method

DOEpatents

Cromer, Robert Harold; Bechtel, William Theodore; Sutcu, Maz

2002-01-01

A method of cooling a transition duct end frame in a gas turbine includes the steps of a) directing cooling air into the end frame from a region external of the transition duct and the impingement cooling sleeve; and b) redirecting the cooling air from the end frame into the annulus between the transition duct and the impingement cooling sleeve.

Seminar 14 - Desiccant Enhanced Air Conditioning: Desiccant Enhanced Evaporative Air Conditioning (Presentation)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kozubal, E.

2013-02-01

This presentation explains how liquid desiccant based coupled with an indirect evaporative cooler can efficiently produce cool, dry air, and how a liquid desiccant membrane air conditioner can efficiently provide cooling and dehumidification without the carryover problems of previous generations of liquid desiccant systems. It provides an overview to a liquid desiccant DX air conditioner that can efficiently provide cooling and dehumidification to high latent loads without the need for reheat, explains how liquid desiccant cooling and dehumidification systems can outperform vapor compression based air conditioning systems in hot and humid climates, explains how liquid desiccant cooling and dehumidification systemsmore » work, and describes a refrigerant free liquid desiccant based cooling system.« less

77 FR 64563 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on November 1-3, 2012, 11545 Rockville...-Term Core Cooling Approach for the Advanced Boiling Water Reactor (ABWR) Design for South Texas Project...

Development of toroid-type HTS DC reactor series for HVDC system

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-11-01

This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Correlation of the Characteristics of Single-Cylinder and Flight Engines in Tests of High-Performance Fuels in an Air-Cooled Engine I : Cooling Characteristics

NASA Technical Reports Server (NTRS)

Wilson, Robert W.; Richard, Paul H.; Brown, Kenneth D.

1945-01-01

Variable charge-air flow, cooling-air pressure drop, and fuel-air ration investigations were conducted to determine the cooling characteristics of a full-scale air-cooled single cylinder on a CUE setup. The data are compared with similar data that were available for the same model multicylinder engine tested in flight in a four-engine airplane. The cylinder-head cooling correlations were the same for both the single-cylinder and the flight engine. The cooling correlations for the barrels differed slightly in that the barrel of the single-cylinder engine runs cooler than the barrel of te flight engine for the same head temperatures and engine conditions.

Sequential cooling insert for turbine stator vane

DOEpatents

Jones, Russel B

2017-04-04

A sequential flow cooling insert for a turbine stator vane of a small gas turbine engine, where the impingement cooling insert is formed as a single piece from a metal additive manufacturing process such as 3D metal printing, and where the insert includes a plurality of rows of radial extending impingement cooling air holes alternating with rows of radial extending return air holes on a pressure side wall, and where the insert includes a plurality of rows of chordwise extending second impingement cooling air holes on a suction side wall. The insert includes alternating rows of radial extending cooling air supply channels and return air channels that form a series of impingement cooling on the pressure side followed by the suction side of the insert.

Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Middleton, Bobby; Pasch, James Jay; Kruizenga, Alan Michael

2016-01-01

This report outlines the thermodynamics of a supercritical carbon dioxide (sCO 2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO 2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related tomore » both Helium and to sCO 2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO 2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO 2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation« less

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

NASA Astrophysics Data System (ADS)

Godfroy, Thomas J.; Kapernick, Richard J.; Bragg-Sitton, Shannon M.

2004-02-01

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

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

NASA Astrophysics Data System (ADS)

Fratoni, Massimiliano

This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be transmuted; this burnup is slightly superior to that attainable in helium-cooled reactors. A preliminary analysis of the modular variant for the PB-AHTR investigated the triple heterogeneity of this design and determined its performance characteristics.

Gas-phase optical fiber photocatalytic reactors for indoor air application: a preliminary study on performance indicators

NASA Astrophysics Data System (ADS)

Palmiste, Ü.; Voll, H.

2017-10-01

The development of advanced air cleaning technologies aims to reduce building energy consumption by reduction of outdoor air flow rates while keeping the indoor air quality at an acceptable level by air cleaning. Photocatalytic oxidation is an emerging technology for gas-phase air cleaning that can be applied in a standalone unit or a subsystem of a building mechanical ventilation system. Quantitative information on photocatalytic reactor performance is required to evaluate the technical and economic viability of the advanced air cleaning by PCO technology as an energy conservation measure in a building air conditioning system. Photocatalytic reactors applying optical fibers as light guide or photocatalyst coating support have been reported as an approach to address the current light utilization problems and thus, improve the overall efficiency. The aim of the paper is to present a preliminary evaluation on continuous flow optical fiber photocatalytic reactors based on performance indicators commonly applied for air cleaners. Based on experimental data, monolith-type optical fiber reactor performance surpasses annular-type optical fiber reactors in single-pass removal efficiency, clean air delivery rate and operating cost efficiency.

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

NASA Astrophysics Data System (ADS)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

40 CFR 92.108 - Intake and cooling air measurements.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 20 2010-07-01 2010-07-01 false Intake and cooling air measurements....108 Intake and cooling air measurements. (a) Intake air flow measurement. Measurement of the flow rate..., the measurement technique shall conform to the following: (1) The air flow measurement method used...

Reflux cooling experiments on the NCSU scaled PWR facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doster, J.M.; Giavedoni, E.

1993-01-01

Under loss of forced circulation, coupled with the loss or reduction in primary side coolant inventory, horizontal stratified flows can develop in the hot and cold legs of pressurized water reactors (PWRs). Vapor produced in the reactor vessel is transported through the hot leg to the steam generator tubes where it condenses and flows back to the reactor vessel. Within the steam generator tubes, the flow regimes may range from countercurrent annular flow to single-phase convection. As a result, a number of heat transfer mechanisms are possible, depending on the loop configuration, total heat transfer rate, and the steam flowmore » rate within the tubes. These include (but are not limited to) two-phase natural circulation, where the condensate flows concurrent to the vapor stream and is transported to the cold leg so that the entire reactor coolant loop is active, and reflux cooling, where the condensate flows back down the interior of the coolant tubes countercurrent to the vapor stream and is returned to the reactor vessel through the hot leg. While operating in the reflux cooling mode, the cold leg can effectively be inactive. Heat transfer can be further influenced by noncondensables in the vapor stream, which accumulate within the upper regions of the steam generator tube bundle. In addition to reducing the steam generator's effective heat transfer area, under these conditions operation under natural circulation may not be possible, and reflux cooling may be the only viable heat transfer mechanism. The scaled PWR (SPWR) facility in the nuclear engineering department at North Carolina State Univ. (NCSU) is being used to study the effectiveness of two-phase natural circulation and reflux cooling under conditions associated with loss of forced circulation, midloop coolant levels, and noncondensables in the primary coolant system.« less

Air cooled turbine component having an internal filtration system

DOEpatents

Beeck, Alexander R [Orlando, FL

2012-05-15

A centrifugal particle separator is provided for removing particles such as microscopic dirt or dust particles from the compressed cooling air prior to reaching and cooling the turbine blades or turbine vanes of a turbine engine. The centrifugal particle separator structure has a substantially cylindrical body with an inlet arranged on a periphery of the substantially cylindrical body. Cooling air enters centrifugal particle separator through the separator inlet port having a linear velocity. When the cooling air impinges the substantially cylindrical body, the linear velocity is transformed into a rotational velocity, separating microscopic particles from the cooling air. Microscopic dust particles exit the centrifugal particle separator through a conical outlet and returned to a working medium.

High-Altitude Flight Cooling Investigation of a Radial Air-Cooled Engine

NASA Technical Reports Server (NTRS)

Manganiello, Eugene J; Valerino, Michael F; Bell, E Barton

1947-01-01

An investigation of the cooling of an 18-cylinder, twin-row, radial, air-cooled engine in a high-performance pursuit airplane has been conducted for variable engine and flight conditions at altitudes ranging from 5000 to 35,000 feet in order to provide a basis for predicting high-altitude cooling performance from sea-level or low altitude experimental results. The engine cooling data obtained were analyzed by the usual NACA cooling-correlation method wherein cylinder-head and cylinder-barrel temperatures are related to the pertinent engine and cooling-air variables. A theoretical analysis was made of the effect on engine cooling of the change of density of the cooling air across the engine (the compressibility effect), which becomes of increasing importance as altitude is increased. Good agreement was obtained between the results of the theoretical analysis and the experimental data.

Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Webster, K. L.

2007-01-01

Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

NASA Astrophysics Data System (ADS)

Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

2010-06-01

In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ross, Kyle W.; Gauntt, Randall O.; Cardoni, Jeffrey N.

2013-11-01

Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.

Safe Affordable Fission Engine-(SAFE-) 100a Heat Exchanger Thermal and Structural Analysis

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

A potential fission power system for in-space missions is a heat pipe-cooled reactor coupled to a Brayton cycle. In this system, a heat exchanger (HX) transfers the heat of the reactor core to the Brayton gas. The Safe Affordable Fission Engine- (SAFE-) 100a is a test program designed to thermally and hydraulically simulate a 95 Btu/s prototypic heat pipe-cooled reactor using electrical resistance heaters on the ground. This Technical Memorandum documents the thermal and structural assessment of the HX used in the SAFE-100a program.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takeda, T.; Shimazu, Y.; Hibi, K.

2012-07-01

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

Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

PubMed

Hill, R N; Nutt, W M; Laidler, J J

2011-01-01

The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. Copyright © 2010 Health Physics Society

Air film cooling in a nonadiabatic wall conical nozzle.

NASA Technical Reports Server (NTRS)

Boldman, D. R.; Papell, S. S.; Ehlers, R. C.

1972-01-01

Experimental data for an air-film cooled conical nozzle operating with a heated-air main stream and a water-cooled wall confirm the validity of Lieu's (1964) method for correlating film cooling data in the accelerated flow of a nonadiabatic-wall nozzle. The film cooling effectiveness modified for nonadiabatic walls by Lieu can be used to correlate film cooling under the condition that the main-stream to coolant velocity ratio at the slot is about 1. Such a ratio provides the optimum cooling effectiveness.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moriarty, M.P.

1993-01-15

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.

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

NASA Astrophysics Data System (ADS)

Moriarty, Michael P.

1993-01-01

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.

Modeling Chilled-Water Storage System Components for Coupling to a Small Modular Reactor in a Nuclear Hybrid Energy System

NASA Astrophysics Data System (ADS)

Misenheimer, Corey Thomas

The intermittency of wind and solar power puts strain on electric grids, often forcing carbonbased and nuclear sources of energy to operate in a load-follow mode. Operating nuclear reactors in a load-follow fashion is undesirable due to the associated thermal and mechanical stresses placed on the fuel and other reactor components. Various Thermal Energy Storage (TES) elements and ancillary energy applications can be coupled to nuclear (or renewable) power sources to help absorb grid instabilities caused by daily electric demand changes and renewable intermittency, thereby forming the basis of a candidate Nuclear Hybrid Energy System (NHES). During the warmer months of the year in many parts of the country, facility air-conditioning loads are significant contributors to the increase in the daily peak electric demand. Previous research demonstrated that a stratified chilled-water storage tank can displace peak cooling loads to off-peak hours. Based on these findings, the objective of this work is to evaluate the prospect of using a stratified chilled-water storage tank as a potential TES reservoir for a nuclear reactor in a NHES. This is accomplished by developing time-dependent models of chilled-water system components, including absorption chillers, cooling towers, a storage tank, and facility cooling loads appropriate for a large office space or college campus, as a callable FORTRAN subroutine. The resulting TES model is coupled to a high-fidelity mPower-sized Small Modular Reactor (SMR) Simulator, with the goal of utilizing excess reactor capacity to operate several sizable chillers in order to keep reactor power constant. Chilled-water production via single effect, lithium bromide (LiBr) absorption chillers is primarily examined in this study, although the use of electric chillers is briefly explored. Absorption chillers use hot water or low-pressure steam to drive an absorption-refrigeration cycle. The mathematical framework for a high-fidelity dynamic absorption chiller model is presented. The transient FORTRAN model is grounded on time-dependent mass, species, and energy conservation equations. Due to the vast computational costs of the high-fidelity model, a low-fidelity absorption chiller model is formulated and calibrated to mimic the behavior of the high-fidelity model. Stratified chilled-water storage tank performance is characterized using Computational Fluid Dynamics (CFD). The geometry employed in the CFD model represents a 5-million-gallon storage tank currently in use at a North Carolina college campus. Simulation results reveal the laminar numerical model most closely aligns with actual tank charging and discharging data. A subsequent parametric study corroborates storage tank behavior documented throughout literature and industry. Two absorption chiller configurations are considered. The first involves bypassing lowpressure steam from the low-pressure turbine to absorption chillers during periods of excess reactor capacity in order to keep reactor power constant. Simulation results show steam conditions downstream of the turbine control valves are a strong function of turbine load, and absorption chiller performance is hindered by reduced turbine impulse pressures at reduced turbine demands. A more suitable configuration entails integrating the absorption chillers into a flash vessel system that is thermally coupled to a sensible heat storage system. The sensible heat storage system is able to maintain reactor thermal output constant at 100% and match turbine output with several different electric demand profiles. High-pressure condensate in the sensible heat storage system is dropped across a let-down orifice and flashed in an ideal separator. Generated steam is sent to a bank of absorption chillers. Simulation results show enough steam is available during periods of reduced turbine demand to power four large absorption chillers to charge a 5-million-gallon stratified chilled-water storage tank, which is used to offset cooling loads in an adjacent facility. The coupled TES systems operating in conjunction with an SMR comprise the foundation of a tightly coupled NHES.

Catalog of experimental projects for a fissioning plasma reactor

NASA Technical Reports Server (NTRS)

Lanzo, C. D.

1973-01-01

Experimental and theoretical investigations were carried out to determine the feasibility of using a small scale fissioning uranium plasma as the power source in a driver reactor. The driver system is a light water cooled and moderated reactor of the MTR type. The eight experiments and proposed configurations for the reactor are outlined.

Application of Molten Salt Reactor Technology to Nuclear Electric Propulsion Mission

NASA Technical Reports Server (NTRS)

Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

Nuclear electric propulsion (NEP) and planetary surface power missions require reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional gas cooled, liquid metal, and heat pipe space reactors.

High-temperature synthesis of silica particles by the chloride method in the regime of counter flow jet quenching

NASA Astrophysics Data System (ADS)

Kartaev, E. V.; Emel'kin, V. A.; Aul'chenko, S. M.

2017-10-01

The experimental and numerical investigations of synthesis of silica (SiO2) nanoparticles from premixed gaseous silicon tetrachloride (SiCl4) and oxygen of dry air in the high-temperature nitrogen flow of plasma-chemical reactor have been carried out. The regime of counter flow jet quenching of high-temperature heterogeneous flow has been utilized. The latter provided a rapid cooling of silica particles under nonequilibrium conditions with substantial temperature gradients. Synthesized silica particles were amorphous, with surface-average size being about 28 nm. The results of numerical calculations are found to agree qualitatively with experimental data.

Propellant actuated nuclear reactor steam depressurization valve

DOEpatents

Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

1992-01-01

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

MERCHANT MARINE SHIP REACTOR

DOEpatents

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Merchant Marine Ship Reactor

DOEpatents

Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

1961-05-01

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hetrick, D.L.; Sowers, G.W.

1978-06-01

This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. Amore » list of variable names and a listing for BRENDA are included as appendices.« less

Engine management during NTRE start up

NASA Technical Reports Server (NTRS)

Bulman, Mel; Saltzman, Dave

1993-01-01

The topics are presented in viewgraph form and include the following: total engine system management critical to successful nuclear thermal rocket engine (NTRE) start up; NERVA type engine start windows; reactor power control; heterogeneous reactor cooling; propellant feed system dynamics; integrated NTRE start sequence; moderator cooling loop and efficient NTRE starting; analytical simulation and low risk engine development; accurate simulation through dynamic coupling of physical processes; and integrated NTRE and mission performance.

Thermionic nuclear reactor with internal heat distribution and multiple duct cooling

DOEpatents

Fisher, C.R.; Perry, L.W. Jr.

1975-11-01

A Thermionic Nuclear Reactor is described having multiple ribbon-like coolant ducts passing through the core, intertwined among the thermionic fuel elements to provide independent cooling paths. Heat pipes are disposed in the core between and adjacent to the thermionic fuel elements and the ribbon ducting, for the purpose of more uniformly distributing the heat of fission among the thermionic fuel elements and the ducts.

Scoping Calculations of Power Sources for Nuclear Electric Propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1994-01-01

This technical memorandum describes models and calculational procedures to fully characterize the nuclear island of power sources for nuclear electric propulsion. Two computer codes were written: one for the gas-cooled NERVA derivative reactor and the other for liquid metal-cooled fuel pin reactors. These codes are going to be interfaced by NASA with the balance of plant in order to make scoping calculations for mission analysis.

ETR, TRA642. EASTWEST SECTION, LOOKING NORTH. PATH OF COOLING WATER ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. EAST-WEST SECTION, LOOKING NORTH. PATH OF COOLING WATER PIPE TUNNEL. WORKING AND STORAGE CANAL. SUB-PILE ROOM. CONTROL ROD ACCESS ROOM. FLOOR NAMES. (THIS WAS A CONCEPT DRAWING.) KAISER ETR-5528-MTR-642-A-5, 11/1955. INL INDEX NO. 532-0642-00-486-100913. REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Vehicle cabin cooling system for capturing and exhausting heated boundary layer air from inner surfaces of solar heated windows

DOEpatents

Farrington, Robert B.; Anderson, Ren

2001-01-01

The cabin cooling system includes a cooling duct positioned proximate and above upper edges of one or more windows of a vehicle to exhaust hot air as the air is heated by inner surfaces of the windows and forms thin boundary layers of heated air adjacent the heated windows. The cabin cooling system includes at least one fan to draw the hot air into the cooling duct at a flow rate that captures the hot air in the boundary layer without capturing a significant portion of the cooler cabin interior air and to discharge the hot air at a point outside the vehicle cabin, such as the vehicle trunk. In a preferred embodiment, the cooling duct has a cross-sectional area that gradually increases from a distal point to a proximal point to the fan inlet to develop a substantially uniform pressure drop along the length of the cooling duct. Correspondingly, this cross-sectional configuration develops a uniform suction pressure and uniform flow rate at the upper edge of the window to capture the hot air in the boundary layer adjacent each window.

Application of the aqueous self-cooled blanket concept to fusion reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deutsch, L.; Steiner, D.; Embrechts, M.J.

1986-01-01

The development of a reliable, safe, and economically attractive tritium breeding blanket is an essential requirement in the path to commercial fusion power. The primary objective of the recently completed Blanket Comparison and Selection Study (BCSS) was to evaluate previously proposed concepts, and thereby identify a limited number of preferred options that would provide the focus for an R and D program. The water-cooled concepts in the BCSS scored relatively low. We consider it prudent that a promising water-cooled blanket concept be included in this program since nearly all power producing reactors currently rely on water technology. It is inmore » this context that we propose the novel water-cooled blanket concept described herein.« less

24 CFR 3280.714 - Appliances, cooling.

Code of Federal Regulations, 2013 CFR

2013-04-01

... refrigerating systems serving any air conditioning or comfort-cooling system installed in a manufactured home... Systems § 3280.714 Appliances, cooling. (a) Every air conditioning unit or a combination air conditioning... Conditioning and Air Source Unitary Heat Pump Equipment and certified by ARI or other nationally recognized...

24 CFR 3280.714 - Appliances, cooling.

Code of Federal Regulations, 2012 CFR

2012-04-01

... refrigerating systems serving any air conditioning or comfort-cooling system installed in a manufactured home... Systems § 3280.714 Appliances, cooling. (a) Every air conditioning unit or a combination air conditioning... Conditioning and Air Source Unitary Heat Pump Equipment and certified by ARI or other nationally recognized...

Analytical investigation of chord size and cooling methods on turbine blade cooling requirements. Book 1: Sections 1 through 8 and appendixes A through I

NASA Technical Reports Server (NTRS)

Faulkner, F. E.

1971-01-01

A study was conducted to determine the effect of chord size on air cooled turbine blades. In the preliminary design phase, eight turbine blade cooling configurations in 0.75-in., 1.0-in., and 1.5-in. chord sizes were analyzed to determine the maximum turbine inlet temperature capabilities. A pin fin convection cooled configuration and a film-impingement cooled configuration were selected for a final design analysis in which the maximum turbine inlet temperature was determined as a function of the cooling air inlet temperature and the turbine inlet total pressure for each of the three chord sizes. The cooling air flow requirements were also determined for a varying cooling air inlet temperature with a constant turbine inlet temperature. It was determined that allowable turbine inlet temperature increases with increasing chord for the convection cooled and transpiration cooled designs, however, the film-convection cooled designs did not have a significant change in turbine inlet temperature with chord.

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

DOEpatents

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

1982-08-19

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

Space Nuclear Reactor Engineering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poston, David Irvin

We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE PAGES

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.; ...

2017-02-26

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

10 CFR 431.92 - Definitions concerning commercial air conditioners and heat pumps.

Code of Federal Regulations, 2014 CFR

2014-01-01

... measurement. Commercial package air-conditioning and heating equipment means air-cooled, water-cooled... Conditioner means a basic model of commercial package air-conditioning and heating equipment (packaged or split) that is: Used in computer rooms, data processing rooms, or other information technology cooling...

40 CFR 63.107 - Identification of process vents subject to this subpart.

Code of Federal Regulations, 2012 CFR

2012-07-01

... process vents associated with an air oxidation reactor, distillation unit, or reactor that is in a source.... (b) Some, or all, of the gas stream originates as a continuous flow from an air oxidation reactor... specified in paragraphs (c)(1) through (3) of this section. (1) Is directly from an air oxidation reactor...

Industrial stator vane with sequential impingement cooling inserts

DOEpatents

Jones, Russell B; Fedock, John A; Goebel, Gloria E; Krueger, Judson J; Rawlings, Christopher K; Memmen, Robert L

2013-08-06

A turbine stator vane for an industrial engine, the vane having two impingement cooling inserts that produce a series of impingement cooling from the pressure side to the suction side of the vane walls. Each insert includes a spar with a row of alternating impingement cooling channels and return air channels extending in a radial direction. Impingement cooling plates cover the two sides of the insert and having rows of impingement cooling holes aligned with the impingement cooling channels and return air openings aligned with the return air channel.

A Review of Gas-Cooled Reactor Concepts for SDI Applications

DTIC Science & Technology

1989-08-01

710 program .) Wire- Core Reactor (proposed by Rockwell). The wire- core reactor utilizes thin fuel wires woven between spacer wires to form an open...reactor is based on results of developmental studies of nuclear rocket propulsion systems. The reactor core is made up of annular fuel assemblies of...XE Addendum to Volume II. NERVA Fuel Development , Westinghouse Astronuclear Laboratory, TNR-230, July 15’ 1972. J I8- Rover Program Reactor Tests

SCW Pressure-Channel Nuclear Reactor Some Design Features

NASA Astrophysics Data System (ADS)

Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

139. ARAIII Index of drwaings of gascooled reactor experiment buildings. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

139. ARA-III Index of drwaings of gas-cooled reactor experiment buildings. Aerojet-general 880-area/GCRE-100. Date: February 1958. Ineel index code no. 063-9999-80-013-102505. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

ETR COOLING TOWER. PUMP HOUSE (TRA645) IN SHADOW OF TOWER ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR COOLING TOWER. PUMP HOUSE (TRA-645) IN SHADOW OF TOWER ON LEFT. AT LEFT OF VIEW, HIGH-BAY BUILDING IS ETR. ONE STORY ATTACHMENT IS ETR ELECTRICAL BUILDING. STACK AT RIGHT IS ETR STACK; MTR STACK IS TOWARD LEFT. CAMERA FACING NORTHEAST. INL NEGATIVE NO. 56-3799. Jack L. Anderson, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MEANS FOR SHIELDING AND COOLING REACTORS

DOEpatents

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1959-02-10

Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

NASA Astrophysics Data System (ADS)

Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

2010-12-01

Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

Cooling Characteristics of an Experimental Tail-pipe Burner with an Annular Cooling-air Passage

NASA Technical Reports Server (NTRS)

Kaufman, Harold R; Koffel, William K

1952-01-01

The effects of tail-pipe fuel-air ratio (exhaust-gas temperatures from approximately 3060 degrees to 3825 degrees R), radial distributiion of tail-pipe fuel flow, and mass flow of combustion gas and the inside wall were determined for an experimental tail-pipe burner cooled by air flowing through and insulated cooling-air to combustion gas mass flow from 0.066 to 0.192 were also determined.

Numerical modelling of heat and mass transfer in adsorption solar reactor of ammonia on active carbon

NASA Astrophysics Data System (ADS)

Aroudam, El. H.

In this paper, we present a modelling of the performance of a reactor of a solar cooling machine based carbon-ammonia activated bed. Hence, for a solar radiation, measured in the Energetic Laboratory of the Faculty of Sciences in Tetouan (northern Morocco), the proposed model computes the temperature distribution, the pressure and the ammonia concentration within the activated carbon bed. The Dubinin-Radushkevich formula is used to compute the ammonia concentration distribution and the daily cycled mass necessary to produce a cooling effect for an ideal machine. The reactor is heated at a maximum temperature during the day and cool at the night. A numerical simulation is carried out employing the recorded solar radiation data measured locally and the daily ambient temperature for the typical clear days. Initially the reactor is at ambient temperature, evaporating pressure; Pev=Pst(Tev=0 ∘C) and maintained at uniform concentration. It is heated successively until the threshold temperature corresponding to the condensing pressure; Pcond=Pst(Tam) (saturation pressure at ambient temperature; in the condenser) and until a maximum temperature at a constant pressure; Pcond. The cooling of the reactor is characterised by a fall of temperature to the minimal values at night corresponding to the end of a daily cycle. We use the mass balance equations as well as energy equation to describe heat and mass transfer inside the medium of three phases. A numerical solution of the obtained non linear equations system based on the implicit finite difference method allows to know all parameters characteristic of the thermodynamic cycle and consider principally the daily evolution of temperature, ammonia concentration for divers positions inside the reactor. The tube diameter of the reactor shows the dependence of the optimum value on meteorological parameters for 1 m2 of collector surface.

Methanation assembly using multiple reactors

DOEpatents

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24

A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

Method and apparatus for cold gas reinjection in through-flow and reverse-flow wave rotors

NASA Technical Reports Server (NTRS)

Nalim, M. Razi (Inventor); Paxson, Daniel E. (Inventor)

1999-01-01

A method and apparatus for cold gas reinjection in through-flow and reverse-flow wave rotors having a plurality of channels formed around a periphery thereof. A first port injects a supply of cool air into the channels. A second port allows the supply of cool air to exit the channels and flow to a combustor. A third port injects a supply of hot gas from the combustor into the channels. A fourth port allows the supply of hot gas to exit the channels and flow to a turbine. A diverting port and a reinjection port are connected to the second and third ports, respectively. The diverting port diverts a portion of the cool air exiting through the second port as reinjection air. The diverting port is fluidly connected to the reinjection port which reinjects the reinjection air back into the channels. The reinjection air evacuates the channels of the hot gas resident therein and cools the channel walls, a pair of end walls of the rotor, ducts communicating with the rotor and subsequent downstream components. In a second embodiment, the second port receives all of the cool air exiting the channels and the diverting port diverts a portion of the cool air just prior to the cool air flowing to the combustor.

Rotor bore and turbine rotor wheel/spacer heat exchange flow circuit

DOEpatents

Caruso, Philip M.; Eldrid, Sacheverel Quentin; Ladhani, Azad A.; DeMania, Alan Richard; Palmer, Gene David; Wilson, Ian David; Rathbun, Lisa Shirley; Akin, Robert Craig

2002-01-01

In a turbine having closed-circuit steam-cooling passages about the rim of the rotor during steady-state operation, compressor discharge air is supplied to the rotor bore for passage radially outwardly into the wheel space cavities between the wheels and spacers. Communicating slots and channels in the spacers and wheels at circumferentially spaced positions enable egress of the compressor discharge air into the hot gas flow path. At turbine startup, cooling air flows through the closed-circuit steam passages to cool the outer rim of the rotor while compressor discharge air pre-warms the wheels and spacers. At steady-state, cooling steam is supplied in the closed-circuit steam-cooling passages and compressor discharge air is supplied through the bore and into the wheel space cavities to cool the rotor.

Cool Flame Quenching

NASA Technical Reports Server (NTRS)

Pearlman, Howard; Chapek, Richard

2001-01-01

Cool flame quenching distances are generally presumed to be larger than those associated with hot flames, because the quenching distance scales with the inverse of the flame propagation speed, and cool flame propagation speeds are often times slower than those associated with hot flames. To date, this presumption has never been put to a rigorous test, because unstirred, non-isothermal cool flame studies on Earth are complicated by natural convection. Moreover, the critical Peclet number (Pe) for quenching of cool flames has never been established and may not be the same as that associated with wall quenching due to conduction heat loss in hot flames, Pe approx. = 40-60. The objectives of this ground-based study are to: (1) better understand the role of conduction heat loss and species diffusion on cool flame quenching (i.e., Lewis number effects), (2) determine cool flame quenching distances (i.e, critical Peclet number, Pe) for different experimental parameters and vessel surface pretreatments, and (3) understand the mechanisms that govern the quenching distances in premixtures that support cool flames as well as hot flames induced by spark-ignition. Objective (3) poses a unique fire safety hazard if conditions exist where cool flame quenching distances are smaller than those associated with hot flames. For example, a significant, yet unexplored risk, can occur if a multi-stage ignition (a cool flame that transitions to a hot flame) occurs in a vessel size that is smaller than that associated with the hot quenching distance. To accomplish the above objectives, a variety of hydrocarbon-air mixtures will be tested in a static reactor at elevated temperature in the laboratory (1g). In addition, reactions with chemical induction times that are sufficiently short will be tested aboard NASA's KC-135 microgravity (mu-g) aircraft. The mu-g results will be compared to a numerical model that includes species diffusion, heat conduction, and a skeletal kinetic mechanism, following the work on diffusion-controlled cool flames by Fairlie et,al., 2000.

ETR, TRA642. ETR COMPLEX NEARLY COMPLETE. CAMERA FACES NORTHWEST, PROBABLY ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. ETR COMPLEX NEARLY COMPLETE. CAMERA FACES NORTHWEST, PROBABLY FROM TOP DECK OF COOLING TOWER. SHADOW IS CAST BY COOLING TOWER UNITS OFF LEFT OF VIEW. HIGH-BAY REACTOR BUILDING IS SURROUNDED BY ITS ATTACHED SERVICES: ELECTRICAL (TRA-648), HEAT EXCHANGER (TRA-644 WITH U-SHAPED YARD), AND COMPRESSOR (TRA-643). THE CONTROL BUILDING (TRA-647) ON THE NORTH SIDE IS HIDDEN FROM VIEW. AT UPPER RIGHT IS MTR BUILDING, TRA-603. INL NEGATIVE NO. 56-3798. Jack L. Anderson, Photographer, 11/26/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anthony L. Alberti; Todd S. Palmer; Javier Ortensi

2016-05-01

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately modelmore » the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.« less

Top shield temperatures, C and K Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Agar, J.D.

1964-12-28

A modification program is now in progress at the C and K Reactors consisting of an extensive renovation of the graphite channels in the vertical safety rod ststems. The present VSR channels are being enlarged by a graphite coring operation and channel sleeves will be installed in the larger channels. One problem associated with the coring operation is the danger of damaging top thermal shield cooling tubes located close to the VSR channels to such an extent that these tubes will have to be removed from service. If such a condition should exist at one or a number of locationsmore » in the top shield of the reactors after reactor startup, the question remains -- what would the resulting temperatures be of the various components of the top shields? This study was initiated to determine temperature distributions in the top shield complex at the C and K Reactors for various top thermal shield coolant system conditions. Since the top thermal shield cooling system at C Reactor is different than those at the K Reactors, the study was conducted separately for the two different systems.« less

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

NASA Astrophysics Data System (ADS)

Guidez, Joel; Saturnin, Anne

2017-11-01

During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

STUDY PROGRAM FOR TURBO-COOLER FOR PRODUCING ENGINE COOLING AIR.

DTIC Science & Technology

VANES , STAGNATION POINT, DECELERATION, ACCELERATION, SUPERSONIC DIFFUSERS, TURBINE BLADES , EVAPOTRANSPIRATION, LIQUID COOLED, HEAT TRANSFER, GAS BEARINGS, SEALS...HYPERSONIC AIRCRAFT , COOLING + VENTILATING EQUIPMENT), (*GAS TURBINES , COOLING + VENTILATING EQUIPMENT), HYPERSONIC FLOW, AIR COOLED, AIRCRAFT ... ENGINES , FEASIBILITY STUDIES, PRESSURE, SUPERSONIC CHARACTERISTICS, DESIGN, HEAT EXCHANGERS, COOLING (U) AXIAL FLOW TURBINES , DUCT INLETS, INLET GUIDE

Dynamic System Simulation of the KRUSTY Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klein, Steven Karl; Kimpland, Robert Herbert

2016-05-09

The proposed KRUSTY experiment is a demonstration of a reactor operating at power. The planned experimental configuration includes a highly enriched uranium (HEU) reflected core, cooled by multiple heat pipes leading to Stirling engines for primary heat rejection. Operating power is expected to be approximately four (4) to five (5) kilowatts with a core temperature above 1,000 K. No data is available on any historical reactor employing HEU metal that operated over the temperature range required for the KRUSTY experiment. Further, no reactor has operated with heat pipes as the primary cooling mechanism. Historic power reactors have employed either naturalmore » or forced convection so data on their operation is not directly applicable to the KRUSTY experiment. The primary purpose of the system model once developed and refined by data from these component experiments, will be used to plan the KRUSTY experiment. This planning will include expected behavior of the reactor from start-up, through various transient conditions where cooling begins to become present and effective, and finally establishment of steady-state. In addition, the model can provide indicators of anticipated off-normal events and appropriate operator response to those conditions. This information can be used to develop specific experiment operating procedures and aids to guide the operators in conduct of the experiment.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bragg-Sitton, S.M.; Propulsion Research Center, NASA Marshall Space Flight Center, Huntsville, AL 35812; Kapernick, R.

2004-02-04

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

Review of Nuclear Thermal Propulsion Ground Test Options

NASA Technical Reports Server (NTRS)

Coote, David J.; Power, Kevin P.; Gerrish, Harold P.; Doughty, Glen

2015-01-01

High efficiency rocket propulsion systems are essential for humanity to venture beyond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rockets with relatively high thrust and twice the efficiency of highest performing chemical propellant engines. NTP utilizes the coolant of a nuclear reactor to produce propulsive thrust. An NTP engine produces thrust by flowing hydrogen through a nuclear reactor to cool the reactor, heating the hydrogen and expelling it through a rocket nozzle. The hot gaseous hydrogen is nominally expected to be free of radioactive byproducts from the nuclear reactor; however, it has the potential to be contaminated due to off-nominal engine reactor performance. NTP ground testing is more difficult than chemical engine testing since current environmental regulations do not allow/permit open air testing of NTP as was done in the 1960's and 1970's for the Rover/NERVA program. A new and innovative approach to rocket engine ground test is required to mitigate the unique health and safety risks associated with the potential entrainment of radioactive waste from the NTP engine reactor core into the engine exhaust. Several studies have been conducted since the ROVER/NERVA program in the 1970's investigating NTP engine ground test options to understand the technical feasibility, identify technical challenges and associated risks and provide rough order of magnitude cost estimates for facility development and test operations. The options can be divided into two distinct schemes; (1) real-time filtering of the engine exhaust and its release to the environment or (2) capture and storage of engine exhaust for subsequent processing.

Cooling Effectiveness Measurements for Air Film Cooling of Thermal Barrier Coated Surfaces in a Burner Rig Environment Using Phosphor Thermometry

NASA Technical Reports Server (NTRS)

Eldridge, Jeffrey I.; Shyam, Vikram; Wroblewski, Adam C.; Zhu, Dongming; Cuy, Michael D.; Wolfe, Douglas E.

2016-01-01

While the effects of thermal barrier coating (TBC) thermal protection and air film cooling effectiveness are usually studied separately, their contributions to combined cooling effectiveness are interdependent and are not simply additive. Therefore, combined cooling effectiveness must be measured to achieve an optimum balance between TBC thermal protection and air film cooling. In this investigation, surface temperature mapping was performed using recently developed Cr-doped GdAlO3 phosphor thermometry. Measurements were performed in the NASA GRC Mach 0.3 burner rig on a TBC-coated plate using a scaled up cooling hole geometry where both the mainstream hot gas temperature and the blowing ratio were varied. Procedures for surface temperature and cooling effectiveness mapping of the air film-cooled TBC-coated surface are described. Applications are also shown for an engine component in both the burner rig test environment as well as an engine afterburner environment. The effects of thermal background radiation and flame chemiluminescence on the measurements are investigated, and advantages of this method over infrared thermography as well as the limitations of this method for studying air film cooling are discussed.

In Situ Measurements of Spectral Emissivity of Materials for Very High Temperature Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

G. Cao; S. J. Weber; S. O. Martin

2011-08-01

An experimental facility for in situ measurements of high-temperature spectral emissivity of materials in environments of interest to the gas-cooled very high temperature reactor (VHTR) has been developed. The facility is capable of measuring emissivities of seven materials in a single experiment, thereby enhancing the accuracy in measurements due to even minor systemic variations in temperatures and environments. The system consists of a cylindrical silicon carbide (SiC) block with seven sample cavities and a deep blackbody cavity, a detailed optical system, and a Fourier transform infrared spectrometer. The reliability of the facility has been confirmed by comparing measured spectral emissivitiesmore » of SiC, boron nitride, and alumina (Al2O3) at 600 C against those reported in literature. The spectral emissivities of two candidate alloys for VHTR, INCONEL{reg_sign} alloy 617 (INCONEL is a registered trademark of the Special Metals Corporation group of companies) and SA508 steel, in air environment at 700 C were measured.« less

Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

NASA Astrophysics Data System (ADS)

Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

2018-02-01

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

[Comparison of shaft temperature related treatment efficacy between "air-cooled" microwave coagulation and traditional microwave coagulation].

PubMed

Zheng, Yun; Li, Jin-Qing; Chen, Min-Shan; Zhang, Yao-Jun; Zhang, Ya-Qi

2004-11-01

The application and development of traditional percutaneous microwave coagulation therapy (PMCT) has been limited by high shaft temperature. The "air-cooled" PMCT is the newest advancement. This study was to compare shaft temperature related treatment efficacy between "air-cooled" PMCT and traditional PMCT. Two pigs underwent traditional PMCT, and "air-cooled" PMCT at 80 W for 10 min separately. Skin injury, surface temperature of guide-needle, charring tissue sticking to the shaft, and lesion shape in 2 pigs were compared. Five patients with liver tumor received traditional PMCT, and 8 patients with liver tumor received "air-cooled" PMCT. Feeling of pain, skin injury, charring tissue sticking to the shaft, local therapeutic efficacy, and recurrence of these 2 groups of patients were compared. In the pig underwent traditional PMCT, surface temperature of guide-needle reached 119-160 Centigrade; skin burn around puncture points was serious; charring tissue stuck to the front of electrodes; a trail sign was observed in coagulated lesion. In the pig underwent "air-cooled" PMCT, surface temperature of guide-needle was 28.8-39.9 Centigrade; no skin injury was found around puncture points; no charring tissue stuck to the front of electrodes; no obvious trail sign was observed in coagulated lesion. In 5 patients received traditional PMCT, 3 had skin injury; 2 had charring tissue stuck to the front of electrode; all felt moderate or serious epigastric pain lasted for 1-8 weeks; 4 had complete coagulation; 1 had local recurrence. In 8 patients received "air-cooled" PMCT, no one had skin injury, and charring tissue stuck to "air-cooled" electrode; 4 felt slight epigastric pain within 1 week; all had complete coagulation; no local recurrence was found. The technique of "air-cooled" electrode may decrease temperature of shaft safely and reliably, and eliminate side effects arose from high temperature of shaft. Treatment efficacy of "air-cooled" PMCT is better than that of traditional PMCT.

The 'fine line' of heat rejection.

PubMed

Carruthers, Phillip

2010-09-01

Selection of heat rejection equipment has traditionally entailed a choice between the higher energy consumption of an air-cooled solution, and the high water consumption of a water-cooled solution. This paper examines advancement in heat rejection technology and the way it can be applied to air conditioning and refrigeration plant in healthcare and other facilities. It also examines field difficulties encountered in pipework design as the knowledge and experience levels of engineers designing systems with remote condensers diminish. With plant larger than 1,000 kW, the only option previously has been water-cooled solutions using an array of cooling towers, or perhaps an evaporative condenser, since air-cooled plant involved massive volumes of chemical refrigerant, which posed a problem ecologically. An additional hurdle was problems associated with limitations on pipe lengths for refrigeration plant. The advent of adiabatically pre-cooled closed circuit coolers and air-cooled condensers has introduced an alternative to cooling towers that offers the potential for "water-cooled performance" from an air-cooled solution with no serious threat of Legionella contamination. However, each application needs to be considered on a case-by-case basis. The paper examines, in detail, the impact of adiabatic pre-cooling, with recent examples of its application in sub-tropical Brisbane providing evidence of the potential performance achievable.

Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

NASA Astrophysics Data System (ADS)

Fic, Adam; Składzień, Jan; Gabriel, Michał

2015-03-01

Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Per; Greenspan, Ehud

2015-02-09

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. This report also includes results for additional studies relevant to the design and analysis of pebble bed reactor cores including the study of forces on shut down blades inserted directly into a packed bed and pebble flow in a cylindrical hopper that is representative of a small test reactor.« less

Compressor ported shroud for foil bearing cooling

DOEpatents

Elpern, David G [Los Angeles, CA; McCabe, Niall [Torrance, CA; Gee, Mark [South Pasadena, CA

2011-08-02

A compressor ported shroud takes compressed air from the shroud of the compressor before it is completely compressed and delivers it to foil bearings. The compressed air has a lower pressure and temperature than compressed outlet air. The lower temperature of the air means that less air needs to be bled off from the compressor to cool the foil bearings. This increases the overall system efficiency due to the reduced mass flow requirements of the lower temperature air. By taking the air at a lower pressure, less work is lost compressing the bearing cooling air.

The effects of cooling systems on CO2-lased human enamel.

PubMed

Lian, H J; Lan, W H; Lin, C P

1996-12-01

The thermal effects on dentin during CO2 laser irradiation on human enamel were investigated. To simulate the clinical practice, two cooling methods (air and water spray) were applied immediately after laser exposure, whereas one group without cooling was served as control. Three hundred and sixty uniform tooth blocks were obtained from freshly extracted human third molars. Temperature change measurements were made via electrical thermocouple implanted within the tooth block 2 mm away from the enamel surface. Experimental treatments consisted of lasing without cooling, lasing with 0.5-ml/sec water cooling, and lasing with 15-psi air cooling. Our results indicated that (1) both air- and water-cooling groups could reduce temperature elevation significantly; (2) the larger power energy resulted in the higher temperature elevation. In conclusion, for CO2 laser irradiation on human enamel both water- and air-cooling methods may be effective on prevention of thermal damage of pulp.

10 CFR 50.55a - Codes and standards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... specified in § 50.55, except that each combined license for a boiling or pressurized water-cooled nuclear... boiling or pressurized water-cooled nuclear power facility is subject to the conditions in paragraphs (f... performed. (2) Systems and components of boiling and pressurized water-cooled nuclear power reactors must...

77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-15

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core... acceptable to implement with regard to initial testing features of emergency core cooling systems (ECCSs) for...

Closed-loop air cooling system for a turbine engine

DOEpatents

North, William Edward

2000-01-01

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

BOILING REACTORS

DOEpatents

Untermyer, S.

1962-04-10

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

Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

NASA Astrophysics Data System (ADS)

Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

2012-06-01

Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

Reactor vessel lower head integrity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rubin, A.M.

1997-02-01

On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) inmore » this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.« less

A SURVEY OF CONVENTIONAL STEAM BOILER EXPERIENCE APPLICABLE TO THE HTGR STEAM GENERATORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paget, J.A.

1959-10-01

BS>The steam generator of a high temperature gas-cooled reactor consists of tubular heating surface inside a shell which forms part of the primary He circuit of the reactor. When a tube fails in such a steam generator, moisture in the form of steam is released into the He steam and is carried through the reactor where it will cause corrosion and mass transfer of C in the core. A paramount consideration in the design of a steam generator for a high temperature gas-cooled reactor is the prevention of tube failures. Preference, therefore, should be given to a forced circulation design.more » The Loeffler Boiler would be the best from this standpoint alone since only steam enters the tubes, and its circulation rate can be maintained at an adequate value to insure cool tubes regardless of load fluctuations. The next type in the order of preference would be the forced recirculation boiler, since at least the boiier tubes always have an adequate cooling flow regardless of output. The third type in order of preference would be a Sulzer Type boiler since it has a separator to remove dissolved material from the water which is comparible in efficiency to a standard boiler drum and although the flow through evaporator and superheater fluctuates with load, the Sulzer Boiler can be operated as a forced recirculation boiler at low loads. The least desirable type would be a Benson or supercritical boiler which is completely dependent on input water purity for its survival. It is not claimed that Benson or supercritical boilers should not or will not be used in the future for gas-cooled reactors, but only that their use would be the least conservative choice from a tube failure standpoint at the present time. (auth)« less

Ambient air cooling arrangement having a pre-swirler for gas turbine engine blade cooling

DOEpatents

Lee, Ching-Pang; Tham, Kok-Mun; Schroeder, Eric; Meeroff, Jamie; Miller, Jr., Samuel R; Marra, John J

2015-01-06

A gas turbine engine including: an ambient-air cooling circuit (10) having a cooling channel (26) disposed in a turbine blade (22) and in fluid communication with a source (12) of ambient air: and an pre-swirler (18), the pre-swirler having: an inner shroud (38); an outer shroud (56); and a plurality of guide vanes (42), each spanning from the inner shroud to the outer shroud. Circumferentially adjacent guide vanes (46, 48) define respective nozzles (44) there between. Forces created by a rotation of the turbine blade motivate ambient air through the cooling circuit. The pre-swirler is configured to impart swirl to ambient air drawn through the nozzles and to direct the swirled ambient air toward a base of the turbine blade. The end walls (50, 54) of the pre-swirler may be contoured.

Economic analysis of condensers for water recovery in steam injected gas turbines

DOE Office of Scientific and Technical Information (OSTI.GOV)

De Paepe, M.; Huvenne, P.; Dick, E.

1998-07-01

Steam injection cycles are interesting for small power ranges because of the high efficiency and the relatively low investment costs. A big disadvantage is the consumption of water by the cycle. Water recovery is seldom realized in industrial practice. In this paper an analysis of the technical and economical possibilities of water recovery by condensation of water out of the exhaust gases is made. Three gas turbines are considered : the Kawasaki M1A-13CC (2.3 MWe), the Allison 501KH (6.8 MWe) and the General Electric LM1600 (17 MWe). For every gas turbine two types of condensers are designed. In the watermore » cooled condenser finned tubes are used to cool the exhaust gases, flowing at the outside of the tubes. The water itself flows at the inside of the tubes and is cooled by a water to air cooler. In the air cooled condenser the exhaust gases flow at the inside of the tubes and the cooling air at the outside. The investment costs of the condensers is compared to the costs of the total installation. The investment costs are relatively smaller if the produced power goes up. The water cooled condenser with water to air cooler is cheaper than the air cooled condenser. Using a condenser results in higher exploitation costs due to the fans and pumps. It is shown that the air cooled condenser has lower exploitation costs than the water cooled one. Pay back time of the total installation does not significantly vary compared to the installation without recovery. Water prices are determined for which water recovery is profitable. For the water cooled condenser the turning point lies at 2.2 Euro/m; for the air cooled condenser this is 0.6 Euro/m.« less

Performance and economic enhancement of cogeneration gas turbines through compressor inlet air cooling

NASA Astrophysics Data System (ADS)

Delucia, M.; Bronconi, R.; Carnevale, E.

1994-04-01

Gas turbine air cooling systems serve to raise performance to peak power levels during the hot months when high atmospheric temperatures cause reductions in net power output. This work describes the technical and economic advantages of providing a compressor inlet air cooling system to increase the gas turbine's power rating and reduce its heat rate. The pros and cons of state-of-the-art cooling technologies, i.e., absorption and compression refrigeration, with and without thermal energy storage, were examined in order to select the most suitable cooling solution. Heavy-duty gas turbine cogeneration systems with and without absorption units were modeled, as well as various industrial sectors, i.e., paper and pulp, pharmaceuticals, food processing, textiles, tanning, and building materials. The ambient temperature variations were modeled so the effects of climate could be accounted for in the simulation. The results validated the advantages of gas turbine cogeneration with absorption air cooling as compared to other systems without air cooling.

Process for making silicon from halosilanes and halosilicons

NASA Technical Reports Server (NTRS)

Levin, Harry (Inventor)

1988-01-01

A reactor apparatus (10) adapted for continuously producing molten, solar grade purity elemental silicon by thermal reaction of a suitable precursor gas, such as silane (SiH.sub.4), is disclosed. The reactor apparatus (10) includes an elongated reactor body (32) having graphite or carbon walls which are heated to a temperature exceeding the melting temperature of silicon. The precursor gas enters the reactor body (32) through an efficiently cooled inlet tube assembly (22) and a relatively thin carbon or graphite septum (44). The septum (44), being in contact on one side with the cooled inlet (22) and the heated interior of the reactor (32) on the other side, provides a sharp temperature gradient for the precursor gas entering the reactor (32) and renders the operation of the inlet tube assembly (22) substantially free of clogging. The precursor gas flows in the reactor (32) in a substantially smooth, substantially axial manner. Liquid silicon formed in the initial stages of the thermal reaction reacts with the graphite or carbon walls to provide a silicon carbide coating on the walls. The silicon carbide coated reactor is highly adapted for prolonged use for production of highly pure solar grade silicon. Liquid silicon (20) produced in the reactor apparatus (10) may be used directly in a Czochralski or other crystal shaping equipment.

Process for making silicon

NASA Technical Reports Server (NTRS)

Levin, Harry (Inventor)

1987-01-01

A reactor apparatus (10) adapted for continuously producing molten, solar grade purity elemental silicon by thermal reaction of a suitable precursor gas, such as silane (SiH.sub.4), is disclosed. The reactor apparatus (10) includes an elongated reactor body (32) having graphite or carbon walls which are heated to a temperature exceeding the melting temperature of silicon. The precursor gas enters the reactor body (32) through an efficiently cooled inlet tube assembly (22) and a relatively thin carbon or graphite septum (44). The septum (44), being in contact on one side with the cooled inlet (22) and the heated interior of the reactor (32) on the other side, provides a sharp temperature gradient for the precursor gas entering the reactor (32) and renders the operation of the inlet tube assembly (22) substantially free of clogging. The precursor gas flows in the reactor (32) in a substantially smooth, substantially axial manner. Liquid silicon formed in the initial stages of the thermal reaction reacts with the graphite or carbon walls to provide a silicon carbide coating on the walls. The silicon carbide coated reactor is highly adapted for prolonged use for production of highly pure solar grade silicon. Liquid silicon (20) produced in the reactor apparatus (10) may be used directly in a Czochralski or other crystal shaping equipment.

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

DOEpatents

Furgerson, W.T.

1963-12-17

A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCulloch, R.W.; Post, D.W.; Lovell, R.T.

1981-04-01

Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relatemore » this profile to that generated by the coils in completed fuel pin simulators.« less

Fukushima Accident: Sequence of Events and Lessons Learned

NASA Astrophysics Data System (ADS)

Morse, Edward C.

2011-10-01

The Fukushima Dai-Ichi nuclear power station suffered a devastating Richter 9.0 earthquake followed by a 14.0 m tsunami on 11 March 2011. The subsequent loss of power for emergency core cooling systems resulted in damage to the fuel in the cores of three reactors. The relief of pressure from the containment in these three reactors led to sufficient hydrogen gas release to cause explosions in the buildings housing the reactors. There was probably subsequent damage to a spent fuel pool of a fourth reactor caused by debris from one of these explosions. Resultant releases of fission product isotopes in air were significant and have been estimated to be in the 3 . 7 --> 6 . 3 ×1017 Bq range (~10 MCi) for 131I and 137Cs combined, or approximately one tenth that of the Chernobyl accident. A synopsis of the sequence of events leading up to this large release of radioactivity will be presented, along with likely scenarios for stabilization and site cleanup in the future. Some aspects of the isotope monitoring programs, both locally and at large, will also be discussed. An assessment of radiological health risk for the plant workers as well as the general public will also be presented. Finally, the impact of this accident on design and deployment of nuclear generating stations in the future will be discussed.

Cooling water distribution system

DOEpatents

Orr, Richard

1994-01-01

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

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

DOEpatents

Hill, Paul R.

1994-01-01

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

Intercooler cooling-air weight flow and pressure drop for minimum drag loss

NASA Technical Reports Server (NTRS)

Reuter, J George; Valerino, Michael F

1944-01-01

An analysis has been made of the drag losses in airplane flight of cross-flow plate and tubular intercoolers to determine the cooling-air weight flow and pressure drop that give a minimum drag loss for any given cooling effectiveness and, thus, a maximum power-plant net gain due to charge-air cooling. The drag losses considered in this analysis are those due to (1) the extra drag imposed on the airplane by the weight of the intercooler, its duct, and its supports and (2) the drag sustained by the cooling air in flowing through the intercooler and its duct. The investigation covers a range of conditions of altitude, airspeed, lift-drag ratio, supercharger-pressure ratio, and supercharger adiabatic efficiency. The optimum values of cooling air pressure drop and weight flow ratio are tabulated. Curves are presented to illustrate the results of the analysis.

Decay Heat Removal from a GFR Core by Natural Convection

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, Wesley C.; Hejzlar, Pavel; Driscoll, Michael J.

2004-07-01

One of the primary challenges for Gas-cooled Fast Reactors (GFR) is decay heat removal after a loss of coolant accident (LOCA). Due to the fact that thermal gas cooled reactors currently under design rely on passive mechanisms to dissipate decay heat, there is a strong motivation to accomplish GFR core cooling through natural phenomena. This work investigates the potential of post-LOCA decay heat removal from a GFR core to a heat sink using an external convection loop. A model was developed in the form of the LOCA-COLA (Loss of Coolant Accident - Convection Loop Analysis) computer code as a meansmore » for 1D steady state convective heat transfer loop analysis. The results show that decay heat removal by means of gas cooled natural circulation is feasible under elevated post-LOCA containment pressure conditions. (authors)« less

Coolant effectiveness in dental cutting with air-turbine handpieces.

PubMed

Leung, Brian T W; Dyson, John E; Darvell, Brian W

2012-03-01

To establish a strategy for evaluating coolant effectiveness and to compare typical cooling conditions used in dental cutting. A test system comprising a resistive heat source and an array of four type K thermocouples was used to compare the cooling effectiveness of air alone, water stream alone, and an air-water spray, as delivered by representative air-turbine handpieces. Mean temperature change at the four sites was recorded for a range of water flow rates in the range 10 to 90 mL min(-1), with and without air, and with and without the turbine running. The thermal resistance of the system, R, was calculated as the temperature change per watt (KW(-1)). For wet cooling (water stream and air-water spray), R was 5.1 to 11.5 KW(-1), whereas for air coolant alone the range was 18.5 to 30.7 KW(-1). R for air-water spray was lower than for water stream cooling at the same flow rate. The thermal resistivity approach is a viable means of comparative testing of cooling efficacy in simulated dental cutting. It may provide a reliable means of testing handpiece nozzle design, thus enabling the development of more efficient cooling.

ETR COMPLEX. CAMERA FACING SOUTH. FROM BOTTOM OF VIEW TO ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR COMPLEX. CAMERA FACING SOUTH. FROM BOTTOM OF VIEW TO TOP: MTR, MTR SERVICE BUILDING, ETR CRITICAL FACILITY, ETR CONTROL BUILDING (ATTACHED TO ETR), ETR BUILDING (HIGH-BAY), COMPRESSOR BUILDING (ATTACHED AT LEFT OF ETR), HEAT EXCHANGER BUILDING (JUST BEYOND COMPRESSOR BUILDING), COOLING TOWER PUMP HOUSE, COOLING TOWER. OTHER BUILDINGS ARE CONTRACTORS' CONSTRUCTION BUILDINGS. INL NEGATIVE NO. 56-4105. Unknown Photographer, ca. 1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Optimization of a heat-pipe-cooled space radiator for use with a reactor-powered Stirling engine

NASA Technical Reports Server (NTRS)

Moriarty, Michael P.; French, Edward P.

1987-01-01

The design optimization of a reactor-Stirling heat-pipe-cooled radiator is presented. The radiator is a self-deploying concept that uses individual finned heat pipe 'petals' to reject waste heat from a Stirling engine. Radiator optimization methodology is presented, and the results of a parametric analysis of the radiator design variables for a 100-kW(e) system are given. The additional steps of optiminzing the radiator resulted in a net system mass savings of 3 percent.

Fast quench reactor and method

DOEpatents

Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

1998-05-12

A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

Key Assets for a Sustainable Low Carbon Energy Future

NASA Astrophysics Data System (ADS)

Carre, Frank

2011-10-01

Since the beginning of the 21st century, concerns of energy security and climate change gave rise to energy policies focused on energy conservation and diversified low-carbon energy sources. Provided lessons of Fukushima accident are evidently accounted for, nuclear energy will probably be confirmed in most of today's nuclear countries as a low carbon energy source needed to limit imports of oil and gas and to meet fast growing energy needs. Future challenges of nuclear energy are then in three directions: i) enhancing safety performance so as to preclude any long term impact of severe accident outside the site of the plant, even in case of hypothetical external events, ii) full use of Uranium and minimization long lived radioactive waste burden for sustainability, and iii) extension to non-electricity energy products for maximizing the share of low carbon energy source in transportation fuels, industrial process heat and district heating. Advanced LWRs (Gen-III) are today's best available technologies and can somewhat advance nuclear energy in these three directions. However, breakthroughs in sustainability call for fast neutron reactors and closed fuel cycles, and non-electric applications prompt a revival of interest in high temperature reactors for exceeding cogeneration performances achievable with LWRs. Both types of Gen-IV nuclear systems by nature call for technology breakthroughs to surpass LWRs capabilities. Current resumption in France of research on sodium cooled fast neutron reactors (SFRs) definitely aims at significant progress in safety and economic competitiveness compared to earlier reactors of this type in order to progress towards a new generation of commercially viable sodium cooled fast reactor. Along with advancing a new generation of sodium cooled fast reactor, research and development on alternative fast reactor types such as gas or lead-alloy cooled systems (GFR & LFR) is strategic to overcome technical difficulties and/or political opposition specific to sodium. In conclusion, research and technology breakthroughs in nuclear power are needed for shaping a sustainable low carbon future. International cooperation is key for sharing costs of research and development of the required novel technologies and cost of first experimental reactors needed to demonstrate enabling technologies. At the same time technology breakthroughs are developed, pre-normative research is required to support codification work and harmonized regulations that will ultimately apply to safety and security features of resulting innovative reactor types and fuel cycles.

Thermoelectric-enhanced, liquid-based cooling of a multi-component electronic system

DOEpatents

Chainer, Timothy J; Graybill, David P; Iyengar, Madhusudan K; Kamath, Vinod; Kochuparambil, Bejoy J; Schmidt, Roger R; Steinke, Mark E

2015-11-10

Methods are provided for facilitating cooling of an electronic component. The methods include providing: a liquid-cooled structure, a thermal conduction path coupling the electronic component and the liquid-cooled structure, a coolant loop in fluid communication with a coolant-carrying channel of the liquid-cooled structure, and an outdoor-air-cooled heat exchange unit coupled to facilitate heat transfer from the liquid-cooled structure via, at least in part, the coolant loop. The thermoelectric array facilitates transfer of heat from the electronic component to the liquid-cooled structure, and the heat exchange unit cools coolant passing through the coolant loop by dissipating heat from the coolant to outdoor ambient air. In one implementation, temperature of coolant entering the liquid-cooled structure is greater than temperature of the outdoor ambient air to which heat is dissipated.

Thermoelectric-enhanced, liquid-based cooling of a multi-component electronic system

DOEpatents

Chainer, Timothy J; Graybill, David P; Iyengar, Madhusudan K; Kamath, Vinod; Kochuparambil, Bejoy J; Schmidt, Roger R; Steinke, Mark E

2015-05-12

Apparatus and method are provided for facilitating cooling of an electronic component. The apparatus includes a liquid-cooled structure, a thermal conduction path coupling the electronic component and the liquid-cooled structure, a coolant loop in fluid communication with a coolant-carrying channel of the liquid-cooled structure, and an outdoor-air-cooled heat exchange unit coupled to facilitate heat transfer from the liquid-cooled structure via, at least in part, the coolant loop. The thermoelectric array facilitates transfer of heat from the electronic component to the liquid-cooled structure, and the heat exchange unit cools coolant passing through the coolant loop by dissipating heat from the coolant to outdoor ambient air. In one implementation, temperature of coolant entering the liquid-cooled structure is greater than temperature of the outdoor ambient air to which heat is dissipated.

Experimental study of condensate subcooling with the use of a model of an air-cooled condenser

NASA Astrophysics Data System (ADS)

Sukhanov, V. A.; Bezukhov, A. P.; Bogov, I. A.; Dontsov, N. Y.; Volkovitsky, I. D.; Tolmachev, V. V.

2016-01-01

Water-supply deficit is now felt in many regions of the world. This hampers the construction of new steam-turbine and combined steam-and-gas thermal power plants. The use of dry cooling systems and, specifically, steam-turbine air-cooled condensers (ACCs) expands the choice of sites for the construction of such power plants. The significance of condensate subcooling Δ t as a parameter that negatively affects the engineering and economic performance of steam-turbine plants is thereby increased. The operation and design factors that influence the condensate subcooling in ACCs are revealed, and the research objective is, thus, formulated properly. The indicated research was conducted through physical modeling with the use of the Steam-Turbine Air-Cooled Condenser Unit specialized, multipurpose, laboratory bench. The design and the combined schematic and measurement diagram of this test bench are discussed. The experimental results are presented in the form of graphic dependences of the condensate subcooling value on cooling ratio m and relative weight content ɛ' of air in steam at the ACC inlet at different temperatures of cooling air t ca ' . The typical ranges of condensate subcooling variation (4 ≤ Δ t ≤ 6°C, 2 ≤ Δ t ≤ 4°C, and 0 ≤ Δ t ≤ 2°C) are identified based on the results of analysis of the attained Δ t levels in the ACC and numerous Δ t reduction estimates. The corresponding ranges of cooling ratio variation at different temperatures of cooling air at the ACC inlet are specified. The guidelines for choosing the adjusted ranges of cooling ratio variation with account of the results of experimental studies of the dependences of the absolute pressure of the steam-air mixture in the top header of the ACC and the heat flux density on the cooling ratio at different temperatures of cooling air at the ACC inlet are given.

Heat Pipe Technology: A bibliography with abstracts

NASA Technical Reports Server (NTRS)

1974-01-01

This bibliography lists 149 references with abstracts and 47 patents dealing with applications of heat pipe technology. Topics covered include: heat exchangers for heat recovery; electrical and electronic equipment cooling; temperature control of spacecraft; cryosurgery; cryogenic, cooling; nuclear reactor heat transfer; solar collectors; laser mirror cooling; laser vapor cavitites; cooling of permafrost; snow melting; thermal diodes variable conductance; artery gas venting; and venting; and gravity assisted pipes.

Core design of a direct-cycle, supercritical-water-cooled fast breeder reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jevremovic, T.; Oka, Yoshiaki; Koshizuka, Seiichi

1994-10-01

The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without muchmore » deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.« less

21 CFR 211.46 - Ventilation, air filtration, air heating and cooling.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 21 Food and Drugs 4 2014-04-01 2014-04-01 false Ventilation, air filtration, air heating and... Buildings and Facilities § 211.46 Ventilation, air filtration, air heating and cooling. (a) Adequate ventilation shall be provided. (b) Equipment for adequate control over air pressure, micro-organisms, dust...

21 CFR 211.46 - Ventilation, air filtration, air heating and cooling.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 21 Food and Drugs 4 2013-04-01 2013-04-01 false Ventilation, air filtration, air heating and... Buildings and Facilities § 211.46 Ventilation, air filtration, air heating and cooling. (a) Adequate ventilation shall be provided. (b) Equipment for adequate control over air pressure, micro-organisms, dust...

21 CFR 211.46 - Ventilation, air filtration, air heating and cooling.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 21 Food and Drugs 4 2012-04-01 2012-04-01 false Ventilation, air filtration, air heating and... Buildings and Facilities § 211.46 Ventilation, air filtration, air heating and cooling. (a) Adequate ventilation shall be provided. (b) Equipment for adequate control over air pressure, micro-organisms, dust...

21 CFR 211.46 - Ventilation, air filtration, air heating and cooling.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 21 Food and Drugs 4 2011-04-01 2011-04-01 false Ventilation, air filtration, air heating and... Buildings and Facilities § 211.46 Ventilation, air filtration, air heating and cooling. (a) Adequate ventilation shall be provided. (b) Equipment for adequate control over air pressure, micro-organisms, dust...

77 FR 42771 - License Renewal for the Dow Chemical TRIGA Research Reactor

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-20

... Chemical Company in Midland, MI and is a part of the Analytical Sciences Laboratory. The reactor is housed...-Radiological Impacts The Dow TRIGA Research Reactor core is cooled by a light water primary system consisting... provided by the volume of primary coolant allows several hours of full-power operation without any...

Experiments on rehabilitation of radioactive metallic waste (RMW) of reactor stainless steels of Siberian chemical plant

NASA Astrophysics Data System (ADS)

Kolpakov, G. N.; Zakusilov, V. V.; Demyanenko, N. V.; Mishin, A. S.

2016-06-01

Stainless steel pipes, used to cool a reactor plant, have a high cost, and after taking a reactor out of service they must be buried together with other radioactive waste. Therefore, the relevant problem is the rinse of pipes from contamination, followed by returning to operation.

Coupled field-structural analysis of HGTR fuel brick using ABAQUS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, S.; Jain, R.; Majumdar, S.

2012-07-01

High-temperature, gas-cooled reactors (HTGRs) are usually helium-gas cooled, with a graphite core that can operate at reactor outlet temperatures much higher than can conventional light water reactors. In HTGRs, graphite components moderate and reflect neutrons. During reactor operation, high temperature and high irradiation cause damage to the graphite crystal and grains and create other defects. This cumulative structural damage during the reactor lifetime leads to changes in graphite properties, which can alter the ability to support the designed loads. The aim of the present research is to develop a finite-element code using commercially available ABAQUS software for the structural integritymore » analysis of graphite core components under extreme temperature and irradiation conditions. In addition, the Reactor Geometry Generator tool-kit, developed at Argonne National Laboratory, is used to generate finite-element mesh for complex geometries such as fuel bricks with multiple pin holes and coolant flow channels. This paper presents the proposed concept and discusses results of stress analysis simulations of a fuel block with H-451 grade material properties. (authors)« less

Emissivity of Candidate Materials for VHTR Applicationbs: Role of Oxidation and Surface Modification Treatments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sridharan, Kumar; Allen, Todd; Anderson, Mark

The Generation IV (GEN IV) Nuclear Energy Systems Initiative was instituted by the Department of Energy (DOE) with the goal of researching and developing technologies and materials necessary for various types of future reactors. These GEN IV reactors will employ advanced fuel cycles, passive safety systems, and other innovative systems, leading to significant differences between these future reactors and current water-cooled reactors. The leading candidate for the Next Generation Nuclear Plant (NGNP) to be built at Idaho National Lab (INL) in the United States is the Very High Temperature Reactor (VHTR). Due to the high operating temperatures of the VHTR,more » the Reactor Pressure Vessel (RPV) will partially rely on heat transfer by radiation for cooling. Heat expulsion by radiation will become all the more important during high temperature excursions during off-normal accident scenarios. Radiant power is dictated by emissivity, a material property. The NGNP Materials Research and Development Program Plan [1] has identified emissivity and the effects of high temperature oxide formation on emissivity as an area of research towards the development of the VHTR.« less

Research Program of a Super Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

40 CFR 1065.122 - Engine cooling and lubrication.

Code of Federal Regulations, 2011 CFR

2011-07-01

....122 Section 1065.122 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR POLLUTION CONTROLS ENGINE-TESTING PROCEDURES Equipment Specifications § 1065.122 Engine cooling and lubrication. (a) Engine cooling. Cool the engine during testing so its intake-air, oil, coolant, block, and...

40 CFR 1065.122 - Engine cooling and lubrication.

Code of Federal Regulations, 2010 CFR

2010-07-01

....122 Section 1065.122 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR POLLUTION CONTROLS ENGINE-TESTING PROCEDURES Equipment Specifications § 1065.122 Engine cooling and lubrication. (a) Engine cooling. Cool the engine during testing so its intake-air, oil, coolant, block, and...

40 CFR 1065.122 - Engine cooling and lubrication.

Code of Federal Regulations, 2013 CFR

2013-07-01

....122 Section 1065.122 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR POLLUTION CONTROLS ENGINE-TESTING PROCEDURES Equipment Specifications § 1065.122 Engine cooling and lubrication. (a) Engine cooling. Cool the engine during testing so its intake-air, oil, coolant, block, and...

40 CFR 1065.122 - Engine cooling and lubrication.

Code of Federal Regulations, 2012 CFR

2012-07-01

....122 Section 1065.122 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR POLLUTION CONTROLS ENGINE-TESTING PROCEDURES Equipment Specifications § 1065.122 Engine cooling and lubrication. (a) Engine cooling. Cool the engine during testing so its intake-air, oil, coolant, block, and...

40 CFR 1065.122 - Engine cooling and lubrication.

Code of Federal Regulations, 2014 CFR

2014-07-01

....122 Section 1065.122 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR POLLUTION CONTROLS ENGINE-TESTING PROCEDURES Equipment Specifications § 1065.122 Engine cooling and lubrication. (a) Engine cooling. Cool the engine during testing so its intake-air, oil, coolant, block, and...

16 CFR 305.7 - Determinations of capacity.

Code of Federal Regulations, 2010 CFR

2010-01-01

... appendix P to 10 CFR part 430, subpart B. (f) Room air conditioners. The capacity shall be the cooling... 430, subpart B. (i) Central air conditioners, cooling. The capacity shall be the cooling capacity in... capacities between 38,000 and 64,999 Btu's per hour. (j) Central air conditioners, heating. The capacity...

Numerical study of the thermo-flow performances of novel finned tubes for air-cooled condensers in power plant

NASA Astrophysics Data System (ADS)

Guo, Yonghong; Du, Xiaoze; Yang, Lijun

2018-02-01

Air-cooled condenser is the main equipment of the direct dry cooling system in a power plant, which rejects heat of the exhaust steam with the finned tube bundles. Therefore, the thermo-flow performances of the finned tubes have an important effect on the optimal operation of the direct dry cooling system. In this paper, the flow and heat transfer characteristics of the single row finned tubes with the conventional flat fins and novel jagged fins are investigated by numerical method. The flow and temperature fields of cooling air for the finned tubes are obtained. Moreover, the variations of the flow resistance and average convection heat transfer coefficient under different frontal velocity of air and jag number are presented. Finally, the correlating equations of the friction factor and Nusselt number versus the Reynolds number are fitted. The results show that with increasing the frontal velocity of air, the heat transfer performances of the finned tubes are enhanced but the pressure drop will increase accordingly, resulting in the average convection heat transfer coefficient and friction factor increasing. Meanwhile, with increasing the number of fin jag, the heat transfer performance is intensified. The present studies provide a reference in optimal designing for the air-cooled condenser of direct air cooling system.

Passive decay heat removal system for water-cooled nuclear reactors

DOEpatents

Forsberg, Charles W.

1991-01-01

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

WATER BOILER REACTOR

DOEpatents

King, L.D.P.

1960-11-22

As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

Dry Air Cooler Modeling for Supercritical Carbon Dioxide Brayton Cycle Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moisseytsev, A.; Sienicki, J. J.; Lv, Q.

Modeling for commercially available and cost effective dry air coolers such as those manufactured by Harsco Industries has been implemented in the Argonne National Laboratory Plant Dynamics Code for system level dynamic analysis of supercritical carbon dioxide (sCO 2) Brayton cycles. The modeling can now be utilized to optimize and simulate sCO 2 Brayton cycles with dry air cooling whereby heat is rejected directly to the atmospheric heat sink without the need for cooling towers that require makeup water for evaporative losses. It has sometimes been stated that a benefit of the sCO 2 Brayton cycle is that it enablesmore » dry air cooling implying that the Rankine steam cycle does not. A preliminary and simple examination of a Rankine superheated steam cycle and an air-cooled condenser indicates that dry air cooling can be utilized with both cycles provided that the cycle conditions are selected appropriately« less

PBF Cooling Tower contextual view. Camera facing southwest. West wing ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Cooling Tower contextual view. Camera facing southwest. West wing and north facade (rear) of Reactor Building (PER-620) is at left; Cooling Tower to right. Photographer: Kirsh. Date: November 2, 1970. INEEL negative no. 70-4913 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

NREL's Energy-Saving Technology for Air Conditioning Cuts Peak Power Loads Without Using Harmful Refrigerants (Fact Sheet)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

2012-07-01

This fact sheet describes how the DEVAP air conditioner was invented, explains how the technology works, and why it won an R&D 100 Award. Desiccant-enhanced evaporative (DEVAP) air-conditioning will provide superior comfort for commercial buildings in any climate at a small fraction of the electricity costs of conventional air-conditioning equipment, releasing far less carbon dioxide and cutting costly peak electrical demand by an estimated 80%. Air conditioning currently consumes about 15% of the electricity generated in the United States and is a major contributor to peak electrical demand on hot summer days, which can lead to escalating power costs, brownouts,more » and rolling blackouts. DEVAP employs an innovative combination of air-cooling technologies to reduce energy use by up to 81%. DEVAP also shifts most of the energy needs to thermal energy sources, reducing annual electricity use by up to 90%. In doing so, DEVAP is estimated to cut peak electrical demand by nearly 80% in all climates. Widespread use of this cooling cycle would dramatically cut peak electrical loads throughout the country, saving billions of dollars in investments and operating costs for our nation's electrical utilities. Water is already used as a refrigerant in evaporative coolers, a common and widely used energy-saving technology for arid regions. The technology cools incoming hot, dry air by evaporating water into it. The energy absorbed by the water as it evaporates, known as the latent heat of vaporization, cools the air while humidifying it. However, evaporative coolers only function when the air is dry, and they deliver humid air that can lower the comfort level for building occupants. And even many dry climates like Phoenix, Arizona, have a humid season when evaporative cooling won't work well. DEVAP extends the applicability of evaporative cooling by first using a liquid desiccant-a water-absorbing material-to dry the air. The dry air is then passed to an indirect evaporative cooling stage, in which the incoming air is in thermal contact with a moistened surface that evaporates the water into a separate air stream. As the evaporation cools the moistened surface, it draws heat from the incoming air without adding humidity to it. A number of cooling cycles have been developed that employ indirect evaporative cooling, but DEVAP achieves a superior efficiency relative to its technological siblings.« less

Thermally Simulated Testing of a Direct-Drive Gas-Cooled Nuclear Reactor

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Bragg-Sitton, Shannon; VanDyke, Melissa

2003-01-01

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.

Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing

NASA Technical Reports Server (NTRS)

Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; VanDyke, M. K.

2002-01-01

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

Design of megawatt power level heat pipe reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors.more » The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.« less

Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

Online Oxide Contamination Measurement and Purification Demonstration

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Godfroy, T. J.; Webster, K. L.; Garber, A. E.; Polzin, K. A.; Childers, D. J.

2011-01-01

Liquid metal sodium-potassium (NaK) has advantageous thermodynamic properties indicating its use as a fission reactor coolant for a surface (lunar, martian) power system. A major area of concern for fission reactor cooling systems is system corrosion due to oxygen contaminants at the high operating temperatures experienced. A small-scale, approximately 4-L capacity, simulated fission reactor cooling system employing NaK as a coolant was fabricated and tested with the goal of demonstrating a noninvasive oxygen detection and purification system. In order to generate prototypical conditions in the simulated cooling system, several system components were designed, fabricated, and tested. These major components were a fully-sealed, magnetically-coupled mechanical NaK pump, a graphite element heated reservoir, a plugging indicator system, and a cold trap. All system components were successfully demonstrated at a maximum system flow rate of approximately 150 cc/s at temperatures up to 550 C. Coolant purification was accomplished using a cold trap before and after plugging operations which showed a relative reduction in oxygen content.

Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

DOE PAGES

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

2009-01-01

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

Commercial-Scale Demonstration of the Liquid Phase Methanol (LPMEOH) Process

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

he Liquid Phase Methanol (LPMEOW) Demonstration Project at Kingsport Tennessee, is a $213.7 million cooperative agreement between the U.S. Department of Energy (DOE) and Air Products Liquid Phase Conversion Company, L.P. (the Partnership) to produce methanol from coal-derived synthesis gas (syngas). Air Products and Chemicals, Inc. (Air Products) and Eastman Chemical Company (Eastman) formed the Partnership to execute the Demonstration Project. The LPMEOEP Process Demonstration Unit was built at a site located at the Eastman coal-to-chemicals complex in Kingsport. The LPMEOHW Demonstration Facility completed its first year of operation on 02 April 1998. The LPMEOW Demonstration Facility also completed themore » longest continuous operating run (65 days) on 21 April 1998. Catalyst activity, as defined by the ratio of the rate constant at any point in time to the rate constant for freshly reduced catalyst (as determined in the laboratory autoclave), was monitored throughout the reporting period. During a six-week test at a reactor temperature of 225oC and Balanced Gas flowrate of 700 KSCFH, the rate of decline in catalyst activity was steady at 0.29-0.36% per day. During a second one-month test at a reactor temperature of 220oC and a Balanced Gas flowrate of 550-600 KSCFH, the rate of decline in catalyst activity was 0.4% per day, which matched the pefiorrnance at 225"C, as well as the 4-month proof-of-concept run at the LaPorte AFDU in 1988/89. Beginning on 08 May 1998, the LPMEOW Reactor temperature was increased to 235oC, which was the operating temperature tier the December 1997 restart with the fresh charge of catalyst (50'Yo of design loading). The flowrate of the primary syngas feed stream (Balanced Gas) was also increased to 700-750 KSCFH. During two stable operating periods between 08 May and 09 June 1998, the average catalyst deactivation rate was 0.8% per day. Due to the scatter of the statistical analysis of the results, this test was extended to better quanti& the catalyst aging behavior. During the reporting perio~ two batches of fresh catalyst were activated and transferred to the reactor (on 02 April and 20 June 1998). The weight of catalyst in the LPMEOW Reactor has reached 80% of the design value. At the end of the reporting period, a step-change in the pressure-drop profile within the LPMEOW Reactor and an increase in the pressure of the steam system which provides cooling to the LPMEOW Reactor were observed. No change in the calculated activity of the catalyst was detected during either of these transients. These parameters will be monitored closely for any additional changes.« less

Effect of Silver or Copper Nanoparticles-Dispersed Silane Coatings on Biofilm Formation in Cooling Water Systems

PubMed Central

Ogawa, Akiko; Kanematsu, Hideyuki; Sano, Katsuhiko; Sakai, Yoshiyuki; Ishida, Kunimitsu; Beech, Iwona B.; Suzuki, Osamu; Tanaka, Toshihiro

2016-01-01

Biofouling often occurs in cooling water systems, resulting in the reduction of heat exchange efficiency and corrosion of the cooling pipes, which raises the running costs. Therefore, controlling biofouling is very important. To regulate biofouling, we focus on the formation of biofilm, which is the early step of biofouling. In this study, we investigated whether silver or copper nanoparticles-dispersed silane coatings inhibited biofilm formation in cooling systems. We developed a closed laboratory biofilm reactor as a model of a cooling pipe and used seawater as a model for cooling water. Silver or copper nanoparticles-dispersed silane coating (Ag coating and Cu coating) coupons were soaked in seawater, and the seawater was circulated in the laboratory biofilm reactor for several days to create biofilms. Three-dimensional images of the surface showed that sea-island-like structures were formed on silane coatings and low concentration Cu coating, whereas nothing was formed on high concentration Cu coatings and low concentration Ag coating. The sea-island-like structures were analyzed by Raman spectroscopy to estimate the components of the biofilm. We found that both the Cu coating and Ag coating were effective methods to inhibit biofilm formation in cooling pipes. PMID:28773758

Performance of Air-cooled Engine Cylinders Using Blower Cooling

NASA Technical Reports Server (NTRS)

Schey, Oscar W; Ellerbrock, Herman H , Jr

1936-01-01

An investigation was made to obtain information on the minimum quantity of air and power required to cool conventional air cooled cylinders at various operating conditions when using a blower. The results of these tests show that the minimum power required for satisfactory cooling with an overall blower efficiency of 100 percent varied from 2 to 6 percent of the engine power depending on the operating conditions. The shape of the jacket had a large effect on the cylinder temperatures. Increasing the air speed over the front of the cylinder by keeping the greater part of the circumference of the cylinder covered by the jacket reduced the temperatures over the entire cylinder.

Hydronic radiant cooling: Overview and preliminary performance assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feustel, H.E.

1993-05-01

A significant amount of electrical energy used to cool non-residential buildings is drawn by the fans used to transport the cool air through the thermal distribution system. Hydronic systems reduce the amount of air transported through the building by separating ventilation and thermal conditioning. Due to the physical properties of water, hydronic distribution systems can transport a given amount of thermal energy using less than 5% of the otherwise necessary fan energy. This savings alone significantly reduces the energy consumption and especially the peak power requirement This survey clearly shows advantages for radiant cooling in combination with hydronic thermal distributionmore » systems in comparison with the All-Air Systems commonly used in California. The report describes a literature survey on the system's development, thermal comfort issues, and cooling performance. The cooling power potential and the cooling power requirement are investigated for several California climates. Peak-power requirement is compared for hydronic radiant cooling and conventional All-Air-Systems.« less

Hydronic radiant cooling: Overview and preliminary performance assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feustel, H.E.

1993-05-01

A significant amount of electrical energy used to cool non-residential buildings is drawn by the fans used to transport the cool air through the thermal distribution system. Hydronic systems reduce the amount of air transported through the building by separating ventilation and thermal conditioning. Due to the physical properties of water, hydronic distribution systems can transport a given amount of thermal energy using less than 5% of the otherwise necessary fan energy. This savings alone significantly reduces the energy consumption and especially the peak power requirement This survey clearly shows advantages for radiant cooling in combination with hydronic thermal distributionmore » systems in comparison with the All-Air Systems commonly used in California. The report describes a literature survey on the system`s development, thermal comfort issues, and cooling performance. The cooling power potential and the cooling power requirement are investigated for several California climates. Peak-power requirement is compared for hydronic radiant cooling and conventional All-Air-Systems.« less

Air cooling of disk of a solid integrally cast turbine rotor for an automotive gas turbine

NASA Technical Reports Server (NTRS)

Gladden, H. J.

1977-01-01

A thermal analysis is made of surface cooling of a solid, integrally cast turbine rotor disk for an automotive gas turbine engine. Air purge and impingement cooling schemes are considered and compared with an uncooled reference case. Substantial reductions in blade temperature are predicted with each of the cooling schemes studied. It is shown that air cooling can result in a substantial gain in the stress-rupture life of the blade. Alternatively, increases in the turbine inlet temperature are possible.

Rapid Induction of Therapeutic Hypothermia Using Transnasal High Flow Dry Air

PubMed Central

Chava, Raghuram; Raghavan, Madhavan Srinivas; Halperin, Henry; Maqbool, Farhan; Geocadin, Romergryko; Quinones-Hinojosa, Alfredo; Kolandaivelu, Aravindan; Rosen, Benjamin A.

2017-01-01

Early induction of therapeutic hypothermia (TH) is recommended in out-of-hospital cardiac arrest (CA); however, currently no reliable methods exist to initiate cooling. We investigated the effect of high flow transnasal dry air on brain and body temperatures in adult porcine animals. Adult porcine animals (n = 23) under general anesthesia were subject to high flow of transnasal dry air. Mouth was kept open to create a unidirectional airflow, in through the nostrils and out through the mouth. Brain, internal jugular, and aortic temperatures were recorded. The effect of varying airflow rate and the air humidity (0% or 100%) on the temperature profiles were recorded. The degree of brain cooling was measured as the differential temperature from baseline. A 10-minute exposure of high flow dry air caused rapid cooling of brain and gradual cooling of the jugular and the aortic temperatures in all animals. The degree of brain cooling was flow dependent and significantly higher at higher airflow rates (0.8°C ± 0.3°C, 1.03°C ± 0.6°C, and 1.3°C ± 0.7°C for 20, 40, and 80 L, respectively, p < 0.05 for all comparisons). Air temperature had minimal effect on the brain cooling over 10 minutes with similar decrease in temperature at 4°C and 30°C. At a constant flow rate (40 LPM) and temperature, the degree of cooling over 10 minutes during dry air exposure was significantly higher compared to humid air (100% saturation) (1.22°C ± 0.35°C vs. 0.21°C ± 0.12°C, p < 0.001). High flow transnasal dry air causes flow dependent cooling of the brain and the core temperatures in intubated porcine animals. The mechanism of cooling appears to be evaporation of nasal mucus as cooling is mitigated by humidifying the air. This mechanism may be exploited to initiate TH in CA. PMID:27635468

Rapid Induction of Therapeutic Hypothermia Using Transnasal High Flow Dry Air.

PubMed

Chava, Raghuram; Zviman, Menekhem; Raghavan, Madhavan Srinivas; Halperin, Henry; Maqbool, Farhan; Geocadin, Romergryko; Quinones-Hinojosa, Alfredo; Kolandaivelu, Aravindan; Rosen, Benjamin A; Tandri, Harikrishna

2017-03-01

Early induction of therapeutic hypothermia (TH) is recommended in out-of-hospital cardiac arrest (CA); however, currently no reliable methods exist to initiate cooling. We investigated the effect of high flow transnasal dry air on brain and body temperatures in adult porcine animals. Adult porcine animals (n = 23) under general anesthesia were subject to high flow of transnasal dry air. Mouth was kept open to create a unidirectional airflow, in through the nostrils and out through the mouth. Brain, internal jugular, and aortic temperatures were recorded. The effect of varying airflow rate and the air humidity (0% or 100%) on the temperature profiles were recorded. The degree of brain cooling was measured as the differential temperature from baseline. A 10-minute exposure of high flow dry air caused rapid cooling of brain and gradual cooling of the jugular and the aortic temperatures in all animals. The degree of brain cooling was flow dependent and significantly higher at higher airflow rates (0.8°C ± 0.3°C, 1.03°C ± 0.6°C, and 1.3°C ± 0.7°C for 20, 40, and 80 L, respectively, p < 0.05 for all comparisons). Air temperature had minimal effect on the brain cooling over 10 minutes with similar decrease in temperature at 4°C and 30°C. At a constant flow rate (40 LPM) and temperature, the degree of cooling over 10 minutes during dry air exposure was significantly higher compared to humid air (100% saturation) (1.22°C ± 0.35°C vs. 0.21°C ± 0.12°C, p < 0.001). High flow transnasal dry air causes flow dependent cooling of the brain and the core temperatures in intubated porcine animals. The mechanism of cooling appears to be evaporation of nasal mucus as cooling is mitigated by humidifying the air. This mechanism may be exploited to initiate TH in CA.

Current Ground Test Options for Nuclear Thermal Propulsion (NTP)

NASA Technical Reports Server (NTRS)

Gerrish, Harold P., Jr.

2014-01-01

About 20 different NTP engines/ reactors were tested from 1959 to 1972 as part of the Rover and Nuclear Engine for Rocket Vehicle Application (NERVA) program. Most were tested in open air at test cell A or test cell C, at the Nevada Test Site (NTS). Even after serious engine breakdowns of the reactor (e.g., Phoebus 1A), the test cells were cleaned up for other engine tests. The engine test stand (ETS) was made for high altitude (approximately 1 psia) testing of an NTP engine with a flight configuration, but still had the exhaust released to open air. The Rover/NERVA program became aware of new environmental regulations which would prohibit the release of any significant quantity of radioactive particulates and noble gases into the open air. The nuclear furnace (NF-1) was the last reactor tested before the program was cancelled in 1973, but successfully demonstrated a scrubber concept on how to filter the NTP exhaust. The NF-1 was demonstrated in the summer of 1972. The NF-1 used a 44MW reactor and operated each run for approximately 90 minutes. The system cooled the hot hydrogen exhaust from the engine with a water spray before entering a particle filter. The exhaust then passed through a series of heat exchangers and water separators to help remove water from the exhaust and further reduce the exhaust temperatures. The exhaust was next prepared for the charcoal trap by passing through a dryer and effluent cooler to bring exhaust temperatures close to liquid nitrogen. At those low temperatures, most of the noble gases (e.g., Xe and Kr made from fission products) get captured in the charcoal trap. The filtered hydrogen is finally passed through a flare stack and released to the air. The concept was overall successful but did show a La plating on some surfaces and had multiple recommendations for improvement. The most recent detailed study on the NTP scrubber concept was performed by the ARES Corporation in 2006. The concept is based on a 50,000 lbf thrust engine (approximately 1 GW) with a maximum burn time of 1 hour. The concept utilized lessons learned from NF-1. The strategy breaks down the exhaust into parallel paths to allow flexibility with engine size and mass flow of exhaust. Similar to NF-1, the exhaust is slowed down, cooled, filtered of particulates, filtered of noble gases, and then the clean hydrogen is flared to open air. Another concept proposed by Steve Howe (currently Director of the Center for Space Nuclear Research) to simplify the NTP exhaust filtering is to run the hydrogen exhaust into boreholes underground to filter the exhaust. The two borehole site locations proposed are at the NTS and at the Idaho National Laboratory (INL). At NTS, the boreholes are 8' diameter and 1200' deep. The permeability of hydrogen through the soil and its buoyancy will allow it to rise up through the soil and allow the filtering of noble gases and radioactive particulates. The exhaust needs to be cooled to 600C before entering the borehole to avoid soil glazing. Preliminary analysis shows a small buildup of back pressure with time which depends on permeability. Noble gases entering the borehole walls deep can take a long time before reaching the surface. Other factors affecting permeability include borehole pressure, water saturation, and turbulence. Also, a possible need to pump out contaminated water collected at the bottom of the borehole. At INL, the borehole concept is slightly different. The underground borehole has openings to the soil at special depths which have impermeable interbeds above the water table and below the surface to allow the exhaust to travel horizontal between the impermeable layers. Preliminary results indicate better permeability than at NTS. The last option is total containment of the exhaust during the test run. The concept involves slowing down the flow to subsonic in a water cooled diffuser. The hydrogen is burned off in an oxygen rich afterburner with the only products being steam, oxygen, and some noble gases. A heat exchanger and water spray pulls heat from the steam and lowers the temperature for condensation. The optimum ratio between the two is being investigated, with a goal to minimize the total volume of the water hold tanks. A water tank farm collects the contaminated water. The amount of water produced from burning the hydrogen is approximately 100,000 gallons (not including cooling water) for a 25k lbf engine operating for 50 minutes. Residual gases (e.g., oxygen and some noble gases) can be captured at cryogenic levels with a liquid nitrogen cooled dewar. After a few weeks post-test, the radiation levels can drop to more favorable levels before slowly draining each capture tank and using existing filters. With today's environmental regulations, the NTP exhaust is filtered to meet 10 mrem/year exposure to the general public (at a DOE site) or 100 mrem/year (via NRC when tested elsewhere), when natural background radiation exposure to the general public is 300- 600 mrem per year. The current society feels more comfortable with filtering even lower to as low as reasonably achievable (ALARA).

Small Reactor for Deep Space Exploration

ScienceCinema

none,

2018-06-06

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.

10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bohachek, Randolph Charles

2015-09-01

The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactorsmore » is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.« less

Technical assumption for Mo-99 production in the MARIA reactor. Feasibility study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jaroszewicz, J.; Pytel, K.; Dabkowski, L.

2008-07-15

The main objective of U-235 irradiation is to obtain the Tc-99m isotope which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short life time, is a reaction of radioactive decay of Mo-99 into Tc- 99m. One of the possible sources of molybdenum can be achieved in course of the U-235 fission reaction. The paper presents activities and the calculations results obtained upon the feasibility study on irradiation of U-235 targets for production of molybdenum in the MARIA reactor. The activities including technical assumption were focused on performing calculation for modelling ofmore » the target and irradiation device as well as adequate equipment and tools for processing in reactor. It has been assumed that the basic component of fuel charge is an aluminium cladded plate with dimensions of 40x230x1.45 containing 4.7 g U-235. The presumed mode of the heat removal generated in the fuel charge of the reactor primary cooling circuit influences the construction of installation to be used for irradiation and the technological instrumentation. The outer channel construction for irradiation has to be identical as the standard fuel channel construction of the MARIA reactor. It enables to use the existing slab and reactor mounting sockets for the fastening of the molybdenum channel as well as the cooling water delivery system. The measurement of water temperature cooling a fuel charge and control of water flow rate in the channel can also be carried out be means of the standard instrumentation of the reactor. (author)« less

SAM Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui

The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM development aims for advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability for reactor transient analyses. To facilitate the code development, SAM utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage modern advanced software environments and numerical methods. SAM focuses on modeling advanced reactormore » concepts such as SFRs (sodium fast reactors), LFRs (lead-cooled fast reactors), and FHRs (fluoride-salt-cooled high temperature reactors) or MSRs (molten salt reactors). These advanced concepts are distinguished from light-water reactors in their use of single-phase, low-pressure, high-temperature, and low Prandtl number (sodium and lead) coolants. As a new code development, the initial effort has been focused on modeling and simulation capabilities of heat transfer and single-phase fluid dynamics responses in Sodium-cooled Fast Reactor (SFR) systems. The system-level simulation capabilities of fluid flow and heat transfer in general engineering systems and typical SFRs have been verified and validated. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As the code is still under ongoing development, this SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.« less

The application of condensate water as an additional cooling media intermittently in condenser of a split air conditioning

NASA Astrophysics Data System (ADS)

Ardita, I. N.; Subagia, I. W. A.

2018-01-01

The condensate water produced by indoor a split air conditioning is usually not utilized and thrown away into the environment. The result of measurement shows that the temperature of condensate water produced by split air conditioning is quite low, that is 19-22 °C at the rate of 16-20 mL / min and it has PH balance. Under such conditions, Air Condensate produced by split air conditioning should still be recovered as an additional cooling medium on the condenser. This research will re-investigate the use of condensate water as an intermittent additional cooling of the condenser to increase the cooling capacity and performance of the air conditioning system. This research is done by experimental method whose implementation includes; designing and manufacturing of experimental equipment, mounting measuring tools, experimental data retrieval, data processing and yield analysis. The experimental results show that the use of condensate water as an intermittent additional cooling medium on split air conditioning condenser can increase the refrigeration effect about 2%, cooling capacity about 4% and 7% of COP system. Experimental results also show a decrease in power consumption in the system compressor about 3%

Personal cooling apparatus and method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Siman-Tov, Moshe; Crabtree, Jerry Allen

2001-01-01

A portable lightweight cooling apparatus for cooling a human body is disclosed, having a channeled sheet which absorbs sweat and/or evaporative liquid, a layer of highly conductive fibers adjacent the channeled sheet; and, an air-moving device for moving air through the channeled sheet, wherein the layer of fibers redistributes heat uniformly across the object being cooled, while the air moving within the channeled sheet evaporates sweat and/or other evaporative liquid, absorbs evaporated moisture and the uniformly distributed heat generated by the human body, and discharges them into the environment. Also disclosed is a method for removing heat generated by themore » human body, comprising the steps of providing a garment to be placed in thermal communication with the body; placing a layer of highly conductive fibers within the garment adjacent the body for uniformly distributing the heat generated by the body; attaching an air-moving device in communication with the garment for forcing air into the garment; removably positioning an exchangeable heat sink in communication with the air-moving device for cooling the air prior to the air entering the garment; and, equipping the garment with a channeled sheet in communication with the air-moving device so that air can be directed into the channeled sheet and adjacent the layer of fibers to expell heat and moisture from the body by the air being directed out of the channeled sheet and into the environment. The cooling system may be configured to operate in both sealed and unsealed garments.« less

Personal cooling apparatus and method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Siman-Tov, Moshe; Crabtree, Jerry Allen

A portable lightweight cooling apparatus for cooling a human body is disclosed, having a channeled sheet which absorbs sweat and/or evaporative liquid, a layer of highly conductive fibers adjacent the channeled sheet; and, an air-moving device for moving air through the channeled sheet, wherein the layer of fibers redistributes heat uniformly across the object being cooled, while the air moving within the channeled sheet evaporates sweat and/or other evaporative liquid, absorbs evaporated moisture and the uniformly distributed heat generated by the human body, and discharges them into the environment. Also disclosed is a method for removing heat generated by themore » human body, comprising the steps of providing a garment to be placed in thermal communication with the body; placing a layer of highly conductive fibers within the garment adjacent the body for uniformly distributing the heat generated by the body; attaching an air-moving device in communication with the garment for forcing air into the garment; removably positioning an exchangeable heat sink in communication with the air-moving device for cooling the air prior to the air entering the garment; and, equipping the garment with a channeled sheet in communication with the air-moving device so that air can be directed into the channeled sheet and adjacent the layer of fibers to expell heat and moisture from the body by the air being directed out of the channeled sheet and into the environment. The cooling system may be configured to operate in both sealed and unsealed garments.« less

Personal cooling apparatus and method

DOEpatents

Siman-Tov, Moshe; Crabtree, Jerry Allen

2001-01-01

A portable lightweight cooling apparatus for cooling a human body is disclosed, having a channeled sheet which absorbs sweat and/or evaporative liquid, a layer of highly conductive fibers adjacent the channeled sheet; and, an air-moving device for moving air through the channeled sheet, wherein the layer of fibers redistributes heat uniformly across the object being cooled, while the air moving within the channeled sheet evaporates sweat and/or other evaporative liquid, absorbs evaporated moisture and the uniformly distributed heat generated by the human body, and discharges them into the environment. Also disclosed is a method for removing heat generated by the human body, comprising the steps of providing a garment to be placed in thermal communication with the body; placing a layer of highly conductive fibers within the garment adjacent the body for uniformly distributing the heat generated by the body; attaching an air-moving device in communication with the garment for forcing air into the garment; removably positioning an exchangeable heat sink in communication with the air-moving device for cooling the air prior to the air entering the garment; and, equipping the garment with a channeled sheet in communication with the air-moving device so that air can be directed into the channeled sheet and adjacent the layer of fibers to expell heat and moisture from the body by the air being directed out of the channeled sheet and into the environment. The cooling system may be configured to operate in both sealed and unsealed garments.

An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

NASA Technical Reports Server (NTRS)

Clark, John S.; Walton, James T.; Mcguire, Melissa L.

1992-01-01

Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines.

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

DOEpatents

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su'ud, Zaki, E-mail: szaki@fi.itba.c.id; Sekimoto, H., E-mail: hsekimot@gmail.com

2014-09-30

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

Nuclear engine flow reactivity shim control

DOEpatents

Walsh, J.M.

1973-12-11

A nuclear engine control system is provided which automatically compensates for reactor reactivity uncertainties at the start of life and reactivity losses due to core corrosion during the reactor life in gas-cooled reactors. The coolant gas flow is varied automatically by means of specially provided control apparatus so that the reactor control drums maintain a predetermined steady state position throughout the reactor life. This permits the reactor to be designed for a constant drum position and results in a desirable, relatively flat temperature profile across the core. (Official Gazette)

Control rod drive for reactor shutdown

DOEpatents

McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

1976-01-20

A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

Experimental and Computational Investigations of Plenum-to-Plenum Heat Transfer and Gas Dynamics under Natural Circulation in a Prismatic Very High Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

AL-Dahhan, Muthanna; Rizwan-Uddin, Rizwan; Usman, S.

All the goals and the objectives set for the project were successfully executed and achieved and all the milestones have been successfully completed. The results that have been obtained for the first time advance the scientific and engineering knowledge and understanding of the plenum-to plenum natural convection of prismatic block nuclear reactors that is encountered during accident or abnormal operation. These have been accomplished by developing and implementing for the first time unique and flexible scaled-down separate and integrated effects experimental plenumto- plenum facility (P2PF) with dual channels at this time that has been equipped with sophisticated measurement techniques integratedmore » in a novel way on the heated and cooled channels. The unique facility is an asset now that can be extended to research multiple channels and to study the effects of hot plumes in the plena for future projects if funding will be available. It can also be modified to research natural convection of pebble bed reactors. Hence, it complement the HTTF at Oregon State University. However, in this study, heat transfer coefficients from the inner wall surface to the flowing gas (both helium and air were used) and the radial temperature and gas velocity profiles have been measured and investigated along the height of the heated and cooled channels using in house developed wall flush mounted heat transfer probes, thermocouple with in house developed adjuster for radial movement with 1 mm increment inside the channel and hot wire anemometry with also in house developed adjuster for 1 mm radial movement inside the channel, respectively. Also advanced tracer technique has been developed to quantify also for the first time the dispersion of the gas dynamics of the hot and cold channels. The research has provided new knowledge and new benchmarking data that can be used to validate computational fluid dynamics (CFD) codes with conjugate heat transfer. The work and its results that have been performed within the budget have demonstrated their superior technical effectiveness and high economic feasibility to perform needed studies for safety analysis and assessment at least cost for these types of gas cooled very high temperature 4th generation nuclear reactors. Accordingly, the results obtained in this project and the unique facility and techniques that have been developed will benefit greatly the public by advancing the technology of the prismatic block nuclear reactors toward commercialization and to ensure they will be designed and operated safely by utilizing the obtained knowledge and having well validated CFD simulations integrated with heat transfer computations« less

NEUTRONIC REACTOR CHARGING AND DISCHARGING

DOEpatents

Zinn, W.H.

1959-07-14

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

Development of ZL400 Mine Cooling Unit Using Semi-Hermetic Screw Compressor and Its Application on Local Air Conditioning in Underground Long-Wall Face

NASA Astrophysics Data System (ADS)

Chu, Zhaoxiang; Ji, Jianhu; Zhang, Xijun; Yan, Hongyuan; Dong, Haomin; Liu, Junjie

2016-12-01

Aiming at heat injuries occurring in the process of deep coal mining in China, a ZL400 mine-cooling unit employing semi-hermetic screw compressor with a cooling capacity of 400 kW is developed. This paper introduced its operating principle, structural characteristics and technical indexes. By using the self-built testing platform, some parameters for indication of its operation conditions were tested on the ground. The results show that the aforementioned cooling unit is stable in operation: cooling capacity of the unit was 420 kW underground-test conditions, while its COP (coefficient of performance) reached 3.4. To address the issue of heat injuries existing in No. 16305 U-shaped long-wall ventilation face of Jining No. 3 coal mine, a local air conditioning system was developed with ZL400 cooling unit as the system's core. The paper presented an analysis of characteristics of the air current flowing in the air-mixing and cooling mode of ZL400 cooling unit used in air intake way. Through i-d patterns we described the process of the airflow treatment, such as cooling, mixing and heating, etc. The cooling system decreased dry bulb temperature on working face by 3°C on average and 3.8°C at most, while lowered the web bulb temperature by 3.6°C on average and 4.8°C at most. At the same time, it reduced relative humidity by 5% on average and 8.6% at most. The field application of the ZL400 cooling unit had gain certain effects in air conditioning and provided support for the solution of mine heat injuries in China in terms of technology and equipment.

Nuclear energy center site survey reactor plant considerations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harty, H.

The Energy Reorganization Act of 1974 required the Nuclear Regulatory Commission (NRC) to make a nuclear energy center site survey (NECSS). Background information for the NECSS report was developed in a series of tasks which include: socioeconomic inpacts; environmental impact (reactor facilities); emergency response capability (reactor facilities); aging of nuclear energy centers; and dry cooled nuclear energy centers.

Nuclear reactor support and seismic restraint with in-vessel core retention cooling features

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwards, Tyler A.; Edwards, Michael J.

A nuclear reactor including a lateral seismic restraint with a vertically oriented pin attached to the lower vessel head and a mating pin socket attached to the floor. Thermally insulating materials are disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head.

REACTOR UNLOADING

DOEpatents

Leverett, M.C.

1958-02-18

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

HEAVY WATER MODERATED NEUTRONIC REACTOR

DOEpatents

Szilard, L.

1958-04-29

A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

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

DOEpatents

Hill, P.R.

1994-12-27

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

Cooling system for a bearing of a turbine rotor

DOEpatents

Schmidt, Mark Christopher

2002-01-01

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rudd, A.

This document covers a description of the need and applied solutions for supplemental dehumidification in warm-humid climates, especially for energy efficient homes where the sensible cooling load has been dramatically reduced. In older homes in warm-humid climates, cooling loads are typically high and cooling equipment runs a lot to cool the air. The cooling process also removes indoor moisture, reducing indoor relative humidity. However, at current residential code levels, and especially for above-code programs, sensible cooling loads have been so dramatically reduced that the cooling system does not run a lot to cool the air, resulting in much less moisturemore » being removed. In these new homes, cooling equipment is off for much longer periods of time especially during spring/fall seasons, summer shoulder months, rainy periods, some summer nights, and some winter days. In warm-humid climates, those long off periods allow indoor humidity to become elevated due to internally generated moisture and ventilation air change. Elevated indoor relative humidity impacts comfort, indoor air quality, and building material durability. Industry is responding with supplemental dehumidification options, but that effort is really in its infancy regarding year-round humidity control in low-energy homes. Available supplemental humidity control options are discussed. Some options are less expensive but may not control indoor humidity as well as more expensive and comprehensive options. The best performing option is one that avoids overcooling and avoids adding unnecessary heat to the space by using waste heat from the cooling system to reheat the cooled and dehumidified air to room-neutral temperature.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rudd, Armin

This document covers a description of the need and applied solutions for supplemental dehumidification in warm-humid climates, especially for energy efficient homes where the sensible cooling load has been dramatically reduced. Cooling loads are typically high and cooling equipment runs a lot to cool the air in older homes in warm-humid climates. The cooling process also removes indoor moisture, reducing indoor relative humidity. However, at current residential code levels, and especially for above-code programs, sensible cooling loads have been so dramatically reduced that the cooling system does not run a lot to cool the air, resulting in much less moisturemore » being removed. In these new homes, cooling equipment is off for much longer periods of time especially during spring/fall seasons, summer shoulder months, rainy periods, some summer nights, and winter days. In warm-humid climates, those long-off periods allow indoor humidity to become elevated due to internally generated moisture and ventilation air change. Elevated indoor relative humidity impacts comfort, indoor air quality, and building material durability. Industry is responding with supplemental dehumidification options, but that effort is really in its infancy regarding year-round humidity control in low-energy homes. Available supplemental humidity control options are discussed. Some options are less expensive but may not control indoor humidity as well as more expensive and comprehensive options. The best performing option is one that avoids overcooling and adding unnecessary heat to the space by using waste heat from the cooling system to reheat the cooled and dehumidified air to room-neutral temperature.« less

Performance of Metal Cutting on Endmills Manufactured by Cooling-Air and Minimum Quantity Lubrication Grinding

NASA Astrophysics Data System (ADS)

Inoue, Shigeru; Aoyama, Tojiro

Grinding fluids have been commonly used during the grinding of tools for their cooling and lubricating effect since the hard, robust materials used for cutting tools are difficult to grind. Grinding fluids help prevent a drop in hardness due to burning of the cutting edge and keep chipping to an absolute minimum. However, there is a heightened awareness of the need to improve the work environment and protect the global environment. Thus, the present study is aimed at applying dry grinding, cooling-air grinding, cooling-air grinding with minimum quantity lubrication (MQL), and oil-based fluid grinding to manufacturing actual endmills (HSS-Co). Cutting tests were performed by a vertical machining center. The results indicated that the lowest surface inclination values and longest tool life were obtained by cooling-air grinding with MQL. Thus, cooling-air grinding with MQL has been demonstrated to be at least as effective as oil-based fluid grinding.

Cooled snubber structure for turbine blades

DOEpatents

Mayer, Clinton A.; Campbell, Christian X.; Whalley, Andrew; Marra, John J.

2014-04-01

A turbine blade assembly in a turbine engine. The turbine blade assembly includes a turbine blade and a first snubber structure. The turbine blade includes an internal cooling passage containing cooling air. The first snubber structure extends outwardly from a sidewall of the turbine blade and includes a hollow interior portion that receives cooling air from the internal cooling passage of the turbine blade.

Passive containment cooling water distribution device

DOEpatents

Conway, Lawrence E.; Fanto, Susan V.

1994-01-01

A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using a series of radial guide elements and cascading weir boxes to collect and then distribute the cooling water into a series of distribution areas through a plurality of cascading weirs. The cooling water is then uniformly distributed over the curved surface by a plurality of weir notches in the face plate of the weir box.

Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

NASA Astrophysics Data System (ADS)

Permana, Sidik; Saputra, Geby; Suzuki, Mitsutoshi; Saito, Masaki

2017-01-01

Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

Non-Intrusive Velocity Measurements with MTV During DCC Event in the HTTF

NASA Technical Reports Server (NTRS)

Andre, M. A.; Bardet, P. M.; Cadell, S. R.; Woods, B.; Burns, R. A.; Danehy, P. M.

2017-01-01

Velocity profiles are measured using molecular tagging velocimetry (MTV) in the high temperature test facility (HTTF) at Oregon State University during a depressurized conduction cooldown (DCC) event. The HTTF is a quarter scale electrically heated nuclear reactor simulator designed to replicate various accident scenarios. During a DCC, a double ended guillotine break results in the reactor pressure vessel (RPV) depressurizing into the reactor cavity and ultimately leading to air ingress in the reactor core (lock-exchange and gas diffusion). It is critical to understand the resulting buoyancy-driven flow to characterize the reactor self-cooling capacity through natural circulation. During tests conducted at ambient pressure and temperature, the RPV containing helium is opened (via the hot and cold legs) to a large vessel filled with nitrogen to simulate the atmosphere. The velocity profile on the hot leg pipe centerline is recorded at 10 Hz with MTV based on NO tracers. The precision of the velocimetry was measured to be 0.02 m/s in quiescent flow prior to the tests. A helium flow from the RPV is initially observed in the top quarter of the pipe. During the first 20 seconds of the event, helium flows out of the RPV with a maximum velocity below 2 m/s. The velocity profile transitions from parabolic to linear in character and decays slowly over the rest of the recording; peak velocities of 0.2 m/s are observed after 30 min. A counter-flow of nitrogen is also observed intermittently, which occurs at lower velocities (>0.1 m/s).

Neutronic Reactor III

NASA Astrophysics Data System (ADS)

Fermi, Enrico; Zinn, Walter H.; Anderson, Herbert L.

An improvement of the reactors described in the previous Patents, aimed at increasing the reproduction factor, is reported here, such improvement being obtained by diminishing the neutron loss due to impurities within the reactor. This is achieved by encasing the reactor in a rubberized balloon cloth housing (or something like this) in order to eliminate the atmospheric air therefrom, thus eliminating both the effect of the danger coefficient of nitrogen (70% of the atmospheric air) and that of the argon present in the air, which can become radioactive. Since the removal of the air from the reactor may result in structural problems, caused by the forces brought into play by that evacuation, the reactor is then filled with a non-reactive (from a chemical and nuclear standpoint) gas such as helium or carbon dioxide. It is interesting to point out that the authors consider also the possibility to control (a little) the reproduction ratio of the reactor by varying the air content of it. Just a rapid mention of the main idea of the present Patent (i.e. the encasing of the pile in a balloon cloth) appeared in [Fermi (1942f)], but no detailed description of the system considered here is reported in any other published paper.

Controlled CO preferential oxidation

DOEpatents

Meltser, Mark A.; Hoch, Martin M.

1997-01-01

Method for controlling the supply of air to a PROX reactor for the preferential oxidation in the presence of hydrogen wherein the concentration of the hydrogen entering and exiting the PROX reactor is monitored, the difference therebetween correlated to the amount of air needed to minimize such difference, and based thereon the air supply to the PROX reactor adjusted to provide such amount and minimize such difference.

Neutronic reactor thermal shield

DOEpatents

Wende, Charles W. J.

1976-06-15

1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

Vented target elements for use in an isotope-production reactor. [LMFBR

DOEpatents

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

1982-08-19

A method is described for producing tritium gas in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins equipped with vents, and tritium gas is recovered from the coolant.

Assemblies with both target and fuel pins in an isotope-production reactor

DOEpatents

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

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

24 CFR 3280.714 - Appliances, cooling.

Code of Federal Regulations, 2010 CFR

2010-04-01

... Systems § 3280.714 Appliances, cooling. (a) Every air conditioning unit or a combination air conditioning...) Mechanical air conditioners shall be rated in accordance with the ARI Standard 210/240-89 Unitary Air Conditioning and Air Source Unitary Heat Pump Equipment and certified by ARI or other nationally recognized...

24 CFR 3280.714 - Appliances, cooling.

Code of Federal Regulations, 2011 CFR

2011-04-01

... Systems § 3280.714 Appliances, cooling. (a) Every air conditioning unit or a combination air conditioning...) Mechanical air conditioners shall be rated in accordance with the ARI Standard 210/240-89 Unitary Air Conditioning and Air Source Unitary Heat Pump Equipment and certified by ARI or other nationally recognized...

16 CFR Appendix H to Part 305 - Cooling Performance and Cost for Central Air Conditioners

Code of Federal Regulations, 2010 CFR

2010-01-01

... 16 Commercial Practices 1 2010-01-01 2010-01-01 false Cooling Performance and Cost for Central Air Conditioners H Appendix H to Part 305 Commercial Practices FEDERAL TRADE COMMISSION REGULATIONS UNDER SPECIFIC... RULEâ) Pt. 305, App. H Appendix H to Part 305—Cooling Performance and Cost for Central Air Conditioners...

Removal of hydrogen sulfide as ammonium sulfate from hydropyrolysis product vapors

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

A system and method for processing biomass into hydrocarbon fuels that includes processing a biomass in a hydropyrolysis reactor resulting in hydrocarbon fuels and a process vapor stream and cooling the process vapor stream to a condensation temperature resulting in an aqueous stream. The aqueous stream is sent to a catalytic reactor where it is oxidized to obtain a product stream containing ammonia and ammonium sulfate. A resulting cooled product vapor stream includes non-condensable process vapors comprising H.sub.2, CH.sub.4, CO, CO.sub.2, ammonia and hydrogen sulfide.

METHOD AND APPARATUS FOR PRODUCING POWER

DOEpatents

Wollan, E.O.

1961-06-27

A neutronic reactor comprising two discrete zones; namely, an inner zone containing fissionable material and an outer zone containing fertile material is described. The inner zone is operated at a low temperature and is cooled by pressurized water. The outer zone is operated at a substantially higher temperature and is cooled by steam flashed from the inner zone. The reactor is particularly advantageous in that it produces high temperature steam; yet the materials of construction in the core (inner zone) are not restricted to materials capable of withstanding high temperature operation.

Removal of hydrogen sulfide as ammonium sulfate from hydropyrolysis product vapors

DOEpatents

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

2014-10-14

A system and method for processing biomass into hydrocarbon fuels that includes processing a biomass in a hydropyrolysis reactor resulting in hydrocarbon fuels and a process vapor stream and cooling the process vapor stream to a condensation temperature resulting in an aqueous stream. The aqueous stream is sent to a catalytic reactor where it is oxidized to obtain a product stream containing ammonia and ammonium sulfate. A resulting cooled product vapor stream includes non-condensable process vapors comprising H.sub.2, CH.sub.4, CO, CO.sub.2, ammonia and hydrogen sulfide.

Gas turbine combustor transition

DOEpatents

Coslow, Billy Joe; Whidden, Graydon Lane

1999-01-01

A method of converting a steam cooled transition to an air cooled transition in a gas turbine having a compressor in fluid communication with a combustor, a turbine section in fluid communication with the combustor, the transition disposed in a combustor shell and having a cooling circuit connecting a steam outlet and a steam inlet and wherein hot gas flows from the combustor through the transition and to the turbine section, includes forming an air outlet in the transition in fluid communication with the cooling circuit and providing for an air inlet in the transition in fluid communication with the cooling circuit.

Gas turbine combustor transition

DOEpatents

Coslow, B.J.; Whidden, G.L.

1999-05-25

A method is described for converting a steam cooled transition to an air cooled transition in a gas turbine having a compressor in fluid communication with a combustor, a turbine section in fluid communication with the combustor, the transition disposed in a combustor shell and having a cooling circuit connecting a steam outlet and a steam inlet and wherein hot gas flows from the combustor through the transition and to the turbine section, includes forming an air outlet in the transition in fluid communication with the cooling circuit and providing for an air inlet in the transition in fluid communication with the cooling circuit. 7 figs.

Reverse Brayton Cycle with Bladeless Turbo Compressor for Automotive Environmental Cooling

NASA Technical Reports Server (NTRS)

Ganapathi, Gani B. (Inventor); Cepeda-Rizo, Juan (Inventor)

2016-01-01

An automotive cabin cooling system uses a bladeless turbocompressor driven by automobile engine exhaust to compress incoming ambient air. The compressed air is directed to an intercooler where it is cooled and then to another bladeless turbine used as an expander where the air cools as it expands and is directed to the cabin interior. Excess energy may be captured by an alternator couple to the expander turbine. The system employs no chemical refrigerant and may be further modified to include another intercooler on the output of the expander turbine to isolate the cooled cabin environment.

Development of Inspection and Repair Technology for Heat Exchanger Tubes in Fast Breeder Reactors

DTIC Science & Technology

2009-06-01

Technology for Heat Exchanger Tubes in Fast Breeder Reactors Akihiko NISHIMURA *1 , Takahisa SHOBU, Kiyoshi OKA, Toshihiko YAMAGUCHI, Yukihiro SHIMADA...fast breeder reactors (FBRs). It comprises a laser processing head combined with an eddy current testing unit. Ultrashort laser pulse ablation is used...be applied in the main- tenance of large structures such as nuclear reactors and chemical factories [1]. Internal access to a blanket cooling pipe

Study of Convection Heat Transfer in a Very High Temperature Reactor Flow Channel: Numerical and Experimental Results

DOE PAGES

Valentin, Francisco I.; Artoun, Narbeh; Anderson, Ryan; ...

2016-12-01

Very High Temperature Reactors (VHTRs) are one of the Generation IV gas-cooled reactor models proposed for implementation in next generation nuclear power plants. A high temperature/pressure test facility for forced and natural circulation experiments has been constructed. This test facility consists of a single flow channel in a 2.7 m (9’) long graphite column equipped with four 2.3kW heaters. Extensive 3D numerical modeling provides a detailed analysis of the thermal-hydraulic behavior under steady-state, transient, and accident scenarios. In addition, forced/mixed convection experiments with air, nitrogen and helium were conducted for inlet Reynolds numbers from 500 to 70,000. Our numerical resultsmore » were validated with forced convection data displaying maximum percentage errors under 15%, using commercial finite element package, COMSOL Multiphysics. Based on this agreement, important information can be extracted from the model, with regards to the modified radial velocity and property gas profiles. Our work also examines flow laminarization for a full range of Reynolds numbers including laminar, transition and turbulent flow under forced convection and its impact on heat transfer under various scenarios to examine the thermal-hydraulic phenomena that could occur during both normal operation and accident conditions.« less

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

Flame tolerant secondary fuel nozzle

DOEpatents

Khan, Abdul Rafey; Ziminsky, Willy Steve; Wu, Chunyang; Zuo, Baifang; Stevenson, Christian Xavier

2015-02-24

A combustor for a gas turbine engine includes a plurality of primary nozzles configured to diffuse or premix fuel into an air flow through the combustor; and a secondary nozzle configured to premix fuel with the air flow. Each premixing nozzle includes a center body, at least one vane, a burner tube provided around the center body, at least two cooling passages, a fuel cooling passage to cool surfaces of the center body and the at least one vane, and an air cooling passage to cool a wall of the burner tube. The cooling passages prevent the walls of the center body, the vane(s), and the burner tube from overheating during flame holding events.

Potential Evaluation of Solar Heat Assisted Desiccant Hybrid Air Conditioning System

NASA Astrophysics Data System (ADS)

Tran, Thien Nha; Hamamoto, Yoshinori; Akisawa, Atsushi; Kashiwagi, Takao

The solar thermal driven desiccant dehumidification-absorption cooling hybrid system has superior advantage in hot-humid climate regions. The reasonable air processing of desiccant hybrid air conditioning system and the utility of clean and free energy make the system environment friendly and energy efficient. The study investigates the performance of the desiccant dehumidification air conditioning systems with solar thermal assistant. The investigation is performed for three cases which are combinations of solar thermal and absorption cooling systems with different heat supply temperature levels. Two solar thermal systems are used in the study: the flat plate collector (FPC) and the vacuum tube with compound parabolic concentrator (CPC). The single-effect and high energy efficient double-, triple-effect LiBr-water absorption cooling cycles are considered for cooling systems. COP of desiccant hybrid air conditioning systems are determined. The evaluation of these systems is subsequently performed. The single effect absorption cooling cycle combined with the flat plate collector solar system is found to be the most energy efficient air conditioning system.

Method of Operating a Neutronic Reactor

NASA Astrophysics Data System (ADS)

Fermi, Enrico; Szilard, Leo

This Patent is a later,1 almost faithful, copy of Patent No. 2,708,656 (which is then not reported in the present volume). This revised version was probably prepared (by the authors) in order to correct several misprints of the previous version. As emphasized in The New York Times of May 19, 1955, Patent No. 2,708,656, an "historic Patent, covering the first nuclear reactor", is the first one on this topic issued by the U.S. Patent Office, and served as a reference for the subsequent Patents on the same subject. In this long Patent, the theory, exper- imental data and principles of construction and operation of "any" type of nuclear reactor known at that time are discussed in an extremely detailed way. Various possible fission fragments produced by the reactor, several forms of the uranium employed (metal, oxide and so on, grouped in different geometrical forms), various materials adopted as moderators, several cooling systems, different geometries of the reactors, etc. are considered accurately. The theoretical description, centered around the achievement of a self-sustaining chain reaction, is exhaustive, and great attention is devoted to any possible cause of neutron loss, to the resonance capture of neutrons and to the effect of the presence of relevant impurities in the reactor. The chain production of neutrons in the pile is described in great detail, along with the theoretical arguments underlying the exponential experiment. The problem of the variation of the multiplication factor due to the production of radioactive elements, such as xenon, is discussed extensively. In particular it is pointed out that, although the initial production of xenon lowers the multiplication factor K due to its relevant neutron absorption, it subsequently increases again due to the decay of xenon into another isotope which absorbs fewer neutrons. The building up of reactors with solid (graphite) or liquid (heavy water) moderators is discussed, as well as other possible moderators such as light water or beryllium. In particular, the ratio is given of the absorption cross section to the scattering cross section for several moderators. Procedures for the purification of uranium are described as well. Several methods (i.e., the exponential pile or the "shotgun" method; see Patent No. 2.969,307) are reported for testing the purity against neutron absorption of different materials. The effect of the boron and vanadium impurities in the graphite and light water in the heavy water are considered. Different cooling systems for the reactors are considered and compared in the Patent, based on the circulation of a gas (typically, air) or a liquid (light or heavy water, diphenyl, etc.). The principles and practice for the construction, functioning and control of several kinds of reactors are reported in detail. One reactor considered in the present Patent is a low power uranium-graphite one without cooling system, where the active part consists in (small) cylinders of metallic uranium or pseudo-spheres of uranium oxide (or cylinders of U3O8). The control rods are made of steel with boron inserts, while limitation and safety rods are made of cadmium. In addition, an uranium-graphite pile cooled by air or even by water or diphenyl is considered. It is pointed out that dyphenil should usually be preferred with respect to water, due to its lower absorption of neutrons and to its higher boiling temperature, but the disadvantage related to its use is mainly due to the closed pumping system required and to the possible occurrence of polymerization which makes the fluid viscous. Another kind of reactor described in detail is made of uranium (vertical) bars immersed in heavy water. When, during the operation, heavy water is dissociated into D2 and O2, these two gaseous elements are carried by an inert gas (helium) into a recombination device. The control and safety rods are made of cadmium. Hybrid reactors composed of different lattices in the same neutronic reactor, in order to increase the multiplication factor K, are considered as well. A description of the possible uses of nuclear reactors, other than as power supplies, including the production of collimated beams of fast neutrons, the production of plutonium (a fissionable material usable in other reactors) or several other radioactive isotopes (for possible utilization in medicine) is as well given. As it results clear, no published reference article behind the present Patent exists. Some partial results may be found in several papers2 of Volume II of [Fermi (1962)] (see, for example, [Fermi (1952)]), but here very many technical data and some information of historic interest (mainly on the experiments performed in order to obtain the data reported) are given. The most "relevant" change of Patent No. 2,798,847 with respect to the original Patent No. 2,708,656 is the replacement of the 8 claims of the previous one by the following only one claim, which well summarizes the work done: "A method of operating a neutronic reactor including an active portion having a neutron reproduction ratio substantially in excess of unity in the absence of high neutron absorbing bodies, said method comprising the steps of inserting in the active portion a shim member consisting essentially of a high neutron absorbing body in an amount to reduce the neutron reproduction ratio to a value slightly higher than unit to prevent a dangerous reactivity level, controlling the reaction by moving a control member consisting essentially of a second high neutron absorbing body inwardly and outwardly in response to variations in neutron density, to maintain the neutron reproduction ratio substantially at unity, and withdrawing successive portions of the shim member to the extent necessary to enable the reactor to be controlled by movement of the control member after the neutron reproduction value has been lowered to the point where the outward movement of the control member is insufficient to maintain the neutron reproduction ratio at the desired point, and thus to maintain the range of control effected by such movement of the control member substantially constant despite diminution of neutron reproduction ratio caused by operation of the reactor, the active portion being substantially free of high neutron absorber other than the control member and the shim member."

Simulation study of air and water cooled photovoltaic panel using ANSYS

NASA Astrophysics Data System (ADS)

Syafiqah, Z.; Amin, N. A. M.; Irwan, Y. M.; Majid, M. S. A.; Aziz, N. A.

2017-10-01

Demand for alternative energy is growing due to decrease of fossil fuels sources. One of the promising and popular renewable energy technology is a photovoltaic (PV) technology. During the actual operation of PV cells, only around 15% of solar irradiance is converted to electricity, while the rest is converted into heat. The electrical efficiency decreases with the increment in PV panel’s temperature. This electrical energy is referring to the open-circuit voltage (Voc), short-circuit current (Isc) and output power generate. This paper examines and discusses the PV panel with water and air cooling system. The air cooling system was installed at the back of PV panel while water cooling system at front surface. The analyses of both cooling systems were done by using ANSYS CFX and PSPICE software. The highest temperature of PV panel without cooling system is 66.3 °C. There is a decrement of 19.2% and 53.2% in temperature with the air and water cooling system applied to PV panel.

Combustion system for hybrid solar fossil fuel receiver

DOEpatents

Mehos, Mark S.; Anselmo, Kenneth M.; Moreno, James B.; Andraka, Charles E.; Rawlinson, K. Scott; Corey, John; Bohn, Mark S.

2004-05-25

A combustion system for a hybrid solar receiver comprises a pre-mixer which combines air and fuel to form an air-fuel mixture. The mixture is introduced tangentially into a cooling jacket. A burner plenum is fluidically connected to the cooling jacket such that the burner plenum and the cooling jacket are arranged in thermal contact with one another. The air-fuel mixture flows through the cooling jacket cooling the burner plenum to reduce pre-ignition of the air-fuel mixture in the burner plenum. A combustion chamber is operatively associated with and open to the burner plenum to receive the air-fuel mixture from the burner plenum. An igniter is operatively positioned in the combustion chamber to combust the air-fuel mixture, releasing heat. A recuperator is operatively associated with the burner plenum and the combustion chamber and pre-heats the air-fuel mixture in the burner plenum with heat from the combustion chamber. A heat-exchanger is operatively associated and in thermal contact with the combustion chamber. The heat-exchanger provides heat for the hybrid solar receiver.

Light shield and cooling apparatus. [high intensity ultraviolet lamp

NASA Technical Reports Server (NTRS)

Meador, T. G., Jr. (Inventor)

1974-01-01

A light shield and cooling apparatus was developed for a high intensity ultraviolet lamp including water and high pressure air for cooling and additional apparatus for shielding the light and suppressing the high pressure air noise.

Center for the Built Environment: UFAD Cooling Load Design Tool

Science.gov Websites

Energy Publications Project Title: Underfloor Air Distribution (UFAD) Cooling Load Design Tool Providing . Webster, 2010. Development of a simplified cooling load design tool for underfloor air distribution Near-ZNE Buildings Setpoint Energy Savings Calculator UFAD Case Studies UFAD Cooling Design Tool UFAD

A roadmap for nuclear energy technology

NASA Astrophysics Data System (ADS)

Sofu, Tanju

2018-01-01

The prospects for the future use of nuclear energy worldwide can best be understood within the context of global population growth, urbanization, rising energy need and associated pollution concerns. As the world continues to urbanize, sustainable development challenges are expected to be concentrated in cities of the lower-middle-income countries where the pace of urbanization is fastest. As these countries continue their trajectory of economic development, their energy need will also outpace their population growth adding to the increased demand for electricity. OECD IEA's energy system deployment pathway foresees doubling of the current global nuclear capacity by 2050 to reduce the impact of rapid urbanization. The pending "retirement cliff" of the existing U.S. nuclear fleet, representing over 60 percent of the nation's emission-free electricity, also poses a large economic and environmental challenge. To meet the challenge, the U.S. DOE has developed the vision and strategy for development and deployment of advanced reactors. As part of that vision, the U.S. government pursues programs that aim to expand the use of nuclear power by supporting sustainability of the existing nuclear fleet, deploying new water-cooled large and small modular reactors to enable nuclear energy to help meet the energy security and climate change goals, conducting R&D for advanced reactor technologies with alternative coolants, and developing sustainable nuclear fuel cycle strategies. Since the current path relying heavily on water-cooled reactors and "once-through" fuel cycle is not sustainable, next generation nuclear energy systems under consideration aim for significant advances over existing and evolutionary water-cooled reactors. Among the spectrum of advanced reactor options, closed-fuel-cycle systems using reactors with fast-neutron spectrum to meet the sustainability goals offer the most attractive alternatives. However, unless the new public-private partnership models emerge to tackle the licensing and demonstration challenges for these advanced reactor concepts, realization of their enormous potential is not likely, at least in the U.S.

Verification of RELAP5-3D code in natural circulation loop as function of the initial water inventory

NASA Astrophysics Data System (ADS)

Bertani, C.; Falcone, N.; Bersano, A.; Caramello, M.; Matsushita, T.; De Salve, M.; Panella, B.

2017-11-01

High safety and reliability of advanced nuclear reactors, Generation IV and Small Modular Reactors (SMR), have a crucial role in the acceptance of these new plants design. Among all the possible safety systems, particular efforts are dedicated to the study of passive systems because they rely on simple physical principles like natural circulation, without the need of external energy source to operate. Taking inspiration from the second Decay Heat Removal system (DHR2) of ALFRED, the European Generation IV demonstrator of the fast lead cooled reactor, an experimental facility has been built at the Energy Department of Politecnico di Torino (PROPHET facility) to study single and two-phase flow natural circulation. The facility behavior is simulated using the thermal-hydraulic system code RELAP5-3D, which is widely used in nuclear applications. In this paper, the effect of the initial water inventory on natural circulation is analyzed. The experimental time behaviors of temperatures and pressures are analyzed. The experimental matrix ranges between 69 % and 93%; the influence of the opposite effects related to the increase of the volume available for the expansion and the pressure raise due to phase change is discussed. Simulations of the experimental tests are carried out by using a 1D model at constant heat power and fixed liquid and air mass; the code predictions are compared with experimental results. Two typical responses are observed: subcooled or two phase saturated circulation. The steady state pressure is a strong function of liquid and air mass inventory. The numerical results show that, at low initial liquid mass inventory, the natural circulation is not stable but pulsated.

Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

NASA Astrophysics Data System (ADS)

Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

2017-01-01

The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

Fuel pins with both target and fuel pellets in an isotope-production reactor

DOEpatents

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

1982-08-19

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

NEUTRONIC REACTORS

DOEpatents

Anderson, H.L.

1958-10-01

The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

Dry coolers and air-condensing units (Review)

NASA Astrophysics Data System (ADS)

Milman, O. O.; Anan'ev, P. A.

2016-03-01

The analysis of factors affecting the growth of shortage of freshwater is performed. The state and dynamics of the global market of dry coolers used at electric power plants are investigated. Substantial increase in number and maximum capacity of air-cooled condensers, which have been put into operation in the world in recent years, are noted. The key reasons facilitating the choice of developers of the dry coolers, in particular the independence of the location of thermal power plant from water sources, are enumerated. The main steam turbine heat removal schemes using air cooling are considered, their comparison of thermal efficiency is assessed, and the change of three important parameters, such as surface area of heat transfer, condensate pump flow, and pressure losses in the steam exhaust system, are estimated. It is shown that the most effective is the scheme of direct steam condensation in the heat-exchange tubes, but other schemes also have certain advantages. The air-cooling efficiency may be enhanced much more by using an air-cooling hybrid system: a combination of dry and wet cooling. The basic applied constructive solutions are shown: the arrangement of heat-exchange modules and the types of fans. The optimal mounting design of a fully shopassembled cooling system for heat-exchange modules is represented. Different types of heat-exchange tubes ribbing that take into account the operational features of cooling systems are shown. Heat transfer coefficients of the plants from different manufacturers are compared, and the main reasons for its decline are named. When using evaporative air cooling, it is possible to improve the efficiency of air-cooling units. The factors affecting the faultless performance of dry coolers (DC) and air-condensing units (ACU) and the ways of their elimination are described. A high velocity wind forcing reduces the efficiency of cooling systems and creates preconditions for the development of wind-driven devices. It is noted that global trends have a significant influence on the application of dry coolers in Russia, in view of the fact that some TPP have a surface condensers arrangement. The reasons that these systems are currently less efficient than the direct steam condensation in an air-cooled condenser are explained. It is shown that, in some cases, it is more reasonable to use mixing-type condensers in combination with a dry cooler. Measures for a full import substitution of steam exhaust heat removal systems are mentioned.

Experimental study of an air-cooled thermal management system for high capacity lithium-titanate batteries

NASA Astrophysics Data System (ADS)

Giuliano, Michael R.; Prasad, Ajay K.; Advani, Suresh G.

2012-10-01

Lithium-titanate batteries have become an attractive option for battery electric vehicles and hybrid electric vehicles. In order to maintain safe operating temperatures, these batteries must be actively cooled during operation. Liquid-cooled systems typically employed for this purpose are inefficient due to the parasitic power consumed by the on-board chiller unit and the coolant pump. A more efficient option would be to circulate ambient air through the battery bank and directly reject the heat to the ambient. We designed and fabricated such an air-cooled thermal management system employing metal-foam based heat exchanger plates for sufficient heat removal capacity. Experiments were conducted with Altairnano's 50 Ah cells over a range of charge-discharge cycle currents at two air flow rates. It was found that an airflow of 1100 mls-1 per cell restricts the temperature rise of the coolant air to less than 10 °C over ambient even for 200 A charge-discharge cycles. Furthermore, it was shown that the power required to drive the air through the heat exchanger was less than a conventional liquid-cooled thermal management system. The results indicate that air-cooled systems can be an effective and efficient method for the thermal management of automotive battery packs.

Design, fabrication, and testing of an external fuel (UO2), full-length thermionic converter

NASA Technical Reports Server (NTRS)

Schock, A.; Raab, B.

1971-01-01

The development of a full-length external-fuel thermionic converter for in-pile testing is described. The development program includes out-of-pile performance testing of the fully fueled-converter, using RF-induction heating, before its installation in the in-pile test capsule. The external-fuel converter is cylindrical in shape, and consists of an inner, centrally cooled collector, and an outer emitter surrounded by nuclear fuel. The term full-length denotes that the converter is long enough to extend over the full height of the reactor core. Thus, the converter is not a scaled-down test device, but a full-scale fuel element of the thermionic reactor. The external-fuel converter concept permits a number of different design options, particularly with respect to the fuel composition and shape, and the collector cooling arrangement. The converter described was developed for the Jet Propulsion Laboratory, and is based on their concept for a thermionic reactor with uninsulated collector cooling as previously described. The converter is double-ended, with through-flow cooling, and with ceramic seals and emitter and collector power take-offs at both ends. The design uses a revolver-shaped tungsten emitter body, with the central emitter hole surrounded by six peripheral fuel holes loaded with cylindrical UO2 pellets.

Application of Molten Salt Reactor Technology to MMW In-Space NEP and Surface Power Missions

NASA Technical Reports Server (NTRS)

Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen (Technical Monitor)

2002-01-01

Anticipated manned nuclear electric propulsion (NEP) and planetary surface power missions will require multimegawatt nuclear reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional multimegawatt gas-cooled and liquid metal concepts.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

2002-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

1998-01-01

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Fast quench reactor and method

DOEpatents

Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

2002-09-24

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buden, D.

1993-08-11

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressedmore » 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.« less

Summary of space nuclear reactor power systems, 1983 - 1992

NASA Astrophysics Data System (ADS)

Buden, D.

1993-08-01

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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Z.; Southwest Science and Technology Univ., No.350 Shushanhu Road, Shushan District, Hefei, Anhui, 230031; Chen, Y.

2012-07-01

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 adjustedmore » 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)« less

Ichthyoplankton entrainment study at the SRS Savannah River water intakes for Westinghouse Savannah River Company. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paller, M.

1992-03-26

Cooling water for L and K Reactors and makeup water for Par Pond is pumped from the Savannah River at the 1G, 3G, and 5G pump houses. Ichthyoplankton (drifting fish larvae and eggs) from the river are entrained into the reactor cooling systems with the river water and passed through the reactor`s heat exchangers where temperatures may reach 70{degrees}C during full power operation. Ichthyoplankton mortality under such conditions is assumed to be 100 percent. The number of ichthyoplankton entrained into the cooling system depends on a variety of variables, including time of year, density and distribution of ichthyoplankton in themore » river, discharge levels in the river, and the volume of water withdrawn by the pumps. Entrainment at the 1 G pump house, which is immediately downstream from the confluence of Upper Three Runs Creek and the Savannah River, is also influenced by discharge rates and ichthyoplankton densities in Upper Three Runs Creek. Because of the anticipated restart of several SRS reactors and the growing concern surrounding striped bass and American shad stocks in the Savannah River, the Department of Energy requested that the Environmental Sciences Section (ESS) of the Savannah River Laboratory sample ichthyoplankton at the SRS Savannah River intakes. Dams & Moore, Inc., under a contract with Westinghouse Savannah River Company performed the sampling and data analysis for the ESS.« less

Passive wall cooling panel with phase change material as a cooling agent

NASA Astrophysics Data System (ADS)

Majid, Masni A.; Tajudin, Rasyidah Ahmad; Salleh, Norhafizah; Hamid, Noor Azlina Abd

2017-11-01

The study was carried out to the determine performance of passive wall cooling panels by using Phase Change Materials as a cooling agent. This passive cooling system used cooling agent as natural energy storage without using any HVAC system. Eight full scale passive wall cooling panels were developed with the size 1500 mm (L) × 500 mm (W) × 100 mm (T). The cooling agent such as glycerine were filled in the tube with horizontal and vertical arrangement. The passive wall cooling panels were casting by using foamed concrete with density between 1200 kg/m3 - 1500 kg/m3. The passive wall cooling panels were tested in a small house and the differences of indoor and outdoor temperature was recorded. Passive wall cooling panels with glycerine as cooling agent in vertical arrangement showed the best performance with dropped of indoor air temperature within 3°C compared to outdoor air temperature. The lowest indoor air temperature recorded was 25°C from passive wall cooling panels with glycerine in vertical arrangement. From this study, the passive wall cooling system could be applied as it was environmental friendly and less maintenance.

Computing Cooling Flows in Turbines

NASA Technical Reports Server (NTRS)

Gauntner, J.

1986-01-01

Algorithm developed for calculating both quantity of compressor bleed flow required to cool turbine and resulting decrease in efficiency due to cooling air injected into gas stream. Program intended for use with axial-flow, air-breathing, jet-propulsion engines with variety of airfoil-cooling configurations. Algorithm results compared extremely well with figures given by major engine manufacturers for given bulk-metal temperatures and cooling configurations. Program written in FORTRAN IV for batch execution.

Sampling and detection of Legionella pneumophila aerosols generated from an industrial cooling tower.

PubMed

Ishimatsu, S; Miyamoto, H; Hori, H; Tanaka, I; Yoshida, S

2001-08-01

Cooling tower water has frequently been cited as a source of infection in outbreaks of Legionnaires' disease. However, there have been few reports on the presence of legionellae in aerosols from cooling towers. This paper describes our use of an impinger or a six-stage microbial impactor for detecting legionellae in air around a cooling tower contaminated with L. pneumophila (1.2+/-0.3x10(5) CFU/100 ml). Phosphate-buffered saline, Page's saline, 2% yeast extract solution and buffered yeast extract (BYE) broth were tested to evaluate their collection efficiency. These solutions were compared in laboratory experiments using an aerosol of L. pneumophila serogroup (SG) 1. Because BYE broth was the most efficient and storable collecting fluid among them, it was used for outdoor air sampling. In the outdoor air sampling, aerosolized L. pneumophila SG 6 was detected in the air around the cooling tower by the impinger (0.09 CFU/l. air). No legionellae were detected by the impactor with Legionella-selective agar plates (WYOalpha) because the plates were overgrown with fungi. Repetitive element PCR (rep-PCR) and arbitrarily primed PCR (AP-PCR) were employed to assess the epidemiological relationship among Legionella isolates from the air sample and the cooling tower water samples. L. pneumophila SG 6 isolated from the aerosols produced rep-PCR and AP-PCR fingerprints identical to those of L. pneumophila SG 6 strains from the cooling tower water, suggesting that the bacterium was aerosolized from the cooling tower.

Preliminary design of an alternate high-temperature turbine. A topical report for Phase II of the High-Temperature-Turbine Technology Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strough, R.I.

The feasibility of designing a convectively air-cooled turbine to operate in the environment of a 3000/sup 0/F combustor exit temperature with maximum turbine airfoil metal temperatures held to 1500/sup 0/F was established. The United Technologies-Kraftwerk Union V84.3 gas turbine design was used as the basic configuration for the design of the 3000/sup 0/F turbine. Turbine cooling requirements were determined based on the use of the modified V84.3 type silo combustor with a pattern factor of 0.1. The convective air-cooling technology levels in terms of cooling effectiveness required to satisfy the airfoil cooling requirements were identified. Cooling schemes and fabrication technologiesmore » required are discussed. Turbine airfoil cooling technology levels required for the 3000/sup 0/F engine were selected. The performance of the 3000/sup 0/F convectively air-cooled gas turbine in simple and combined cycle was calculated. The 3000/sup 0/F gas turbine combined-cycle system provides an increase in power of 61% and a decrease in heat rate of 10% compared to a similar system with a combustor exit temperature of 2210/sup 0/F and the same airflow. The development of a successful 3000/sup 0/F convectively air-cooled turbine can be accomplished with a reasonable design and fabrication development effort on the cooled turbine airfoils. Use of the convectively air-cooled turbine provides the transfer of technology from extensive aircraft engines developed programs and operating experience to industrial gas turbines. It eliminates the requirement for large investments in alternate cooling techniques tailored specifically for industrial engines which offer no additional benefits.« less

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1995-08-01

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fitzpatrick, F.C.; Gray, D.D.; Hyndman, J.R.

The thermal, ecological, and social impacts of a 40-reactor NEC are compared to impacts from four 10-reactor NECs and ten 4-reactor power plants. The comparison was made for surrogate sites in western Tennessee. The surrogate site for the 40-reactor NEC is located on Kentucky Lake. A layout is postulated for ten clusters of four reactors each with 2.5-mile spacing between clusters. The plants use natural-draft cooling towers. A transmission system is proposed for delivering the power (48,000 MW) to five load centers. Comparable transmission systems are proposed for the 10-reactor NECs and the 4-reactor dispersed sites delivering power to themore » same load centers. (auth)« less

Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forsberg, C.W.; Reich, W.J.

1991-09-01

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactormore » concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.« less

Interfacial area transport of steam-water two-phase flow in a vertical annulus at elevated pressures

NASA Astrophysics Data System (ADS)

Ozar, Basar

Analysis of accident scenarios in nuclear reactors are done by using codes such as TRACE and RELAP5. Large oscillations in the core void fraction are observed in calculations of advanced passive light water reactors (ALWRs), especially during the low pressure long-term cooling phase. These oscillations are attributed to be numerical in nature and served to limit the accuracy as well as the credibility of the calculations. One of the root causes of these unphysical oscillations is determined to be flow regime transitions caused by the usage of static flow regime maps. The interfacial area transport equation was proposed earlier in order to address these issues. Previous research successfully developed the foundation of the interfacial area transport equation and the experimental techniques needed for the measurement of interfacial area, bubble diameters and velocities. In the past, an extensive database has been then generated for adiabatic air-water conditions in vertical upward and downward bubbly-churn turbulent flows in pipes. Using this database, mechanistic models for the creation (bubble breakup) and destruction (bubble coalescence) of interfacial area have been developed for the bubblyslug flow regime transition. However, none of these studies investigated the effect of phase change. To address this need, a heated annular test section was designed and constructed. The design relied on a three level scaling approach: geometric scaling; hydrodynamic scaling; thermal scaling. The test section consisted of a heated and unheated section in order to study the sub-cooled boiling and bulk condensation/flashing and evaporation phenomena, respectively. Steam-water two-phase flow tests were conducted under sub-cooled boiling conditions in the heated section and with sub-cooled/super-heated bulk liquid in the unheated section. The modeling of interfacial area transport equation with phase change effects was introduced and discussed. Constitutive relations, which took phase change effects into account, for interfacial area transport equation were proposed and implemented. Effects of these constitutive relations on the prediction capability of the transport equation were discussed.

Dynamics of heat-pipe reactors

NASA Technical Reports Server (NTRS)

Niederauer, G. F.

1971-01-01

A split-core heat pipe reactor, fueled with either U(233)C or U(235)C in a tungsten cermet and cooled by 7-Li-W heat pipes, was examined for the effects of the heat pipes on reactor while trying to safely absorb large reactivity inputs through inherent shutdown mechanisms. Limits on ramp reactivity inputs due to fuel melting temperature and heat pipe wall heat flux were mapped for the reactor in both startup and at-power operating modes.

HEDL FACILITIES CATALOG 400 AREA

DOE Office of Scientific and Technical Information (OSTI.GOV)

MAYANCSIK BA

1987-03-01

The purpose of this project is to provide a sodium-cooled fast flux test reactor designed specifically for irradiation testing of fuels and materials and for long-term testing and evaluation of plant components and systems for the Liquid Metal Reactor (LMR) Program. The FFTF includes the reactor, heat removal equipment and structures, containment, core component handling and examination, instrumentation and control, and utilities and other essential services. The complex array of buildings and equipment are arranged around the Reactor Containment Building.

Modeling evaporation from spent nuclear fuel storage pools: A diffusion approach

NASA Astrophysics Data System (ADS)

Hugo, Bruce Robert

Accurate prediction of evaporative losses from light water reactor nuclear power plant (NPP) spent fuel storage pools (SFPs) is important for activities ranging from sizing of water makeup systems during NPP design to predicting the time available to supply emergency makeup water following severe accidents. Existing correlations for predicting evaporation from water surfaces are only optimized for conditions typical of swimming pools. This new approach modeling evaporation as a diffusion process has yielded an evaporation rate model that provided a better fit of published high temperature evaporation data and measurements from two SFPs than other published evaporation correlations. Insights from treating evaporation as a diffusion process include correcting for the effects of air flow and solutes on evaporation rate. An accurate modeling of the effects of air flow on evaporation rate is required to explain the observed temperature data from the Fukushima Daiichi Unit 4 SFP during the 2011 loss of cooling event; the diffusion model of evaporation provides a significantly better fit to this data than existing evaporation models.

Air Conditioner Ready to Change Industry - Continuum Magazine | NREL

Science.gov Websites

create very dry air, ideal for cooling with evaporative techniques. Desiccants, which can be liquids or into an innovative "cooling core." This would marry the desiccants' capacity to create dry air using heat and evaporative coolers' capability to turn dry air into cold air. If it worked, it

Thermal baffle for fast-breeder reacton

DOEpatents

Rylatt, John A.

1977-01-01

A liquid-metal-cooled fast-breeder reactor includes a bridge structure for separating hot outlet coolant from relatively cool inlet coolant consisting of an annular stainless steel baffle plate extending between the core barrel surrounding the core and the thermal liner associated with the reactor vessel and resting on ledges thereon, there being inner and outer circumferential webs on the lower surface of the baffle plate and radial webs extending between the circumferential webs, a stainless steel insulating plate completely covering the upper surface of the baffle plate and flex seals between the baffle plate and the ledges on which the baffle plate rests to prevent coolant from washing through the gaps therebetween. The baffle plate is keyed to the core barrel for movement therewith and floating with respect to the thermal liner and reactor vessel.

Desiccant Enhanced Evaporative Air-Conditioning (DEVap): Evaluation of a New Concept in Ultra Efficient Air Conditioning

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kozubal, E.; Woods, J.; Burch, J.

2011-01-01

NREL has developed the novel concept of a desiccant enhanced evaporative air conditioner (DEVap) with the objective of combining the benefits of liquid desiccant and evaporative cooling technologies into an innovative 'cooling core.' Liquid desiccant technologies have extraordinary dehumidification potential, but require an efficient cooling sink. DEVap's thermodynamic potential overcomes many shortcomings of standard refrigeration-based direct expansion cooling. DEVap decouples cooling and dehumidification performance, which results in independent temperature and humidity control. The energy input is largely switched away from electricity to low-grade thermal energy that can be sourced from fuels such as natural gas, waste heat, solar, or biofuels.

Closed loop air cooling system for combustion turbines

DOEpatents

Huber, David John; Briesch, Michael Scot

1998-01-01

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

Closed loop air cooling system for combustion turbines

DOEpatents

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

1998-07-21

Convective cooling of turbine hot parts using a closed loop system is disclosed. Preferably, the present invention is applied to cooling the hot parts of combustion turbine power plants, and the cooling provided permits an increase in the inlet temperature and the concomitant benefits of increased efficiency and output. In preferred embodiments, methods and apparatus are disclosed wherein air is removed from the combustion turbine compressor and delivered to passages internal to one or more of a combustor and turbine hot parts. The air cools the combustor and turbine hot parts via convection and heat is transferred through the surfaces of the combustor and turbine hot parts. 1 fig.

Serial cooling of a combustor for a gas turbine engine

DOEpatents

Abreu, Mario E.; Kielczyk, Janusz J.

2001-01-01

A combustor for a gas turbine engine uses compressed air to cool a combustor liner and uses at least a portion of the same compressed air for combustion air. A flow diverting mechanism regulates compressed air flow entering a combustion air plenum feeding combustion air to a plurality of fuel nozzles. The flow diverting mechanism adjusts combustion air according to engine loading.

Effect of Several Factors on the Cooling of a Radial Engine in Flight

NASA Technical Reports Server (NTRS)

Schey, Oscar W; Pinkel, Benjamin

1936-01-01

Flight tests of a Grumman Scout (XSF-2) airplane fitted with a Pratt & Whitney 1535 supercharged engine were conducted to determine the effect of engine power, mass flow of the cooling air, and atmospheric temperature on cylinder temperature. The tests indicated that the difference in temperature between the cylinder wall and the cooling air varied as the 0.38 power of the brake horsepower for a constant mass flow of cooling air, cooling-air temperature, engine speed, and brake fuel consumption. The difference in temperature was also found to vary inversely as the 0.39 power of the mass flow for points on the head and the 0.35 power for points on the barrel, provided that engine power, engine speed, brake fuel consumption, and cooling-air temperature were kept constant. The results of the tests of the effect of atmospheric temperature on cylinder temperature were inconclusive owing to unfavorable weather conditions prevailing at the time of the tests. The method used for controlling the test conditions, however, was found to be feasible.

Municipal Waste Incinerator Public Works Center, Yokosuka Japan Evaluation and Recommendations

DTIC Science & Technology

1993-04-01

Incinerator and Pollution Control Equipment 24 XIV. Gas Cooling Chamber Water Injection Sites and Control Valve 25 XV. Quencher Reactor 27 XVI...discussed below.I 11I.B.1. Exhaust Gas Cooling Chamber Within the exhaust gas cooling chamber, water is atomized into the gas stream cools the gases...as it evaporates. The feed rate of water is controlled to provide gases entering the quencher at 3000C (Figure XIV). The gases exit the exhaust gas

Technical note on results of production test IP-628-AI: Crosstie flow test and flushing calibration, and the earlier production tests IP-573-AI and IP-499-AI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jones, S.S.

1964-06-15

A series of three experimental tests was conducted on the operation and adequacy of the K-Reactors` third, or last-ditch, cooling system. The first test showed considerable line corrosion the second test was performed directly after line cleaning, and the third test showed a significant amount of additional line corrosion after only nine months of service. The present cooling adequacy of this last-ditch system at the KE and KW reactors is summarized these show the power levels for which we have adequate last-ditch cooling as a function of the crosstie coolant temperature. These figures include the effects of increasing the numbermore » of pumps that remain in operation at the other K-Reactor, and various other operating or emergency conditions. These curves are for diesel pump speeds of 750 rpm which are planned for June of this year. In these figures the crosstie temperature is assumed to be a conservative 5{degrees}C above the process water inlet temperature.« less

THERMAL PROPERTIES AND HEATING AND COOLING DURABILITY OF REACTOR SHIELDING CONCRETE (in Japanese)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hosoi, J.; Chujo, K.; Saji, K.

1959-01-01

A study was made of the thermal properties of various concretes made of domestic raw materials for radiation shields of a power reactor and of a high- flux research reactor. The results of measurements of thermal expansion coefficient, specific heat, thermal diffusivity, thermal conductivity, cyclical heating, and cooling durability are described. Relationships between thermal properties and durability are discussed and several photographs of the concretes are given. It is shown that the heating and cooling durability of such a concrete which has a large thermal expansion coefficient or a considerable difference between the thermal expansion of coarse aggregate and themore » one of cement mortar part or aggregates of lower strength is very poor. The decreasing rates of bending strength and dynamical modulus of elasticity and the residual elongation of the concrete tested show interesting relations with the modified thermal stress resistance factor containing a ratio of bending strength and thermal expansion coefficient. The thermal stress resistance factor seems to depend on the conditions of heat transfer on the surface and on heat release in the concrete. (auth)« less

A Balanced-pressure Sliding Seal for Transfer of Pressurized Air Between Stationary and Rotating Parts

NASA Technical Reports Server (NTRS)

Curren, Arthur N; Cochran, Reeves P

1957-01-01

A combination sliding-ring and pressure-balancing seal capable of transferring pressurize air from stationary to rotating parts was developed and experimentally investigated at sliding velocities and cooling-air pressures up to 10,000 feet per minute and 38.3 pounds per square inch absolute, respectively. Leakage of cooling air was completely eliminated with an expenditure of balance air less than one-fourth the leakage loss of air from labyrinth seals under the same conditions. Additional cooling of the carbon-base seal rings was required, and the maximum wear rate on the rings was about 0.0005 inch per hour.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Powers, Jeffrey J.; Mueller, Don

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy researchmore » and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.« less

14 CFR 29.1109 - Carburetor air cooling.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 14 Aeronautics and Space 1 2012-01-01 2012-01-01 false Carburetor air cooling. 29.1109 Section 29... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Induction System § 29.1109 Carburetor air... to maintain the air temperature, at the carburetor inlet, at or below the maximum established value...

14 CFR 29.1109 - Carburetor air cooling.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 14 Aeronautics and Space 1 2011-01-01 2011-01-01 false Carburetor air cooling. 29.1109 Section 29... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Induction System § 29.1109 Carburetor air... to maintain the air temperature, at the carburetor inlet, at or below the maximum established value...

14 CFR 29.1109 - Carburetor air cooling.

Code of Federal Regulations, 2010 CFR

2010-01-01

... AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Induction System § 29.1109 Carburetor air... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Carburetor air cooling. 29.1109 Section 29... to maintain the air temperature, at the carburetor inlet, at or below the maximum established value...

Performance of Introducing Outdoor Cold Air for Cooling a Plant Production System with Artificial Light

PubMed Central

Wang, Jun; Tong, Yuxin; Yang, Qichang; Xin, Min

2016-01-01

The commercial use of a plant production system with artificial light (PPAL) is limited by its high initial construction and operation costs. The electric-energy consumed by heat pumps, applied mainly for cooling, accounts for 15–35% of the total electric-energy used in a PPAL. To reduce the electric-energy consumption, an air exchanger with low capacity (180 W) was used for cooling by introducing outdoor cold air. In this experiment, the indoor air temperature in two PPALs (floor area: 6.2 m2 each) was maintained at 25 and 20°C during photoperiod and dark period, respectively, for lettuce production. A null CO2 balance enrichment method was used in both PPALs. In one PPAL (PPALe), an air exchanger (air flow rate: 250 m3·h−1) was used along with a heat pump (cooling capacity: 3.2 kW) to maintain the indoor air temperature at the set-point. The other PPAL (PPALc) with only a heat pump (cooling capacity: 3.2 kW) was used for reference. Effects of introducing outdoor cold air on energy use efficiency, coefficient of performance (COP), electric-energy consumption for cooling and growth of lettuce were investigated. The results show that: when the air temperature difference between indoor and outdoor ranged from 20.2 to 30.0°C: (1) the average energy use efficiency of the air exchanger was 2.8 and 3.4 times greater than the COP of the heat pumps in the PPALe and PPALc, respectively; (2) hourly electric-energy consumption for cooling in the PPALe reduced by 15.8–73.7% compared with that in the PPALc; (3) daily supply of CO2 in the PPALe reduced from 0.15 to 0.04 kg compared with that in the PPALc with the outdoor air temperature ranging from −5.6 to 2.7°C; (4) no significant difference in lettuce growth was observed in both PPALs. The results indicate that using air exchanger to introduce outdoor cold air should be considered as an effective way to reduce electric-energy consumption for cooling with little effects on plant growth in a PPAL. PMID:27066012

An experimental study on the design, performance and suitability of evaporative cooling system using different indigenous materials

NASA Astrophysics Data System (ADS)

Alam, Md. Ferdous; Sazidy, Ahmad Sharif; Kabir, Asif; Mridha, Gowtam; Litu, Nazmul Alam; Rahman, Md. Ashiqur

2017-06-01

The present study aimed to evaluate the feasibility of coconut coir pads, jute fiber pads and sackcloth pads as alternative pad materials. Experimental measurements were conducted and the experimental data were quantitative. The experimental work mainly focused on the effects of different types and thicknesses of evaporative cooling pads by using forced draft fan while changing the environmental conditions. Experiments are conducted in a specifically constructed test chamber having dimensions of 12'X8'X8', using a number of cooling pads (36"X26") with a variable thickness parameters of the evaporative cooling pads i.e., 50, 75 and 100 mm. Moreover, the experimental work involved the measurement of environmental parameters such as temperature, relative humidity, air velocity, water mass flow rate and pressure drops at different times during the day. Experiments were conducted at three different water mass flow rates (0.25 kgs-1, 0.40 kgs-1 & 0.55 kgs-1) and three different air velocities (3.6 ms-1, 4.6 ms-1& 5.6 ms-1). There was a significant difference between evaporative cooling pad types and cooling efficiency. The coconut coir pads yielded maximum cooling efficiency of 85%, whereas other pads yielded the following maximum cooling efficiency: jute fiber pads 78% and sackcloth 69% for higher air velocity and minimum mass flow rate. It is found that the maximum reduction in temperature between cooling pad inlet and outlet is 4°C with a considerable increase in humidity. With the increase of pad thickness there was an increment of cooling efficiency. The results obtained for environmental factors, indicated that there was a significant difference between environmental factors and cooling efficiency. In terms of the effect of air velocity on saturation efficiency and pressure drop, higher air velocity decreases saturation efficiency and increases pressure drop across the wetted pad for maximum flow rate. Convective heat transfer co-efficient has an almost linear relationship with air Velocity. Water consumption or evaporation rate increases with the increase in air velocity. Finally, the present study indicated that the coconut coir pads perform better than the other evaporative cooling pads and have higher potential as wetted-pad material. The outcomes of this study can provide an effective and low-cost solution in the form of evaporative cooling system, especially in an agricultural country like Bangladesh.

Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

PubMed

Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

2017-12-01

The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

DTIC Science & Technology

2006-12-01

24 Table 3.3 Hazards of Sodium Reaction Products, Hydride And Oxide...........................26 Table 3.4 Chemical Reactivity Of Selected...Liquid Metal Fast Breeder Reactor ORIGEN Oak Ridge Isotope Generator ORIGENARP Oak Ridge Isotope Generator Automated Rapid Processing PWR ...nuclear reactors, both because of the possibility of increased reactivity due to boiling and the potential loss of effectiveness of coolant heat transfer

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Pulse detonation engines and components thereof

NASA Technical Reports Server (NTRS)

Tangirala, Venkat Eswarlu (Inventor); Rasheed, Adam (Inventor); Vandervort, Christian Lee (Inventor); Dean, Anthony John (Inventor)

2009-01-01

A pulse detonation engine comprises a primary air inlet; a primary air plenum located in fluid communication with the primary air inlet; a secondary air inlet; a secondary air plenum located in fluid communication with the secondary air inlet, wherein the secondary air plenum is substantially isolated from the primary air plenum; a pulse detonation combustor comprising a pulse detonation chamber, wherein the pulse detonation chamber is located downstream of and in fluid communication with the primary air plenum; a coaxial liner surrounding the pulse detonation combustor defining a cooling plenum, wherein the cooling plenum is in fluid communication with the secondary air plenum; an axial turbine assembly located downstream of and in fluid communication with the pulse detonation combustor and the cooling plenum; and a housing encasing the primary air plenum, the secondary air plenum, the pulse detonation combustor, the coaxial liner, and the axial turbine assembly.

40 CFR 60.611 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-07-01

... From the Synthetic Organic Chemical Manufacturing Industry (SOCMI) Air Oxidation Unit Processes § 60... given them. Air Oxidation Reactor means any device or process vessel in which one or more organic... compounds. Ammoxidation and oxychlorination reactions are included in this definition. Air Oxidation Reactor...

40 CFR 60.611 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-07-01

... From the Synthetic Organic Chemical Manufacturing Industry (SOCMI) Air Oxidation Unit Processes § 60... given them. Air Oxidation Reactor means any device or process vessel in which one or more organic... compounds. Ammoxidation and oxychlorination reactions are included in this definition. Air Oxidation Reactor...

40 CFR 60.611 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-07-01

... From the Synthetic Organic Chemical Manufacturing Industry (SOCMI) Air Oxidation Unit Processes § 60... given them. Air Oxidation Reactor means any device or process vessel in which one or more organic... compounds. Ammoxidation and oxychlorination reactions are included in this definition. Air Oxidation Reactor...

40 CFR 60.611 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... From the Synthetic Organic Chemical Manufacturing Industry (SOCMI) Air Oxidation Unit Processes § 60... given them. Air Oxidation Reactor means any device or process vessel in which one or more organic... compounds. Ammoxidation and oxychlorination reactions are included in this definition. Air Oxidation Reactor...

Geomechanical Analysis of Underground Coal Gasification Reactor Cool Down for Subsequent CO2 Storage

NASA Astrophysics Data System (ADS)

Sarhosis, Vasilis; Yang, Dongmin; Kempka, Thomas; Sheng, Yong

2013-04-01

Underground coal gasification (UCG) is an efficient method for the conversion of conventionally unmineable coal resources into energy and feedstock. If the UCG process is combined with the subsequent storage of process CO2 in the former UCG reactors, a near-zero carbon emission energy source can be realised. This study aims to present the development of a computational model to simulate the cooling process of UCG reactors in abandonment to decrease the initial high temperature of more than 400 °C to a level where extensive CO2 volume expansion due to temperature changes can be significantly reduced during the time of CO2 injection. Furthermore, we predict the cool down temperature conditions with and without water flushing. A state of the art coupled thermal-mechanical model was developed using the finite element software ABAQUS to predict the cavity growth and the resulting surface subsidence. In addition, the multi-physics computational software COMSOL was employed to simulate the cavity cool down process which is of uttermost relevance for CO2 storage in the former UCG reactors. For that purpose, we simulated fluid flow, thermal conduction as well as thermal convection processes between fluid (water and CO2) and solid represented by coal and surrounding rocks. Material properties for rocks and coal were obtained from extant literature sources and geomechanical testings which were carried out on samples derived from a prospective demonstration site in Bulgaria. The analysis of results showed that the numerical models developed allowed for the determination of the UCG reactor growth, roof spalling, surface subsidence and heat propagation during the UCG process and the subsequent CO2 storage. It is anticipated that the results of this study can support optimisation of the preparation procedure for CO2 storage in former UCG reactors. The proposed scheme was discussed so far, but not validated by a coupled numerical analysis and if proved to be applicable it could provide a significant optimisation of the UCG process by means of CO2 storage efficiency. The proposed coupled UCG-CCS scheme allows for meeting EU targets for greenhouse gas emissions and increases the coal yield otherwise impossible to exploit.

Active (air-cooled) vs. passive (phase change material) thermal management of high power lithium-ion packs: Limitation of temperature rise and uniformity of temperature distribution

NASA Astrophysics Data System (ADS)

Sabbah, Rami; Kizilel, R.; Selman, J. R.; Al-Hallaj, S.

The effectiveness of passive cooling by phase change materials (PCM) is compared with that of active (forced air) cooling. Numerical simulations were performed at different discharge rates, operating temperatures and ambient temperatures of a compact Li-ion battery pack suitable for plug-in hybrid electric vehicle (PHEV) propulsion. The results were also compared with experimental results. The PCM cooling mode uses a micro-composite graphite-PCM matrix surrounding the array of cells, while the active cooling mode uses air blown through the gaps between the cells in the same array. The results show that at stressful conditions, i.e. at high discharge rates and at high operating or ambient temperatures (for example 40-45 °C), air-cooling is not a proper thermal management system to keep the temperature of the cell in the desirable operating range without expending significant fan power. On the other hand, the passive cooling system is able to meet the operating range requirements under these same stressful conditions without the need for additional fan power.

Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Pointer, William David; Sieger, Matt

2016-04-01

The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.