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1

Stability analysis of supercritical water cooled reactors  

E-Print Network (OSTI)

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

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

2005-01-01T23:59:59.000Z

2

Electrochemistry of Water-Cooled Nuclear Reactors  

DOE Green Energy (OSTI)

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

Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy, Amit Jain, Han Sang Kim, Vishisht Gupta; Jonathan Pitt

2006-08-08T23:59:59.000Z

3

Candidate Materials Evaluation for Supercritical Water-Cooled Reactor  

SciTech Connect

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

T. R. Allen and G. S. Was

2008-12-12T23:59:59.000Z

4

Development of Materials for Supercritical-Water-Cooled Reactor |  

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

Development of Materials for Supercritical-Water-Cooled Reactor Development of Materials for Supercritical-Water-Cooled Reactor Development of Materials for Supercritical-Water-Cooled Reactor Supercritical-Water-Cooled Reactor (SCWR) was selected as one of the promising candidates in Generation IV reactors for its prominent advantages; those are the high thermal efficiency, the system simplification, the R&D cost minimization and the flexibility for core design. As the demand for advanced nuclear system increases, Japanese R&D project started in 1999 aiming to provide technical information essential to demonstration of SCPR technologies through three sub-themes of 1. Plant conceptual design, 2. Thermal-hydraulics, and 3. Material. Although the material development is critical issue of SCWR development, previous studies were limited for the screening tests on commercial alloys

5

REACTOR COOLING  

DOE Patents (OSTI)

A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

Quackenbush, C.F.

1959-09-29T23:59:59.000Z

6

Program on Technology Innovation: Cooling Water Review of the Advanced Light Water Reactor Utility Requirements Document  

Science Conference Proceedings (OSTI)

The EPRI Utility Requirements Document (URD) was developed and last revised in 1999 to provide a list of requirements for the design and construction of new nuclear power plants. The objective of this project was to review URD Vol. III. This volume covers passive advanced light water reactors (ALWRs) for plant design requirements with respect to operations and maintenance (O&M) practices of the plant's cooling water systems (not including the circulating water system used for condenser cooling). The revi...

2007-07-26T23:59:59.000Z

7

Sustained Recycle in Light Water and Sodium-Cooled Reactors  

Science Conference Proceedings (OSTI)

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

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

2010-11-01T23:59:59.000Z

8

Passive decay heat removal system for water-cooled nuclear reactors  

DOE Patents (OSTI)

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.

Forsberg, Charles W. (Oak Ridge, TN)

1991-01-01T23:59:59.000Z

9

THE DETECTION OF BOILING IN A WATER-COOLED NUCLEAR REACTOR  

SciTech Connect

Measurements made at ORNL to study the feasibility of boiling detection in a water-cooled nuclear reactor are described. The methods selected for the detection of boiling include measurement of the acoustical noise produced by the generation of bubbles and measurement of changes in the reactor-power spectral density produced by bubbles. Preliminary results indicating that both methods could detect boiling are shown. (auth)

Colomb, A.L.; Binford, F.T.

1962-08-17T23:59:59.000Z

10

RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors  

SciTech Connect

The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

2003-04-01T23:59:59.000Z

11

Cooling water distribution system  

DOE Patents (OSTI)

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.

Orr, Richard (Pittsburgh, PA)

1994-01-01T23:59:59.000Z

12

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

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

13

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

DOE Patents (OSTI)

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

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

1993-12-14T23:59:59.000Z

14

Passive decay heat removal system for water-cooled nuclear reactors  

DOE Patents (OSTI)

This document describes passive decay-heat removal system for a water-cooled nuclear reactor which employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated evaporator 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.

Forseberg, C.W.

1990-01-01T23:59:59.000Z

15

Passive decay heat removal system for water-cooled nuclear reactors  

DOE Patents (OSTI)

This document describes passive decay-heat removal system for a water-cooled nuclear reactor which employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated evaporator 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.

Forseberg, C.W.

1990-12-31T23:59:59.000Z

16

Reactor water cleanup system  

DOE Patents (OSTI)

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

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

1994-01-01T23:59:59.000Z

17

Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor  

Science Conference Proceedings (OSTI)

Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to {approx}3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged.

Zhang Zuoyi [Tsinghua University (China); Dong Yujie [Tsinghua University (China); Scherer, Winfried [Forschungszentrum Juelich (Germany)

2005-03-15T23:59:59.000Z

18

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

DOE Patents (OSTI)

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.

Hill, P.R.

1994-12-27T23:59:59.000Z

19

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

DOE Patents (OSTI)

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.

Hill, Paul R. (Tucson, AZ)

1994-01-01T23:59:59.000Z

20

Gas-cooled reactors  

SciTech Connect

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing.

Schulten, R.; Trauger, D.B.

1976-01-01T23:59:59.000Z

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


21

Reactor core isolation cooling system  

DOE Patents (OSTI)

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.

Cooke, F.E.

1992-12-08T23:59:59.000Z

22

Reactor core isolation cooling system  

DOE Patents (OSTI)

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.

Cooke, Franklin E. (San Jose, CA)

1992-01-01T23:59:59.000Z

23

AIR COOLED NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

Fermi, E.; Szilard, L.

1958-05-27T23:59:59.000Z

24

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production  

Science Conference Proceedings (OSTI)

The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

2005-02-13T23:59:59.000Z

25

Critical Design Issues of Tokamak Cooling Water System of ITER's Fusion Reactor  

SciTech Connect

U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). The TCWS transfers heat generated in the Tokamak to cooling water during nominal pulsed operation 850 MW at up to 150 C and 4.2 MPa water pressure. This water contains radionuclides because impurities (e.g., tritium) diffuse from in-vessel components and the vacuum vessel by water baking at 200 240 C at up to 4.4MPa, and corrosion products become activated by neutron bombardment. The system is designated as safety important class (SIC) and will be fabricated to comply with the French Order concerning nuclear pressure equipment (December 2005) and the EU Pressure Equipment Directive using ASME Section VIII, Div 2 design codes. The complexity of the TCWS design and fabrication presents unique challenges. Conceptual design of this one-of-a-kind cooling system has been completed with several issues that need to be resolved to move to next stage of the design. Those issues include flow balancing between over hundreds of branch pipelines in parallel to supply cooling water to blankets, determination of optimum flow velocity while minimizing the potential for cavitation damage, design for freezing protection for cooling water flowing through cryostat (freezing) environment, requirements for high-energy piping design, and electromagnetic impact to piping and components. Although the TCWS consists of standard commercial components such as piping with valves and fittings, heat exchangers, and pumps, complex requirements present interesting design challenges. This paper presents a brief description of TCWS conceptual design and critical design issues that need to be resolved.

Kim, Seokho H [ORNL; Berry, Jan [ORNL

2011-01-01T23:59:59.000Z

26

GAS COOLED NUCLEAR REACTORS  

DOE Patents (OSTI)

A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

Long, E.; Rodwell, W.

1958-06-10T23:59:59.000Z

27

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

SciTech Connect

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

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

2011-01-01T23:59:59.000Z

28

Gas-cooled nuclear reactor  

DOE Patents (OSTI)

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

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

1985-01-01T23:59:59.000Z

29

Solar energy as an alternate energy source to mixed oxide fuels in light-water cooled reactors  

DOE Green Energy (OSTI)

Supplemental information pertaining to the generic environmental impact statement on the Pu recycling process for mixed oxide light-water cooled reactors (GESMO) was requested from several sources. In particular, the role of alternate sources of energy was to be explored and the implications of these alternate sources to the question of Pu recycle in LWRs were to be investigated. In this vein, solar energy as an alternate source is the main subject of this report, along with other information related to solar energy. The general conclusion is that solar energy should have little effect on the decisions concerning GESMO.

Bertini, H.W.

1977-06-30T23:59:59.000Z

30

Operating experience of natural circulation core cooling in boiling water reactors  

SciTech Connect

General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed.

Kullberg, C.; Jones, K.; Heath, C.

1993-08-01T23:59:59.000Z

31

Liquid metal cooled nuclear reactors with passive cooling system  

SciTech Connect

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.

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

1991-01-01T23:59:59.000Z

32

THE DEVELOPMENT OF COMPOSITE CONTROL RODS FOR WATER-COOLED POWER REACTORS  

SciTech Connect

The phrase "composite control rod" is used to describe a hafnium-tipped titanium-boron control component with a titanium cladding. Blades for such cortrol rods were successfully prepared in cooperation with the Battelle Memorial Institute by a picture-frame rolling technique. The rolling packs, which are machined from type 304 stainless steel, contain slntered titanium boron and wrought hafnium core materials in a commercially pure titanium envelope. Such packs are evacuated, sealod off, and rolled at 16O0 F with a total reduction of 3/1 using 20% reduction per roll setting. Postfabrication treatments include mechanical removal of the stainless steel envelope, flat annealing, machining, and stress relief annealing. Data on the mechanical properties, corrosion performance, thermal cycling resistance, and irradiation damage resistance of composite control rod components are presented. This information strongly indicates that composite control rods will perform satisfactorily in water-coolod reactors. (aut)h

Ray, W.E.

1957-10-15T23:59:59.000Z

33

Cooling Water System Optimization  

E-Print Network (OSTI)

During summer months, many manufacturing plants have to cut back in rates because the cooling water system is not providing sufficient cooling to support higher production rates. There are many low/no-cost techniques available to improve tower performance. To understand the importance of the optimization techniques, cooling tower theory will be discussed first.

Aegerter, R.

2005-01-01T23:59:59.000Z

34

Development of Modeling Capabilities for the Analysis of Supercritical Water-Cooled Reactor Thermal-Hydraulics and Dynamics  

Science Conference Proceedings (OSTI)

Develop an experimental and theoretical data base for heat transfer in tubes and channels cooled by water and CO2 at supercritical pressures.

Dr. Michael Z. Podowski

2009-04-16T23:59:59.000Z

35

Passive containment cooling water distribution device  

DOE Patents (OSTI)

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.

Conway, Lawrence E. (Hookstown, PA); Fanto, Susan V. (Plum Borough, PA)

1994-01-01T23:59:59.000Z

36

Development of flaw evaluation and acceptance procedures for flaw indications in the cooling water system at the Savannah River Site K Reactor  

SciTech Connect

This paper describes the methodology used in determining the criteria for acceptance of inspection indications in the K-Reactor Cooling Water System at the Savannah River Plant. These criteria have been developed in a manner consistent with the development of similar criteria in the ASME Code Section 11 for commercial light water reactors, but with a realistic treatment of the operating conditions in the cooling water system. The technical basis for the development of these criteria called {open_quotes}Acceptance Standards{close_quotes} is contained in this paper. A second portion of this paper contains the methodology used in the construction of flaw evaluation charts which have been developed for each specific line size in the cooling water system. The charts provide the results of detailed fracture mechanics calculations which have been completed to determine the largest flaw which can be accepted in the cooling water system without repair. These charts are designed for use in conjunction with inservice inspections of the cooling water system, and only require inspection results to determine acceptability.

Tandon, S.; Bamford, W.H. [Westinghouse Electric Corp., Pittsburgh, PA (US); Cowfer, C.D.; Ostrowski, R. [Westinghouse Savannah River Co., Aiken, SC (US)

1993-06-01T23:59:59.000Z

37

Development of flaw evaluation and acceptance procedures for flaw indications in the cooling water system at the Savannah River Site K Reactor  

SciTech Connect

This paper describes the methodology used in determining the criteria for acceptance of inspection indications in the K-Reactor Cooling Water System at the Savannah River Plant. These criteria have been developed in a manner consistent with the development of similar criteria in the ASME Code Section 11 for commercial light water reactors, but with a realistic treatment of the operating conditions in the cooling water system. The technical basis for the development of these criteria called [open quotes]Acceptance Standards[close quotes] is contained in this paper. A second portion of this paper contains the methodology used in the construction of flaw evaluation charts which have been developed for each specific line size in the cooling water system. The charts provide the results of detailed fracture mechanics calculations which have been completed to determine the largest flaw which can be accepted in the cooling water system without repair. These charts are designed for use in conjunction with inservice inspections of the cooling water system, and only require inspection results to determine acceptability.

Tandon, S.; Bamford, W.H. (Westinghouse Electric Corp., Pittsburgh, PA (United States)); Cowfer, C.D.; Ostrowski, R. (Westinghouse Savannah River Co., Aiken, SC (United States))

1993-01-01T23:59:59.000Z

38

Optimization of Cooling Water  

E-Print Network (OSTI)

A cooling water system can be optimized by operation at the highest possible cycles of concentration without risking sealing and fouling on heat exchanger surfaces. The way to optimize will be shown, with a number of examples of new systems.

Matson, J.

1985-05-01T23:59:59.000Z

39

Passive cooling safety system for liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

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

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

1991-01-01T23:59:59.000Z

40

Indirect passive cooling system for liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

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

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

1990-01-01T23:59:59.000Z

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


41

Water Cooling | Open Energy Information  

Open Energy Info (EERE)

Cooling: Cooling: Water cooling is commonly defined as a method of using water as a heat conduction to remove heat from an object, machine, or other substance by passing cold water over or through it. In energy generation, water cooling is typically used to cool steam back into water so it can be used again in the generation process. Other definitions:Wikipedia Reegle Water Cooling Typical water cooled condenser used for condensing steam Water or liquid cooling is the most efficient cooling method and requires the smallest footprint when cold water is readily available. When used in power generation the steam/vapor that exits the turbine is condensed back into water and reused by means of a heat exchanger. Water cooling requires a water resource that is cold enough to bring steam, typically

42

Water cooled steam jet  

DOE Patents (OSTI)

A water cooled steam jet for transferring fluid and preventing vapor lock, or vaporization of the fluid being transferred, has a venturi nozzle and a cooling jacket. The venturi nozzle produces a high velocity flow which creates a vacuum to draw fluid from a source of fluid. The venturi nozzle has a converging section connected to a source of steam, a diffuser section attached to an outlet and a throat portion disposed therebetween. The cooling jacket surrounds the venturi nozzle and a suction tube through which the fluid is being drawn into the venturi nozzle. Coolant flows through the cooling jacket. The cooling jacket dissipates heat generated by the venturi nozzle to prevent vapor lock.

Wagner, Jr., Edward P. (Idaho Falls, ID)

1999-01-01T23:59:59.000Z

43

Liquid metal cooled nuclear reactor plant system  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

44

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

SciTech Connect

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

Philip E. MacDonald

2003-09-01T23:59:59.000Z

45

WATER COOLED RETORT COVER  

DOE Patents (OSTI)

A retort cover is designed for use in the production of magnesium metal by the condensation of vaporized metal on a collecting surface. The cover includes a condensing surface, insulating means adjacent to the condensing surface, ind a water-cooled means for the insulating means. The irrangement of insulation and the cooling means permits the magnesium to be condensed at a high temperature and in massive nonpyrophoric form. (AEC)

Ash, W.J.; Pozzi, J.F.

1962-05-01T23:59:59.000Z

46

Development of Mechanistic Modeling Capabilities for Local Neutronically-Coupled Flow-Induced Instabilities in Advanced Water-Cooled Reactors  

SciTech Connect

The major research objectives of this project included the formulation of flow and heat transfer modeling framework for the analysis of flow-induced instabilities in advanced light water nuclear reactors such as boiling water reactors. General multifield model of two-phase flow, including the necessary closure laws. Development of neurton kinetics models compatible with the proposed models of heated channel dynamics. Formulation and encoding of complete coupled neutronics/thermal-hydraulics models for the analysis of spatially-dependent local core instabilities. Computer simulations aimed at testing and validating the new models of reactor dynamics.

Michael Podowski

2009-11-30T23:59:59.000Z

47

WATER BOILER REACTOR  

DOE Patents (OSTI)

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.

King, L.D.P.

1960-11-22T23:59:59.000Z

48

Passive cooling system for top entry liquid metal cooled nuclear reactors  

SciTech Connect

This patent describes a passive cooling system for liquid metal cooled, top entry loop nuclear fission reactors. It comprises: a liquid metal cooled nuclear reactor plant; a passive cooling system; and a secondary passive cooling system.

Boardman, C.E.; Hunsbedt, A.; Hui, M.M.

1992-10-27T23:59:59.000Z

49

HEPTAFLUOROPROPANE WITH WATER SPRAY COOLING ...  

Science Conference Proceedings (OSTI)

HEPTAFLUOROPROPANE WITH WATER SPRAY COOLING SYSTEM AS A TOTAL ... and evaluation studies on active and passive fire protection ...

2011-10-13T23:59:59.000Z

50

Open Cooling Water Chemistry Guideline  

Science Conference Proceedings (OSTI)

State-of-the-art chemistry programs help to ensure the continued operation of open cooling water systems while mitigating corrosion and fouling mechanisms. This document, Open Cooling Water Chemistry Guideline, prepared by a committee of industry experts, reflects field and laboratory data on corrosion and fouling issues of open cooling systems.BackgroundService Water System Chemical Addition Guideline (Electric Power Research Institute ...

2012-09-17T23:59:59.000Z

51

Cooling system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1982-01-01T23:59:59.000Z

52

Dry cooling tower operating experience in the LOFT reactor  

SciTech Connect

A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features.

Hunter, J.A.

1980-01-01T23:59:59.000Z

53

Indirect passive cooling system for liquid metal cooled nuclear reactors  

SciTech Connect

This patent describes a passive cooling system. It is for liquid metal cooled nuclear reactors having a pool of liquid metal coolant with the heat generating fissionable fuel core substantially immersed in the pool of liquid metal coolant. The passive cooling system including a combination of spaced apart side-by-side partitions in generally concentric arrangement and providing for intermediate fluid circulation and heat transfer therebetween.

Hunsbedt, A.; Boardman, C.E.

1990-09-25T23:59:59.000Z

54

EVALUATION OF MERCURY COOLED BREEDER REACTORS  

SciTech Connect

A technical and economic evaluation of a mercury-cooled fast breeder reactor is presented. The objectives of the program were to establish the technical feasibility of a fast breeder reactor cooled with boiling mercury and to evaluate the long-range potential of such a reactor power plant for production of economic power. Details of the conceptual design of a 100-Mw(e) reactor and system are discussed. The power cost from a mercury cooled fast breeder reactor was estimated as 21.4 mills/kwh which is competitive with the power cost for the initial Enrico Fermi plant. It was concluded that this reactor concept is technically feasible and has promising long-range economic potential. (M.C.G.)

Battles, D.W.

1960-12-14T23:59:59.000Z

55

IEP - Water-Energy Interface: Cooling Water Intake Structures  

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

The types of cooling water systems to be evaluated are: Wet Cooling Tower - The condenser is cooled with water recirculated to a mechanical draft cooling tower. Because there...

56

Reactor physics design of supercritical CO?-cooled fast reactors  

E-Print Network (OSTI)

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

Pope, Michael A. (Michael Alexander)

2004-01-01T23:59:59.000Z

57

Gas-Cooled Reactors: the importance of their development  

SciTech Connect

Gas-Cooled Reactors are considered to have a significant future impact on the application of fission energy. The specific types are the steam-cycle High-Temperature Gas-Cooled Reactor, the Gas-Cooled Fast Breeder Reactor, the gas-turbine HTGR, and the Very High-Temperature Process Heat Reactor. The importance of developing the above systems is discussed relative to alternative fission power systems involving Light Water Reactors, Heavy Water Reactors, Spectral Shift Controlled Reactors, and Liquid-Metal-Cooled Fast Breeder Reactors. A primary advantage of developing GCRs as a class lies in the technology and cost interrelations, permitting cost-effective development of systems having diverse applications. Further, HTGR-type systems have highly proliferation-resistant characteristics and very attractive safety features. Finally, such systems and GCFRs are mutally complementary. Overall, GCRs provide interrelated systems that serve different purposes and needs; their development can proceed in stages that provide early benefits while contributing to future needs. It is concluded that the long-term importance of the various GCRs is as follows: HTGR, providing a technology for economic GCFRs and HTGR-GTs, while providing a proliferation-resistant reactor system having early economic and fuel utilization benefits; GCFR, providing relatively low cost fissile fuel and reducing overall separative work needs at capital costs lower than those for LMFBRs; HTGR-GT (in combination with a bottoming cycle), providing a very high thermal efficiency system having low capital costs and improved fuel utilization and technology pertinent to VHTRs; HTGR-GT, providing a power system well suited for dry cooling conditions for low-temperature process heat needs; and VHTR, providing a high-temperature heat source for hydrogen production processes.

Kasten, P.R.

1978-11-16T23:59:59.000Z

58

Flow Stability of Supercritical Water Cooled Systems  

SciTech Connect

Research activities are ongoing worldwide to develop nuclear power plants with supercritical water cooled reactor (SCWR) with the purpose to achieve a high thermal efficiency and to improve their economical competitiveness. However, the strong variation of the thermal-physical properties of water in the vicinity of the pseudo-critical line results in challenging tasks in thermal-hydraulic design of a SCWR. One of the challenging tasks is to understand and to predict the dynamic behavior and flow stability of supercritical water cooled systems. Although extensive thermal-hydraulic research activities have been carried out worldwide, studies on flow stability of SC water cooled systems are scarce. The present study deals with the flow behavior of SC water cooled systems. For this purpose the computer code SASC was developed, which is applied to a simplified cooling system. The effect of various parameters on the flow behavior is investigated. The first results achieved up to now reveals a complicated dynamic performance of a system cooled by supercritical water. (authors)

Cheng, X.; Kuang, B.; Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 1954 Hua Shan Road, Shanghai 200030 (China)

2006-07-01T23:59:59.000Z

59

Nuclear reactor cooling system decontamination reagent regeneration  

DOE Patents (OSTI)

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

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

1985-01-01T23:59:59.000Z

60

HEAVY WATER MODERATED NEUTRONIC REACTOR  

DOE Patents (OSTI)

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.

Szilard, L.

1958-04-29T23:59:59.000Z

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


61

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

DOE Patents (OSTI)

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

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

1991-01-01T23:59:59.000Z

62

Conceptual design for a re-entrant type fuel channel for supercritical water-cooled nuclear reactors.  

E-Print Network (OSTI)

??Current CANDU-type nuclear reactors use a once-through fuel-channel with an annulus gas insulating it from the moderator. The current reference design for a CANDU-type SuperCritical… (more)

Samuel, Jeffrey

2011-01-01T23:59:59.000Z

63

Definition: Water Cooling | Open Energy Information  

Open Energy Info (EERE)

Water Cooling Water Cooling Water cooling is commonly defined as a method of using water as a heat conduction to remove heat from an object, machine, or other substance by passing cold water over or through it. In energy generation, water cooling is typically used to cool steam back into water so it can be used again in the generation process.[1] View on Wikipedia Wikipedia Definition Water cooling is a method of heat removal from components and industrial equipment. As opposed to air cooling, water is used as the heat conductor. Water cooling is commonly used for cooling automobile internal combustion engines and large industrial facilities such as steam electric power plants, hydroelectric generators, petroleum refineries and chemical plants. Other uses include cooling the barrels of machine guns, cooling of

64

"Hot" for Warm Water Cooling  

E-Print Network (OSTI)

liquid cooling, dry cooler, cooling tower 1. INTRODUCTIONsolutions for cooling. Substituting cooling towers,hybrid cooling towers, or dry coolers that provide warmer

Coles, Henry

2012-01-01T23:59:59.000Z

65

Boiling water neutronic reactor incorporating a process inherent safety design  

DOE Patents (OSTI)

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.

Forsberg, C.W.

1985-02-19T23:59:59.000Z

66

Light Water Reactors Technology Development - Nuclear Reactors  

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

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

67

Boiling water neutronic reactor incorporating a process inherent safety design  

DOE Patents (OSTI)

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.

Forsberg, Charles W. (Kingston, TN)

1987-01-01T23:59:59.000Z

68

Thermal and flow design of helium-cooled reactors  

Science Conference Proceedings (OSTI)

This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors.

Melese, G.; Katz, R.

1984-01-01T23:59:59.000Z

69

Water Cooling Systems  

Science Conference Proceedings (OSTI)

...coil. Usually, two pumps are provided, one as a standby with an automatic switchover, because it is extremely important to circulate water through the coil continuously to prevent it from being damaged. To protect induction furnace equipment during momentary power interruptions or prolonged power...

70

GAS COOLED NUCLEAR REACTOR STUDY. Final Report  

SciTech Connect

An investigntion was made of the performance of a gas-cooled reactor, designed to provide a source of high temperature heat to a stream of helium. This reactor, in turn, is used as a source of heat for the air stream in a gas- turbine power plant. The reactor design was predicted primarily on the requirement for transferring a large amount of heat to the helium stream with a pressure drop low enough that it will not represent a major loss of power in the power plant. The mass of uranium e uired far criticality under various circumstances was investigated by multigroup calculations, both on desk calculators and on an IBM-704 machine. The gasturbine power plant perfarmance was studied based on a Studebaker-Packard-designed gas-turbine power plant for the propulsion of destroyer-escort vessels. A small experimental program was carried out to study some effects of helium on graphite and on structural steels. (auth)

Thompson, A.S.

1956-07-31T23:59:59.000Z

71

STEAM-COOLED POWER REACTOR EVALUATION, STEAM-COOLED FAST BREEDER REACTOR  

SciTech Connect

Conceptual design and economic studies of a steamcooled fast breeder reactor that can also be used as a source of power are presented. Two reactor plant sizes were considered: a 300-Mw(e) central power station plant and a 40 Mw(e) plant. It was concluded that attractive economics and good breeding characteristics breeding ratios from 1.27 to 1.42) can be achieved in steam- cooled PuO/sub 2/UO/sub 2/ fueled fast reactors. Low capital costs can be obtained by a compact reactor core and the absence of large heat exchangers and complicated process systems. Reactor design data are discussed. Analysis showed that these reactors can be prevented from going prompt critical, when fully flooded, by incorporating a tolerable amount of high resonance absorption materials such as hafnium or indium. An increase in reactivity on loss of coolant was indicated by preliminary calculations. (M.C.G.)

Sofer, G.; Hankel, R.; Goldstein, L.; Birman, G.

1961-04-15T23:59:59.000Z

72

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

DOE Patents (OSTI)

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.

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

1993-01-01T23:59:59.000Z

73

Integrated reactor-containment hyperbolic-cooling-tower system  

Science Conference Proceedings (OSTI)

A preliminary feasibility analysis has been conducted to evaluate placing a nuclear reactor containment building inside a large hyperbolic cooling tower, a concept previously suggested for fossil-fired units but for reasons other than those that motivate this evaluation. The geometry of the design, the amount of water available, and the shielding provided by the cooling tower are beneficial to the safety characteristics of the containment under accident conditions. Three means of decay heat management are employed: an initial water spray on the containment exterior, long-term air convection on side of the containment, and creation of a water pool inside the containment. A continuously spraying water tank on top of the containment allows for a completely passive decay heat removal system. An annular air chimney around the containment is effective in long-term removal of {approximately} 1O MW (thermal) through air convection. Five percent of the water inventory in the cooling-tower pond surrounding the containment is sufficient to flood the containment interior to a depth of 14.6 ft, thereby providing an internal containment heat sink. The packing and the height of the tower provide major scrubbing and dispersing sources for any uncontrolled radioactive leak. The cooling tower veil also protects the containment from external events such as lane crashes.

Patel, A.R.; Todreas, N.E.; Driscoll, M.J. [Massachusetts Institute of Technology, Cambridge, MA (United States)

1994-12-31T23:59:59.000Z

74

Lessons Learned From Gen I Carbon Dioxide Cooled Reactors  

Science Conference Proceedings (OSTI)

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

David E. Shropshire

2004-04-01T23:59:59.000Z

75

Passive cooling system for top entry liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

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.

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

1992-01-01T23:59:59.000Z

76

Monitoring system for a liquid-cooled nuclear fission reactor  

SciTech Connect

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

DeVolpi, Alexander (Bolingbrook, IL)

1987-01-01T23:59:59.000Z

77

GAS COOLED POWER REACTOR COOLANT CHOICE  

SciTech Connect

The current status of helium and carbon dioxide technology is described in the light of the Gas Cooled Reactor Program requiremoents. The problem of containing high-pressure helium at high temperature is discussed, and it is concluded that, by proper attention to the design, construction and maintenance of a plant, a high degree of helium leak-tightness can be achieved at small additional cost when compared with a carbon dioxide system. What is more, the cost of making up helium losses in a practically achievable system is estimated to be small compared with other fixed and operating costs. Graphite-carbon dioxide reaction data are reviewed. It is shown that carbon dioxide at atmospheric pressure and low flow rates should be compatible with a graphite mooderator up to 525 C. No data are available at the high pressures and fiow rates that would be encountered in power reactors. Significantiy higher oxidation rates may result, however, perhaps limiting bulk moderator temperatures to 450 to 500 C. Improved carbon materials, protective coatings and inhibitors, and/or operating practices may be developed that will allow significant future increases in these limiting temperatures. (auth)

Heacock, H.W.; Nightingale, R.E.

1958-06-12T23:59:59.000Z

78

Reactor Pressure Vessel Task of Light Water Reactor Sustainability...  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure...

79

Validation of the RVACS (Reactor Vessel Auxiliary Cooling System)/RACS (Reactor Air Cooling System) model in SASSYS-1  

SciTech Connect

The SASSYS-1 LMR systems analysis code contains a model for transient analysis of heat removal by a RVACS (Reactor Vessel Auxiliary Cooling System) or a RACS (Reactor Air Cooling System) in an LMR (Liquid Metal Reactor). This model has been validated by comparisons of model predictions with experimental data from a large scale RVACS/RACS simulation experiment performed at Argonne National Laboratory. 4 refs., 1 fig.

Dunn, F.E.

1987-01-01T23:59:59.000Z

80

Design Considerations for Economically Competitive Sodium Cooled Fast Reactors  

SciTech Connect

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

Hongbin Zhang; Haihua Zhao

2009-05-01T23:59:59.000Z

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


81

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

SciTech Connect

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

WRIGHT,STEVEN A.; HOUTS,MICHAEL

2000-11-22T23:59:59.000Z

82

"Hot" for Warm Water Cooling  

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

Published 112011 Conference Location Seattle, WA Call Number LBNL-5128E Abstract Liquid cooling is key to reducing energy consumption for this generation of supercomputers and...

83

Natural circulating passive cooling system for nuclear reactor containment structure  

DOE Patents (OSTI)

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

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

1990-01-01T23:59:59.000Z

84

Passive cooling system for nuclear reactor containment structure  

DOE Patents (OSTI)

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

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

1989-01-01T23:59:59.000Z

85

CHF Enhancement by Vessel Coating for External Reactor Vessel Cooling  

SciTech Connect

In-vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding of the reactor cavity during a severe accident. As part of a joint Korean – United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boiling experiments were performed in the SBLB (Subscale Boundary Layer Boiling) facility at Penn State using test vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady state boiling, the vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the vessel outer surface but with very little effect of aging.

Fan-Bill Cheung; Joy L. Rempe

2004-06-01T23:59:59.000Z

86

Closed Cooling Water Chemistry Guideline: Revision 2  

Science Conference Proceedings (OSTI)

This Closed Cooling Water Chemistry Guideline addresses the use of chemicals and monitoring methods to mitigate corrosion, fouling, and microbiological growth in the closed cooling-water (CCW) systems of nuclear and fossil-fueled power plants. The chemical additives used for these purposes depend on plant-design characteristics, water quality, operating parameters, and the specifications of the Nuclear Steam Supply System (NSSS) suppliers. The list of chemicals is not as extensive as that ...

2013-12-09T23:59:59.000Z

87

Conceptual design description for the tritium recovery system for the US ITER (International Thermonuclear Experimental Reactor) Li sub 2 O/Be water cooled blanket  

Science Conference Proceedings (OSTI)

The tritium recovery system for the US ITER Li{sub 2}O/Be water cooled blanket processes two separate helium purge streams to recover tritium from the Li{sub 2}O zones and the Be zones of the blanket, to process the waste products, and to recirculate the helium back to the blanket. The components are selected to minimize the tritium inventory of the recovery system, and to minimize waste products. The system is robust to either an increase in the tritium release rate or to an in-leak of water in the purge system. Three major components were used to process these streams, first, 5A molecular sieves at {minus}196{degree}C separate hydrogen from the helium, second, a solid oxide electrolysis unit is used to reduce all molecular water, and third, a palladium/silver diffuser is used to ensure that only hydrogen (H{sub 2}, HT) species reach the cryogenic distillation unit. Other units are present to recover tritium from waste products but the three major components are the basis of the blanket tritium recovery system. 32 refs.

Finn, P.A.; Sze, D.K. (Argonne National Lab., IL (USA). Fusion Power Program); Clemmer, R.G. (Pacific Northwest Lab., Richland, WA (USA))

1990-11-01T23:59:59.000Z

88

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report  

SciTech Connect

The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks: Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking Task 3. Plant Engineering and Reactor Safety Analysis. moderator rods. materials.

Philip E. MacDonald

2005-01-01T23:59:59.000Z

89

"Hot" for Warm Water Cooling  

Science Conference Proceedings (OSTI)

Liquid cooling is key to reducing energy consumption for this generation of supercomputers and remains on the roadmap for the foreseeable future. This is because the heat capacity of liquids is orders of magnitude larger than that of air and once heat has been transferred to a liquid, it can be removed from the datacenter efficiently. The transition from air to liquid cooling is an inflection point providing an opportunity to work collectively to set guidelines for facilitating the energy efficiency of liquid-cooled High Performance Computing (HPC) facilities and systems. The vision is to use non-compressor-based cooling, to facilitate heat re-use, and thereby build solutions that are more energy-efficient, less carbon intensive and more cost effective than their air-cooled predecessors. The Energy Efficient HPC Working Group is developing guidelines for warmer liquid-cooling temperatures in order to standardize facility and HPC equipment, and provide more opportunity for reuse of waste heat. This report describes the development of those guidelines.

IBM Corporation; Energy Efficient HPC Working Group; Hewlett Packard Corporation; SGI; Cray Inc.; Intel Corporation; U.S. Army Engineer Research Development Center; Coles, Henry; Ellsworth, Michael; Martinez, David J.; Bailey, Anna-Maria; Banisadr, Farhad; Bates, Natalie; Coghlan, Susan; Cowley, David E.; Dube, Nicholas; Fields, Parks; Greenberg, Steve; Iyengar, Madhusudan; Kulesza, Peter R.; Loncaric, Josip; McCann, Tim; Pautsch, Greg; Patterson, Michael K.; Rivera, Richard G.; Rottman, Greg K.; Sartor, Dale; Tschudi, William; Vinson, Wade; Wescott, Ralph

2011-08-26T23:59:59.000Z

90

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

E-Print Network (OSTI)

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

Buongiorno, J.

91

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

E-Print Network (OSTI)

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

Buongiorno, Jacopo, 1971-

2001-01-01T23:59:59.000Z

92

Light Water Reactor Sustainability Program: Integrated Program...  

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

Light Water Reactor Sustainability Program: Integrated Program Plan Light Water Reactor Sustainability Program: Integrated Program Plan Nuclear power has safely, reliably, and...

93

Modeling and performance of the MHTGR (Modular High-Temperature Gas-Cooled Reactor) reactor cavity cooling system  

SciTech Connect

The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab.

Conklin, J.C. (Oak Ridge National Lab., TN (USA))

1990-04-01T23:59:59.000Z

94

New and Underutilized Technology: Water Cooled Oil Free Magnetic...  

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

Water Cooled Oil Free Magnetic Bearing Compressors New and Underutilized Technology: Water Cooled Oil Free Magnetic Bearing Compressors October 4, 2013 - 3:58pm Addthis The...

95

A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA  

E-Print Network (OSTI)

A Review of Light-Water Reactor Safety Studies," by A.V.due to a break in the reactor cooling cooling water the therecirculation - Failure of the reactor protection system.

Nero, A.V.

2010-01-01T23:59:59.000Z

96

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

DOE Patents (OSTI)

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

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

1980-06-06T23:59:59.000Z

97

Gravity Scaling of a Power Reactor Water Shield  

Science Conference Proceedings (OSTI)

Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson

Robert S. Reid; J. Boise Pearson

2008-01-01T23:59:59.000Z

98

Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors  

SciTech Connect

This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs.

Orvis, D.D.; Raabe, P.H.

1980-01-01T23:59:59.000Z

99

Liquid metal reactor air cooling baffle  

DOE Patents (OSTI)

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.

Hunsbedt, Anstein (Los Gatos, CA)

1994-01-01T23:59:59.000Z

100

Air and water cooled modulator  

DOE Patents (OSTI)

A compact high power magnetic compression apparatus and method are disclosed for delivering high voltage pulses of short duration at a high repetition rate and high peak power output which does not require the use of environmentally unacceptable fluids such as chlorofluorocarbons either as a dielectric or as a coolant, and which discharges very little waste heat into the surrounding air. A first magnetic switch has cooling channels formed therethrough to facilitate the removal of excess heat. The first magnetic switch is mounted on a printed circuit board. A pulse transformer comprised of a plurality of discrete electrically insulated and magnetically coupled units is also mounted on said printed board and is electrically coupled to the first magnetic switch. The pulse transformer also has cooling means attached thereto for removing heat from the pulse transformer. A second magnetic switch also having cooling means for removing excess heat is electrically coupled to the pulse transformer. Thus, the present invention is able to provide high voltage pulses of short duration at a high repetition rate and high peak power output without the use of environmentally unacceptable fluids and without discharging significant waste heat into the surrounding air. 9 figs.

Birx, D.L.; Arnold, P.A.; Ball, D.G.; Cook, E.G.

1995-09-05T23:59:59.000Z

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


101

Air and water cooled modulator  

DOE Patents (OSTI)

A compact high power magnetic compression apparatus and method for delivering high voltage pulses of short duration at a high repetition rate and high peak power output which does not require the use of environmentally unacceptable fluids such as chlorofluorocarbons either as a dielectric or as a coolant, and which discharges very little waste heat into the surrounding air. A first magnetic switch has cooling channels formed therethrough to facilitate the removal of excess heat. The first magnetic switch is mounted on a printed circuit board. A pulse transformer comprised of a plurality of discrete electrically insulated and magnetically coupled units is also mounted on said printed board and is electrically coupled to the first magnetic switch. The pulse transformer also has cooling means attached thereto for removing heat from the pulse transformer. A second magnetic switch also having cooling means for removing excess heat is electrically coupled to the pulse transformer. Thus, the present invention is able to provide high voltage pulses of short duration at a high repetition rate and high peak power output without the use of environmentally unacceptable fluids and without discharging significant waste heat into the surrounding air.

Birx, Daniel L. (Oakley, CA); Arnold, Phillip A. (Livermore, CA); Ball, Don G. (Livermore, CA); Cook, Edward G. (Livermore, CA)

1995-01-01T23:59:59.000Z

102

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

DOE Patents (OSTI)

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

Hunsbedt, A.; Boardman, C.E.

1995-04-11T23:59:59.000Z

103

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

SciTech Connect

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

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

1995-01-01T23:59:59.000Z

104

Water simulation of sodium reactors  

Science Conference Proceedings (OSTI)

The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system.

Grewal, S.S.; Gluekler, E.L.

1981-06-28T23:59:59.000Z

105

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

Science Conference Proceedings (OSTI)

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

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

106

Use of Thorium in Light Water Reactors  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Use of Alternate Fuels in Light Water Reactors

Michael Todosow; A. Galperin; S. Herring; M. Kazimi; T. Downar; A. Morozov

107

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

SciTech Connect

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

Not Available

1980-05-01T23:59:59.000Z

108

Containment system for supercritical water oxidation reactor  

DOE Patents (OSTI)

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

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

1994-01-01T23:59:59.000Z

109

Containment system for supercritical water oxidation reactor  

DOE Patents (OSTI)

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

Chastagner, P.

1991-12-31T23:59:59.000Z

110

Design and Operating Guidelines Manual for Cooling-Water Treatment - Treatment of Recirculated Cooling Water  

Science Conference Proceedings (OSTI)

This preliminary manual was developed to provide a systematic procedure for evaluating candidate strategies for the treatment of recirculated cooling water. It presents (1) a means of selecting optimal treatment methods and facilities on the basis of technical and economic considerations, and (2) guidelines for proper cooling-water system operation. Descriptions of, and user's manuals for, the cooling-system process and chemical equilibrium computer simulation models are included.

1982-03-01T23:59:59.000Z

111

Management of Non-Cooling Water Releases  

Science Conference Proceedings (OSTI)

This report assesses the efficacy and cost of water management practices that can be used by power companies to address non-cooling water and stormwater release issues with respect to siting, design, and operation of facilities, including generating stations, substations, and rights-of-way in urban and rural settings. The report will be of value to environmental and generation managers within power companies, as well as regulators, water resource managers, and environmentalists.

2008-03-03T23:59:59.000Z

112

Preparation of high temperature gas-cooled reactor fuel element  

DOE Patents (OSTI)

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.

Bradley, Ronnie A. (Oak Ridge, TN); Sease, John D. (Knoxville, TN)

1976-01-01T23:59:59.000Z

113

Gas-cooled reactors: the importance of their development  

SciTech Connect

The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U/sub 3/O/sub 8/ before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production.

Kasten, P.R.

1979-06-01T23:59:59.000Z

114

Fuel leak detection apparatus for gas cooled nuclear reactors  

SciTech Connect

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.

Burnette, Richard D. (San Diego, CA)

1977-01-01T23:59:59.000Z

115

Fuel Rod Cooling in Natural Uranium Reactors  

SciTech Connect

An analysis is presented of the transfer of heat from a cylindrical fuel rod surrounded by a fast flowing coolant in an annular duct, with maximum power output limited by fuel rod temperatures, coolant pressure drop and pumping power requirements. A method is also presented for comparing and evaluating various liquid and gaseous coolants within these limitations. The report also shows and discusses some calculated results obtained for the systems considred in the study of natural U reactors for the production of Pu and useful power (NAA-SR-137).

Trilling, C.A.

1952-01-28T23:59:59.000Z

116

Water reactor fuel cladding  

Science Conference Proceedings (OSTI)

This patent describes a nuclear reactor fuel element cladding tube. It comprises: an outer cylindrical layer of a first zirconium alloy selected from the group consisting of Zircaloy-2 and Zircaloy-4; an inner cylindrical layer of a second zirconium alloy consisting essentially of about 0.19 to 0.6 wt.% tin, about 0.19 to less than 0.5 wt.% iron, about 100 to 700 ppm oxygen, less than 2000 ppm total impurities, and the remainder essentially zirconium; the inner layer characterized by aqueous corrosion resistance substantially the same as the first zirconium alloy; the inner layer characterized by improved resistance to PCI crack propagation under reactor operating conditions compared to the first zirconium alloy and substantially the same PCI crack propagation resistance compared to unalloyed zirconium; and the inner cylindrical layer is metallurgically bonded to the outer layer.

Foster, J.P.; McDonald, S.G.

1990-06-12T23:59:59.000Z

117

Materials Degradation Issues in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

CASL: The Consortium for Advanced Simulation of Light Water Reactors: A U.S. ... Strategies for Studying High Dose Irradiation Effects in Reactor Components.

118

Survey of Optimization of Reactor Coolant Cleanup Systems: For Boiling Water Reactors and Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Optimization of the reactor coolant cleanup systems in the boiling water reactor (BWR) and pressurized water reactor (PWR) environment is important for controlling the transport of corrosion products (metals and activated metals), fission products, and coolant impurities (soluble and insoluble) throughout the reactor coolant loop, and this optimization contributes to reducing primary system radiation fields. The removal of radionuclides and corrosion products is just one of many functions (both ...

2013-09-27T23:59:59.000Z

119

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING  

Science Conference Proceedings (OSTI)

The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment including the dome was removed, a concrete cover was to be placed over the remaining footprint and the groundwater monitored for an indefinite period to ensure compliance with environmental regulations.

Austin, W.; Brinkley, D.

2011-10-13T23:59:59.000Z

120

CHIMNEY FOR BOILING WATER REACTOR  

DOE Patents (OSTI)

A boiling-water reactor is described which has vertical fuel-containing channels for forming steam from water. Risers above the channels increase the head of water radially outward, whereby water is moved upward through the channels with greater force. The risers are concentric and the radial width of the space between them is somewhat small. There is a relatively low rate of flow of water up through the radially outer fuel-containing channels, with which the space between the risers is in communication. (AE C)

Petrick, M.

1961-08-01T23:59:59.000Z

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


121

Repair and Replacement Applications Center: Stress Corrosion Cracking in Closed Cooling Water Systems  

Science Conference Proceedings (OSTI)

The results of a recent EPRI project "Stress Corrosion Cracking in PWR and BWR Closed Cooling Water Systems," (EPRI Report 1009721, October 2004) indicated that approximately 10 of 143 light water reactor (LWR) plants surveyed had through-wall leaks in carbon steel piping in their closed cooling water (CCW) systems. The root cause of this leakage was intergranular stress corrosion cracking. Since there has not been extensive non-destructive testing in these systems, it is likely that the incidence rate o...

2006-09-28T23:59:59.000Z

122

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

Science Conference Proceedings (OSTI)

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

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

2004-10-06T23:59:59.000Z

123

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

Science Conference Proceedings (OSTI)

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

Moisseytsev, A.; Sienicki, J.J.

2004-10-06T23:59:59.000Z

124

Irradiation Effects on High-Temperature Gas-Cooled Reactor Structural Materials  

Science Conference Proceedings (OSTI)

G. Irradiation Behavior / Status of Metallic Materials Development for Application in Advanced High-Temperature Gas-Cooled Reactor / Material

James R. Lindgren

125

Gas-cooled reactor for space power systems  

Science Conference Proceedings (OSTI)

Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors.

Walter, C.E.; Pearson, J.S.

1987-05-01T23:59:59.000Z

126

Water cooling of HVDC thyristor valves  

SciTech Connect

It is generally accepted that water is a very effective medium to remove heat losses from any type of equipment. When used for HVDC thyristor valves, the fundamentals of electrolyte conduction and water chemistry need to be considered in the design of the cooling circuit. The characteristics of the materials used, in conjunction with high voltage stresses and circuit configuration, play an important role to assure longevity and corrosion-free performance.

Lips, H.P. (Siemens AG, Erlangen (Germany))

1994-10-01T23:59:59.000Z

127

The sodium-cooled fast reactor (SFR).  

DOE Green Energy (OSTI)

The primary mission for the SFR is the management of high-level wastes, and in particular, management of plutonium and other actinides. The Generation IV Roadmap Fuel Cycle Crosscut Group (FCCG) found that the limiting factor facing an essential role for nuclear energy with the once-through cycle is the availability of repository space worldwide [FCCG Report]. This becomes an important issue, requiring new repository development in only a few decades. Systems that employ a fully closed fuel cycle hold the promise to reduce repository space and performance requirements, although their costs must be held to acceptable levels. Closed fuel cycles, working alone or symbiotically with systems using a once-through cycle, permit partitioning the nuclear waste and management of each partitioned fraction. In the longer term, beyond 50 years, or if major new missions requiring nuclear energy production (such as a major growth in the use of hydrogen as an energy carrier) develop, uranium resource availability also becomes a limiting factor unless breakthroughs occur in mining or extraction technologies. Fast spectrum reactors have the ability to utilize almost all of the energy in the natural uranium versus the 1% utilized in thermal spectrum systems.

Lineberry, M. J.; Allen, T. R.

2002-10-25T23:59:59.000Z

128

NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES  

DOE Patents (OSTI)

BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

McCorkle, W.H.

1960-02-23T23:59:59.000Z

129

Development of MELCOR Input Techniques for High Temperature Gas-Cooled Reactor Analysis  

E-Print Network (OSTI)

High Temperature Gas-cooled Reactors (HTGRs) can provide clean electricity,as well as process heat that can be used to produce hydrogen for transportation and other sectors. A prototypic HTGR, the Next Generation Nuclear Plant (NGNP),will be built at Idaho National Laboratory.The need for HTGR analysis tools and methods has led to the addition of gas-cooled reactor (GCR) capabilities to the light water reactor code MELCOR. MELCOR will be used by the Nuclear Regulatory Commission licensing of the NGNP and other HTGRs. In the present study, new input techniques have been developed for MELCOR HTGR analysis. These new techniques include methods for modeling radiation heat transfer between solid surfaces in an HTGR, calculating fuel and cladding geometric parameters for pebble bed and prismatic block-type HTGRs, and selecting appropriate input parameters for the reflector component in MELCOR. The above methods have been applied to input decks for a water-cooled reactor cavity cooling system (RCCS); the 400 MW Pebble Bed Modular Reactor (PBMR), the input for which is based on a code-to-code benchmark activity; and the High Temperature Test Facility (HTTF), which is currently in the design phase at Oregon State University. RCCS results show that MELCOR accurately predicts radiation heat transfer rates from the vessel but may overpredict convective heat transfer rates and RCCS coolant flow rates. PBMR results show that thermal striping from hot jets in the lower plenum during steady-state operations, and in the upper plenum during a pressurized loss of forced cooling accident, may be a major design concern. Hot jets could potentially melt control rod drive mechanisms or cause thermal stresses in plenum structures. For the HTTF, results will provide data to validate MELCOR for HTGR analyses. Validation will be accomplished by comparing results from the MELCOR representation of the HTTF to experimental results from the facility. The validation process can be automated using a modular code written in Python, which is described here.

Corson, James

2010-05-01T23:59:59.000Z

130

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from

131

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the operation of commercial nuclear power plants, require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including

132

THE EXPERIMENTAL BERYLLIUM OXIDE REACTOR. MARITIME GAS-COOLED REACTOR PROGRAM  

SciTech Connect

LIUM OXIDE REACTOR. MARITIME GAS-COOLED The Experimental Beryllium Oxide Reactor, EBOR, will be constructed at the National Reactor Testing Station as the AEC portion of the joint Maritime Administration--AEC Maritime Gas Cooled Reactor Program. The ultimate goal of the Program is the development of nuclear power plants employing a helium cooled and beryllium oxide moderated reactor directly coupled to a closed cycle gas turbine. The objective is to obtain compact nuclear engines suitable for use either in a merchant ship propulsion system or an intermediate size central station power plant in the 20 to 100 Mw(e) size range. The EBOR is a l0 Mw(t) test of the basic fuel element and moderator designs. It is capable of being up-graded in power at a later date to a test of the nuclear reactor turbine concept. The objective of the experiment is outlined. The principal reactor components to be tested and the test facility are described. (auth)

Moore, W.C.

1961-07-01T23:59:59.000Z

133

STEAM GENERATORS FOR HIGH-TEMPERATURE GAS-COOLED REACTORS  

SciTech Connect

An analytical approach and an IBM machine code were prepared for the design of gas-cooled reactor once-through steam generators for both axial-flow and cross-flow tube matrices. The codes were applied to investigate the effects of steam generator configuration, tube diameter, extended surface, type of cooling gas, steam and gas temperature and pressure conditions, and the pumping power-to-heat removal ratio on the size, weight, and cost of steam generators. The results indicate that the least expensive and most promising unit for high- temperature high-pressure gascooled reactor plants employs axial-gas flow over 0.5-in.dia bare U-tubes arranged with their axes parallel to that of the shell. The proposed design is readily adaptable to the installation of a reheater and is suited to conventional fabrication techniques. Charts are presented to facilitate tlie design of both axial-flow and cross-flow steam generators for gas- cooled reactor applications. (auth)

Fraas, A.P.; Ozisik, M.N.

1963-04-23T23:59:59.000Z

134

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

Science Conference Proceedings (OSTI)

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

Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

2012-06-01T23:59:59.000Z

135

Water inventory management in condenser pool of boiling water reactor  

DOE Patents (OSTI)

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

Gluntz, Douglas M. (San Jose, CA)

1996-01-01T23:59:59.000Z

136

Water inventory management in condenser pool of boiling water reactor  

DOE Patents (OSTI)

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

Gluntz, D.M.

1996-03-12T23:59:59.000Z

137

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS  

DOE Patents (OSTI)

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)

Furgerson, W.T.

1963-12-17T23:59:59.000Z

138

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

DOE Green Energy (OSTI)

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

Not Available

1980-10-09T23:59:59.000Z

139

lullIIlllIllLLLII DESIGN WINDOWS FOR A He COOLED FUSION REACTOR*  

E-Print Network (OSTI)

_ ii|l iImMmmm lullIIlllIllLLLII #12;. #12;DESIGN WINDOWS FOR A He COOLED FUSION REACTOR '....."[',":-,_.30 B93 _I_TFIII_31_ONOF THIS DO_.JMENT IS LJNLIMITED 0 S 1" I #12;,l° Design Windows for a He Cooled A design window concept is developed for a He-cooled of a helium cooled reactor are: fusion reactor blanket

Harilal, S. S.

140

Design guide for category VI reactors: air-cooled graphite reactors  

SciTech Connect

The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned air-cooled graphite reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC).

Brynda, W.J.; Karol, R.; Powell, R.W.

1979-02-01T23:59:59.000Z

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


141

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

SciTech Connect

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

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

2002-02-26T23:59:59.000Z

142

Best Management Practices Manual for Preventing Cooling Water Intake Blockages  

Science Conference Proceedings (OSTI)

Blockage of the cooling water intake structure (CWIS) occurs frequently at nuclear and fossil power facilities worldwide, regardless of fuel type or cooling water body source. The loss of cooling water impacts facility safety and reliabilityprincipally at nuclear facilitiesand results in a loss of revenue. This Best Management Practices Manual for Preventing Cooling Water Intake Blockages presents a review of debris management at existing facilities and provides procedural and operation and maintenance (...

2009-12-22T23:59:59.000Z

143

LIGHT WATER MODERATED NEUTRONIC REACTOR  

DOE Patents (OSTI)

A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

Christy, R.F.; Weinberg, A.M.

1957-09-17T23:59:59.000Z

144

Best Practice for Energy Efficient Cleanrooms: Cooling tower and condenser water optimization  

E-Print Network (OSTI)

Cleanrooms: Cooling Tower and Condenser Water OptimizationCleanrooms: Cooling tower and condenser water optimization2 Cooling tower and condenser water

Xu, Tengfang

2005-01-01T23:59:59.000Z

145

Best Practice for Energy Efficient Cleanrooms: Cooling tower and condenser water optimization  

E-Print Network (OSTI)

Efficient Cleanrooms: Cooling Tower and Condenser WaterEfficient Cleanrooms: Cooling tower and condenser water2 Cooling tower and condenser water

Xu, Tengfang

2005-01-01T23:59:59.000Z

146

Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.  

Science Conference Proceedings (OSTI)

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage, and cleaning stations-have accumulated satisfactory construction and operation experiences. In addition, two special issues for future development are described in this report: large capacity interim storage and transuranic-bearing fuel handling.

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

2009-03-01T23:59:59.000Z

147

Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors  

SciTech Connect

The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors.

1978-09-01T23:59:59.000Z

148

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

E-Print Network (OSTI)

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

Yarsky, Peter

2005-01-01T23:59:59.000Z

149

Developments in Molten Salt and Liquid-Salt-Cooled Reactors  

Science Conference Proceedings (OSTI)

In the last 5 years, there has been a rapid growth in interest in the use of high-temperature (700 to 1000 deg C) molten and liquid fluoride salts as coolants in nuclear systems. This renewed interest is a consequence of new applications for high-temperature heat and the development of new reactor concepts. Fluoride salts have melting points between 350 and 500 deg C; thus, they are of use only in high-temperature systems. Historically, steam cycles with temperature limits of {approx}550 deg C have been the only efficient method to convert heat to electricity. This limitation produced few incentives to develop high-temperature reactors for electricity production. However, recent advances in Brayton gas turbine technology now make it possible to convert higher-temperature heat efficiency into electricity on an industrial scale and thus have created the enabling technology for more efficient nuclear reactors. Simultaneously, there is a growing interest in using high-temperature nuclear heat for the production of hydrogen and shale oil. Five nuclear-related applications are being investigated: (1) liquid-salt heat-transport systems in hydrogen and shale oil production systems; (2) the advanced high-temperature reactor, which uses a graphite-matrix coated-particle fuel and a liquid salt coolant; (3) the liquid-salt-cooled fast reactor which uses metal-clad fuel and a liquid salt coolant; (4) the molten salt reactor, with the fuel dissolved in the molten salt coolant; and (5) fusion energy systems. The reasons for the new interest in liquid salt coolants, the reactor concepts, and the relevant programs are described. (author)

Forsberg, Charles W. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6165 (United States)

2006-07-01T23:59:59.000Z

150

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

Science Conference Proceedings (OSTI)

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

Scheele, Randall D.; Casella, Andrew M.

2010-09-28T23:59:59.000Z

151

Federal Energy Management Program: FEMP Designated Product: Water-Cooled  

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

FEMP Designated FEMP Designated Product: Water-Cooled Ice Machines to someone by E-mail Share Federal Energy Management Program: FEMP Designated Product: Water-Cooled Ice Machines on Facebook Tweet about Federal Energy Management Program: FEMP Designated Product: Water-Cooled Ice Machines on Twitter Bookmark Federal Energy Management Program: FEMP Designated Product: Water-Cooled Ice Machines on Google Bookmark Federal Energy Management Program: FEMP Designated Product: Water-Cooled Ice Machines on Delicious Rank Federal Energy Management Program: FEMP Designated Product: Water-Cooled Ice Machines on Digg Find More places to share Federal Energy Management Program: FEMP Designated Product: Water-Cooled Ice Machines on AddThis.com... Energy-Efficient Products Federal Requirements

152

Federal Energy Management Program: Covered Product Category: Water-Cooled  

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

Covered Product Covered Product Category: Water-Cooled Electric Chillers to someone by E-mail Share Federal Energy Management Program: Covered Product Category: Water-Cooled Electric Chillers on Facebook Tweet about Federal Energy Management Program: Covered Product Category: Water-Cooled Electric Chillers on Twitter Bookmark Federal Energy Management Program: Covered Product Category: Water-Cooled Electric Chillers on Google Bookmark Federal Energy Management Program: Covered Product Category: Water-Cooled Electric Chillers on Delicious Rank Federal Energy Management Program: Covered Product Category: Water-Cooled Electric Chillers on Digg Find More places to share Federal Energy Management Program: Covered Product Category: Water-Cooled Electric Chillers on AddThis.com...

153

SUPERHEATING IN A BOILING WATER REACTOR  

DOE Patents (OSTI)

A boiling-water reactor is described in which the steam developed in the reactor is superheated in the reactor. This is accomplished by providing means for separating the steam from the water and passing the steam over a surface of the fissionable material which is not in contact with the water. Specifically water is boiled on the outside of tubular fuel elements and the steam is superheated on the inside of the fuel elements.

Treshow, M.

1960-05-31T23:59:59.000Z

154

Process for treating effluent from a supercritical water oxidation reactor  

DOE Patents (OSTI)

The present invention relates generally to a method for treating and recycling the effluent from a supercritical water oxidation reactor and more specifically to a method for treating and recycling the effluent by expanding the effluent without extensive cooling. Supercritical water oxidation is the oxidation of fuel, generally waste material, in a body of water under conditions above the thermodynamic critical point of water. The current state of the art in supercritical water oxidation plant effluent treatment is to cool the reactor effluent through heat exchangers or direct quench, separate the cooled liquid into a gas/vapor stream and a liquid/solid stream, expand the separated effluent, and perform additional purification on gaseous, liquid, brine and solid effluent. If acid gases are present, corrosion is likely to occur in the coolers. During expansion, part of the condensed water will revaporize. Vaporization can damage the valves due to cavitation and erosion. The present invention expands the effluent stream without condensing the stream. Radionuclides and suspended solids are more efficiently separated in the vapor phase. By preventing condensation, the acids are kept in the much less corrosive gaseous phase thereby limiting the damage to treatment equipment. The present invention also reduces the external energy consumption, by utilizing the expansion step to also cool the effluent.

Barnes, C.M.; Shapiro, C.

1995-12-31T23:59:59.000Z

155

Pressurized Water Reactor Secondary Water Chemistry Guidelines – Revision 6  

Science Conference Proceedings (OSTI)

State-of-the-art water chemistry programs reduce equipment corrosion and enhance steam generator reliability. A committee of industry experts prepared these revised "Pressurized Water Reactor Secondary Water Chemistry Guidelines" to incorporate the latest field and laboratory data on secondary system corrosion and performance issues. Pressurized water reactor (PWR) operators can use these guidelines to update their secondary water chemistry programs.

2004-12-13T23:59:59.000Z

156

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

E-Print Network (OSTI)

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

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

2007-01-01T23:59:59.000Z

157

Best Practice for Energy Efficient Cleanrooms: Cooling tower and condenser water optimization  

E-Print Network (OSTI)

condenser water optimization Tengfang Xu Contents HVAC WATERHVAC Water Systems Cooling tower and condenser water optimization

Xu, Tengfang

2005-01-01T23:59:59.000Z

158

The Small Modular Liquid Metal Cooled Reactor: A New Approach to Proliferation Risk Management  

DOE Green Energy (OSTI)

There is an ongoing need to supply energy to small markets and remote locations with limited fossil fuel infrastructures. The Small, Modular, Liquid-Metal-Cooled Reactor, also referred to as SSTAR (Small, Secure, Transportable, Autonomous Reactor), can provide reliable and cost-effective electricity, heat, fresh water, and potentially hydrogen transportation fuels for these markets. An evaluation of a variety of reactor designs indicates that SSTAR, with its secure, long-life core, has many advantages for deployment into a variety of national and international markets. In this paper, we describe the SSTAR concept and its approach to safety, security, environmental and non-proliferation. The system would be design-certified using a new license-by-test approach, and demonstrated for commercial deployment anywhere in the world. The project addresses a technology development need (i.e., a small secure modular system for remote sites) that is not otherwise addressed in other currently planned research programs.

Smith, C F; Crawford, D; Cappiello, M; Minato, A; Herczeg, J W

2003-11-12T23:59:59.000Z

159

Boiling Water Reactor Sampling Summary: 2012 Update  

Science Conference Proceedings (OSTI)

This report documents boiling water reactor (BWR) sampling practices for key reactor water and feedwater parameters. It includes information on analysis methods, sampling frequencies, and compliance with the recommended sampling frequencies in BWRVIP-190: BWR Vessels and Internals Project, BWR Water Chemistry Guidelines – 2008 Revision (EPRI report 1016579).

2013-03-28T23:59:59.000Z

160

Fracture Mechanics Investigations on High-Temperature Gas-Cooled Reactor Materials  

Science Conference Proceedings (OSTI)

C.5. Fracture Mechanic / Status of Metallic Materials Development for Application in Advanced High-Temperature Gas-Cooled Reactor / Material

Klaus Krompholz; Erik Bodmann; Günter K. H. Gnirss; Horst Huthmann

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


161

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

E-Print Network (OSTI)

??Sodium cooled fast reactors (SFRs) are considered as a novel example to exercise the Technology Neutral Framework (TNF) proposed in NUREG- 1860. One reason for… (more)

Johnson, Brian C. (Brian Carl)

2010-01-01T23:59:59.000Z

162

Mechanical Properties of Welds in Commercial Alloys for High-Temperature Gas-Cooled Reactor Components  

Science Conference Proceedings (OSTI)

C. 1. Mechanical Property / Status of Metallic Materials Development for Application in Advanced High-Temperature Gas-Cooled Reactor / Material

James R. Lindgren; Brian E. Thurgood; Robin H. Ryder; Chia-Chuan Li

163

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

E-Print Network (OSTI)

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

Whitman, Joshua (Joshua J.)

2009-01-01T23:59:59.000Z

164

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

E-Print Network (OSTI)

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

Whitman, Joshua (Joshua J.)

2007-01-01T23:59:59.000Z

165

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

DOE Green Energy (OSTI)

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

Lee O. Nelson

2011-04-01T23:59:59.000Z

166

Chemical Treatment Fosters Zero Discharge by Making Cooling Water Reusable  

E-Print Network (OSTI)

Over the past decade, the water requirements for cooling industrial manufacturing processes have changed dramatically. Once-through cooling has been largely replaced by open recirculating cooling water methods. This approach reduces water consumption by increasing the use of recycled water. Simplistically, the circulating cooling water flows through heat exchanger equipment and is cooled by passing through a cooling tower. The recycled water is cooled by evaporation of some of the circulating water as it passes through the tower. As a result of the evaporation process, the dissolved solids in the water become concentrated. The evaporated water is replaced by fresh makeup water. The dissolved solids content of the water is maintained by the rate of water discharge (blowdown). As the amount of dissolved solids increases, their solubility is exceeded and the solids tend to precipitate from the cooling water. The precipitated scale adheres to heat transfer surfaces and reduces heat transfer efficiency. In order to achieve zero discharge of water, it is paramount that the potential for scale formation and deposition be minimized. This can be accomplished through physical separation of scale-forming ions and particulate matter. Two widely used mechanical methods in this category are lime-soda side stream softening and vapor compression blowdown evaporation. Another approach is chemical treatment to promote scale inhibition and dispersion.

Boffardi, B. P.

1996-04-01T23:59:59.000Z

167

Accident Tolerant Fuels for Light Water Reactors  

Science Conference Proceedings (OSTI)

Presentation Title, Accident Tolerant Fuels for Light Water Reactors. Author(s), Steven J. Zinkle, Kurt A. Terrani, Lance L. Snead. On-Site Speaker (Planned) ...

168

Topical report : NSTF facilities plan for water-cooled VHTR RCCS : normal operational tests.  

DOE Green Energy (OSTI)

As part of the Department of Energy (DOE) Generation IV roadmapping activity, the gas-cooled Very High Temperature Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept.

Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C. P.; Lomperski, S.; Aeschlimann, R. W.; Nuclear Engineering Division

2006-09-01T23:59:59.000Z

169

THERMAL STRESS CALCULATIONS FOR HEATPIPE-COOLED REACTOR POWER SYSTEMS.  

DOE Green Energy (OSTI)

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

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

2001-01-01T23:59:59.000Z

170

GAS COOLED PEBBLE BED REACTOR FOR A LARGE CENTRAL STATION. Reactor Design and Feasibility Study  

SciTech Connect

An optimum econonic design for a high temperature, helium cooled, central station reactor power plant of about 400 Mw of electric power was determined. The core consists of a randomly packed bed of unclad graphite spheres, approximately one in. in diameter, impregnated with U/sup 233/ and thorium such that a conversion ratio of near unity is achieved. The high temperature helium permits steam conditions, at the turbine throttle, of 1000 deg F and 1450 psia. (auth)

Schock, A.; Bruley, D.F.; Culver, H.N.; Ianni, P.W.; Kaufman, W.F.; Schmidt, R.A.; Supp, R.E.

1957-08-01T23:59:59.000Z

171

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Assessment of High Value Surveillance Materials Assessment of High Value Surveillance Materials Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely

172

IEP - Water-Energy Interface: Advanced Cooling Technology  

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

Cooling Technology Cooling Technology This component of the program is focused on research to develop technologies that improve performance and reduce costs associated with wet cooling, dry cooling, and hybrid cooling technologies. In addition, the research area covers innovative methods to control bio-fouling of cooling water intake structures as well as advances in intake structure systems. Read More! It is technically possible to cool power plants with minimal water use. However, at this time such cooling methods are not as economically feasible as traditional cooling systems. Additional research and development is necessary to develop cooling systems that use as little water as possible, but at a reasonable cost. Water intake structures are also an area of concern, especially considering the Clean Water Act 316(b) regulation which requires that the location, design, construction, and capacity of cooling water intake structures reflect the best technology available for minimizing adverse environmental impact. With plant intake structures, the particular concern is impingement and entrainment of aquatic organisms.

173

PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP  

DOE Patents (OSTI)

A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

Puechl, K.H.

1963-09-24T23:59:59.000Z

174

Process Optimization of Cast Alloy 718 for Water Cooled Gas ...  

Science Conference Proceedings (OSTI)

FOR WATER COOLED GAS TURBINE APPLICATION. G.K. Bouse+ and P.W. Schilke*. Gene@ Electric Company+ Materials and Processes Laboratory, and.

175

Water-side Economizer for Non-Fan Cooling Systems  

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

changes to the commercial provisions of the 2012 IECC: Water-side Economizer for Non-Fan Cooling Systems R Hart Pacific Northwest National Laboratory January 2013 Proposal...

176

Gas-Cooled Fast Reactor (GFR) FY05 Annual Report  

SciTech Connect

The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom and Switzerland), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above for this fiscal year. In addition, this report fulfills the Level 2 milestones, ''Complete annual status report on GFR reactor design'', and ''Complete annual status report on pre-conceptual GFR reactor designs'' in work package GI0401K01. GFR funding for FY05 included FY04 carryover funds, and was comprised of multiple tasks. These tasks involved a consortium of national laboratories and universities, including the Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Oak Ridge National Laboratory (ORNL), Auburn University (AU), Idaho State University (ISU), and the University of Wisconsin-Madison (UW-M). The total funding for FY05 was $1000K, with FY04 carryover of $174K. The cost breakdown can be seen in Table 1.

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

2005-09-01T23:59:59.000Z

177

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

SciTech Connect

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

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

2005-09-01T23:59:59.000Z

178

Where are we and where are we going in reactor shielding  

SciTech Connect

Information on reactor shielding is presented concerning nuclear data in shielding calculations, analysis methods, water-cooled reactors, fast-breeder reactors, and fusion reactors.

Maienschein, F.C.

1977-01-01T23:59:59.000Z

179

Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2  

Science Conference Proceedings (OSTI)

The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

2002-09-01T23:59:59.000Z

180

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

SciTech Connect

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

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

2011-01-01T23:59:59.000Z

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


181

CIVILIAN POWER REACTOR PROGRAM. PART III. STATUS REPORT ON GAS-COOLED REACTORS AS OF 1959. Book 8  

SciTech Connect

The technology of natural-uranium-fueled graphitemoderated gas-cooled reactor power plants is summarized for its relevance to the technology of enriched-fuel graphite-moderated systems. The technology of D/sub 2/Omoderated gas-cooled reactors is also summarized. Estimated technical performance parameters are given for the enriched-fuel prototype and for a large natural- uraniumfueled plant. Current technical status is discussed in terms of reactor physics, heat transfer and fluid flow, core materials, components, plant design and conctruction, and hazards. Detailed tables of characteristics for various reactors are given. An extensive bibliography is included. (W.D.M.)

1960-01-01T23:59:59.000Z

182

A Free Cooling Based Chilled Water System at Kingston  

E-Print Network (OSTI)

In efforts to reduce operating costs, the IBM site at Kingston, New York incorporated the energy saving concept of 'free cooling' (direct cooling of chilled water with condenser water) with the expansion of the site chilled water system. Free cooling was employed to satisfy the winter chilled water load of approximately 3000 tons resulting in electrical savings of up to 70% in the winter with wet bulb temperatures below 38 oF. Other energy efficient features included variable speed pumping, high efficiency motors and chillers with reduced entering condenser water limits. This paper will describe the various possible operating modes and their associated savings using computer simulation techniques.

Jansen, P. R.

1984-01-01T23:59:59.000Z

183

Use of nanofiltration to reduce cooling tower water usage.  

SciTech Connect

Nanofiltration (NF) can effectively treat cooling-tower water to reduce water consumption and maximize water usage efficiency of thermoelectric power plants. A pilot is being run to verify theoretical calculations. A side stream of water from a 900 gpm cooling tower is being treated by NF with the permeate returning to the cooling tower and the concentrate being discharged. The membrane efficiency is as high as over 50%. Salt rejection ranges from 77-97% with higher rejection for divalent ions. The pilot has demonstrated a reduction of makeup water of almost 20% and a reduction of discharge of over 50%.

Sanchez, Andres L.; Everett, Randy L.; Jensen, Richard Pearson; Cappelle, Malynda A.; Altman, Susan Jeanne

2010-09-01T23:59:59.000Z

184

Use of nanofiltration to reduce cooling tower water consumption.  

Science Conference Proceedings (OSTI)

Nanofiltration (NF) can effectively treat cooling-tower water to reduce water consumption and maximize water usage efficiency of thermoelectric power plants. A pilot is being run to verify theoretical calculations. A side stream of water from a 900 gpm cooling tower is being treated by NF with the permeate returning to the cooling tower and the concentrate being discharged. The membrane efficiency is as high as over 50%. Salt rejection ranges from 77-97% with higher rejection for divalent ions. The pilot has demonstrated a reduction of makeup water of almost 20% and a reduction of discharge of over 50%.

Altman, Susan Jeanne; Ciferno, Jared

2010-10-01T23:59:59.000Z

185

Water-Cooled Ice Machines, Purchasing Specifications for Energy...  

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

Ice Makers. b) Measured in accordance with ARI Standard 810-2003. Does not include condenser water use. Buying Energy-Efficient Water-Cooled Ice Machines Several types of...

186

TA-2 Water Boiler Reactor Decommissioning Project  

Science Conference Proceedings (OSTI)

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m{sup 3} of low-level solid radioactive waste and 35 m{sup 3} of mixed waste. 15 refs., 25 figs., 3 tabs.

Durbin, M.E. (ed.); Montoya, G.M.

1991-06-01T23:59:59.000Z

187

Multifunctional robot to maintain boiler water-cooling tubes  

Science Conference Proceedings (OSTI)

A robot has been developed to maintain boiler water-cooling tubes. This robot has a double tracked moving mechanism, an ash cleaning device, a slag purging device, a tubes' thickness measurement device, a marking device, and a control system. This robot ... Keywords: Boiler maintenance, Boiler water-cooling tube, Climbing robot, Mobile robot

Xueshan Gao; Dianguo Xu; Yan Wang; Huanhuan Pan; Weimin Shen

2009-10-01T23:59:59.000Z

188

Best Practices for Energy Efficient Cleanrooms: Cooling Tower and Condenser Water Optimization  

E-Print Network (OSTI)

LBNL-58634 Best Practices for Energy Efficient Cleanrooms: Cooling Tower and Condenser Water Efficient Cleanrooms: Cooling tower and condenser water optimization Tengfang Xu Contents HVAC WATER SYSTEMS.............................................................................................. 2 Cooling tower and condenser water optimization

189

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Initial Assessment of Thermal Annealing Needs and Challenges Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from 40y to 80y implies a doubling of the neutron exposure for the RPV. Thus,

190

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

SciTech Connect

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

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

1994-05-01T23:59:59.000Z

191

Closed Cooling Water Chemistry Guideline, Revision 1: Revision 1 to TR-107396, Closed Cooling Water Chemistry Guideline  

Science Conference Proceedings (OSTI)

This "Closed Cooling Water Chemistry Guideline" addresses the use of chemicals to mitigate corrosion, fouling, and microbiological growth in the closed cooling water (CCW) systems of nuclear power plants. The chemical additives used for these purposes depend on plant design characteristics, water quality, operating parameters, and the specifications of nuclear steam supply system (NSSS) suppliers. The list of chemicals is not as extensive as that for service water systems but can be confusing to utility ...

2004-04-23T23:59:59.000Z

192

Cooling Towers: Understanding Key Components of Cooling Towers and How to Improve Water Efficiency  

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

Paul Johnston-Knight Introduction Federal laws and regulations require Federal agencies to reduce water use and improve water efficiency. Namely, Executive Order 13514 Federal Leadership in Environmental, Energy, and Economic Performance, requires an annual two percent reduction of water use intensity (water use per square foot of building space) for agency potable water consumption as well as a two percent reduction of water use for industrial, landscaping, and agricultural applica- tions. Cooling towers can be a significant

193

Electric Power Plant Cooling Water Intakes and Related Water  

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

Impact of Drought on U.S. Steam Impact of Drought on U.S. Steam Electric Power Plant Cooling Water Intakes and Related Water Resource Management Issues April 2009 DOE/NETL-2009/1364 Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference therein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement,

194

Effect of Watering and Non-Watering Cooling Rates on the ...  

Science Conference Proceedings (OSTI)

Effect of Watering and Non-Watering Cooling Rates on the Mechanical Properties of an Aluminum Smelter's Potshell · Energy Reduction Technology for  ...

195

Compact intermediate heat transport system for sodium cooled reactor  

SciTech Connect

This patent describes a combination with a sodium cooled reactor having an intermediate heat exchanger for extracting heat in a nonradioactive secondary sodium loop from the sodium rector. It comprises: first and second upstanding closed cylindrical vessels, one of the cylindrical vessels being exterior of the other of the cylindrical vessels; the other of the cylindrical vessels being interior, smaller, and concentric of the larger cylindrical vessel so as to define between the inside of the larger vessel and the outside of the smaller vessel an interstitial annular volume; at least one feedwater inlet plenums at the bottom of the larger vessel communicated to the interstitial annular volume; at least one feedwater outlet plenums at the top of the larger and outer vessel communicated to the interstitial annular volume; tubes communicated to the feedwater inlet plenum at the bottom of the vessels and to the steam outlet plenum at the top of the vessel; a first conduit; a large submersible electromagnetic pump; and a jet pump having an inlet, a venturi, and a diffusing outlet.

Boardman, C.E.; Maurer, J.P.

1990-03-06T23:59:59.000Z

196

High temperature gas-cooled reactor: gas turbine application study  

SciTech Connect

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

Not Available

1980-12-01T23:59:59.000Z

197

Use of Reclaimed Water for Power Plant Cooling  

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

CONTENTS Chapter 1 - Introduction .......................................................................................................... 1 Power Plants Need Water .................................................................................................. 1 Meeting Water Demands in a Water-Constrained Environment ....................................... 3 Purpose and Structure of the Report .................................................................................. 3 Chapter 2 - Database of Reclaimed Water Use for Cooling ................................................... 5 Data Collection .................................................................................................................. 5 The Database...................................................................................................................... 7

198

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

E-Print Network (OSTI)

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

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

199

Pressurized Water Reactor Secondary Water Chemistry Guidelines - Revision 7  

Science Conference Proceedings (OSTI)

State-of-the-art water chemistry programs reduce equipment corrosion and enhance steam generator reliability. A committee of industry experts prepared these revised PWR Secondary Water Chemistry Guidelines to incorporate the latest field and laboratory data on secondary system corrosion and performance issues. Pressurized water reactor (PWR) operators can use these guidelines to update their secondary water chemistry programs.

2009-02-17T23:59:59.000Z

200

Use of Degraded Water Sources as Cooling Water in Power Plants  

Science Conference Proceedings (OSTI)

In electricity production, nearly all thermal power plants reject heat either to a large body of water (once-through cooling) or to the atmosphere via wet cooling towers — the predominant form of cooling in California. These towers, however, use considerable quantities of water. Competing state demands for freshwater have forced California thermal power plants to consider alternative cooling water supplies, though the availability of such supplies and data on their use and impact is limited. In fac...

2003-10-13T23:59:59.000Z

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


201

Gravity Scaling of a Power Reactor Water Shield  

SciTech Connect

Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa{sup n}. These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

Reid, Robert S.; Pearson, J. Boise [NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States)

2008-01-21T23:59:59.000Z

202

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT, MAY 1962  

SciTech Connect

Research progress is reported on water-cooled reactors, liquid-metal- cooled reactors, general reactor technology, plutonium recycle, advanced systems research and development, and nuclear safety. (M.C.G.)

1962-06-15T23:59:59.000Z

203

Study on Water-Cooled Solar Semiconductor Air Conditioner  

Science Conference Proceedings (OSTI)

Water-cooled solar semiconductor air conditioner was designed. Relevant calculation was made to determine the room's cooling load, which export the solar panels and battery capacity, followed by selection of CNC matcher. Development work also involves ... Keywords: solar energy, peltier effect, semiconductor air conditioner

Dong Zhi-Ming; Chang Ji-Bin; Xiang Li-Juan; Zhou Xue-Bin

2012-04-01T23:59:59.000Z

204

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

E-Print Network (OSTI)

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

Diecker, Jane T

2005-01-01T23:59:59.000Z

205

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

E-Print Network (OSTI)

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

Nitta, Christopher C

2010-01-01T23:59:59.000Z

206

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

E-Print Network (OSTI)

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

McKee, Stephanie A

2008-01-01T23:59:59.000Z

207

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

E-Print Network (OSTI)

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

Johnson, Brian C.

208

Application of the technology neutral framework to sodium cooled fast reactors  

E-Print Network (OSTI)

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

Johnson, Brian C. (Brian Carl)

2010-01-01T23:59:59.000Z

209

Performance of a hotel chilled water plant with cool storage  

SciTech Connect

A comprehensive monitoring suite was installed at a large convention hotel located in San Francisco, CA. The instrumentation was used for a research project to evaluate the effectiveness of electricity price based controls that automate response to real time pricing and to characterize the operation and performance of the hotel's chilled water plant that included a newly installed ice cool storage system. The hotel operates under real-time electricity rates. To date, over four years of data have been collected. Data included electricity use for all chillers, secondary coolant, chilled water, condenser pumps, and the cooling tower fans. Thermal flow data were also collected for the storage system, ice chiller, direct cooling chillers, and chilled water load loops. This paper (1) describes the chilled water plant, (2) defines the performance measurement objectives for the project, (3) discusses operational experience with the plant, focusing on the cool storage system, (4) analyzes chilled water plant and cool storage system operation by examining the charge/discharge heat flow data, and (5) evaluates how well the plant as a whole and the cool storage system specifically met cooling loads of the facility, and how this affected their use.

Gillespie, K.L.; Blanc, S.L.; Parker, S.

1999-07-01T23:59:59.000Z

210

Use of Alternate Water Sources for Power Plant Cooling  

Science Conference Proceedings (OSTI)

This report lays out a framework developed to evaluate the potential use of non-traditional water supplies for cooling new or existing power plants. The report will be of value to environment, generation, and planning managers within power companies.

2008-03-31T23:59:59.000Z

211

Water Cooling of High Power Light Emitting Diode Henrik Srensen  

E-Print Network (OSTI)

Water Cooling of High Power Light Emitting Diode Henrik Sørensen Department of Energy Technology and product lifetime. The high power Light Emitting Diodes (LED) belongs to the group of electronics

Sørensen, Henrik

212

CALIOP: a multichannel design code for gas-cooled fast reactors. Code description and user's guide  

Science Conference Proceedings (OSTI)

CALIOP is a design code for fluid-cooled reactors composed of parallel fuel tubes in hexagonal or cylindrical ducts. It may be used with gaseous or liquid coolants. It has been used chiefly for design of a helium-cooled fast breeder reactor and has built-in cross section information to permit calculations of fuel loading, breeding ratio, and doubling time. Optional cross-section input allows the code to be used with moderated cores and with other fuels.

Thompson, W.I.

1980-10-01T23:59:59.000Z

213

A STEAM POWER INSTALLATION FOR NUCLEAR POWER PLANT WITH GAS-COOLED REACTORS  

SciTech Connect

A steam power plant is designed for use with gas-cooled power reactors. In this plant, the turbine is divided into two sections, one high pressure and the other low pressure, the low-pressure turbine being the condensing turbine. The feed water from the condensing turbine is divided into two streams, one of which is brought to a higher pressure than the other. The high-pressure feed water is evaporated and superheated in the heat exchanger and then supplied to the high-pressure turbine, while the low-pressure feed water is evaporated and mixed with the exhaust steam of the highpressure turbine before superhenting and then passing to the low-pressure condensing turbine. Circulation of the reactor coolant is effected by a blower driven by a series turbine with no regulating devices and arranged in the steam plant circuit upstream of the low-pressure turbine; such a turbine works with constant efficiency over its whole load range. (D.L.C.)

1961-03-01T23:59:59.000Z

214

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these experienced cladding failures as operational capabilities of the different designs were being established. In addition, numerous spills of heavy water occurred within the facility. Currently, radiation and radioactive contamination levels are low within HWCTR with most of the radioactivity contained within the reactor vessel. There are no known insults to the environment, however with the increasing deterioration of the facility, the possibility exists that contamination could spread outside the facility if it is not decommissioned. An interior panoramic view of the ground floor elevation taken in August 2009 is shown in Figure 2. The foreground shows the transfer coffin followed by the reactor vessel and control rod drive platform in the center. Behind the reactor vessel is the fuel pool. Above the ground level are the polar crane and the emergency deluge tank at the top of the dome. Note the considerable rust and degradation of the components and the interior of the containment building. Alternative studies have concluded that the most environmentally safe, cost effective option for final decommissioning is to remove the reactor vessel, steam generators, and all equipment above grade including the dome. Characterization studies along with transport models have concluded that the remaining below grade equipment that is left in place including the transfer coffin will not contribute any significant contamination to the environment in the future. The below grade space will be grouted in place. A concrete cover will be placed over the remaining footprint and the groundwater will be monitored for an indefinite period to ensure compliance with environmental regulations. The schedule for completion of decommissioning is late FY2011. This paper describes the concepts planned in order to remove the major components including the dome, the reactor vessel (RV), the two steam generators (SG), and relocating the transfer coffin (TC).

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

215

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these experienced cladding failures as operational capabilities of the different designs were being established. In addition, numerous spills of heavy water occurred within the facility. Currently, radiation and radioactive contamination levels are low within HWCTR with most of the radioactivity contained within the reactor vessel. There are no known insults to the environment, however with the increasing deterioration of the facility, the possibility exists that contamination could spread outside the facility if it is not decommissioned. An interior panoramic view of the ground floor elevation taken in August 2009 is shown in Figure 2. The foreground shows the transfer coffin followed by the reactor vessel and control rod drive platform in the center. Behind the reactor vessel is the fuel pool. Above the ground level are the polar crane and the emergency deluge tank at the top of the dome. Note the considerable rust and degradation of the components and the interior of the containment building. Alternative studies have concluded that the most environmentally safe, cost effective option for final decommissioning is to remove the reactor vessel, steam generators, and all equipment above grade including the dome. Characterization studies along with transport models have concluded that the remaining below grade equipment that is left in place including the transfer coffin will not contribute any significant contamination to the environment in the future. The below grade space will be grouted in place. A concrete cover will be placed over the remaining footprint and the groundwater will be monitored for an indefinite period to ensure compliance with environmental regulations. The schedule for completion of decommissioning is late FY2011. This paper describes the concepts planned in order to remove the major components including the dome, the reactor vessel (RV), the two steam generators (SG), and relocating the transfer coffin (TC).

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

216

Licensing topical report: applicability of Division 1 regulatory guides to high-temperature gas-cooled reactors  

SciTech Connect

The application of Division 1 (power reactors) regulatory guides to high-temperature gas-cooled reactors (HTGRs) is discussed. About eighty of the Division 1 guides can be applied to any type of reactor; the remaining sixty, mostly written for light water reactors (LWRs), are divided between (1) those not applicable to the HTGR because of fundamental differences in design, (2) those applicable in intent but containing positions specific to LWRs, and (3) those written for LWRs but of sufficient generality to be applied to the HTGR without major exception. Emphasis is placed on issues which involve the unique characteristics of the HTGR. The regulatory guides evaluated are those extant as of early 1980. The positions presented are subject to periodic updating owing to the continuing modification of the guides and because the design options for the HTGR are at various stages of development. Nevertheless, this report is believed to provide a sound basis for evaluating conformance with existing Division 1 guides.

Lewis, J.H.

1980-12-01T23:59:59.000Z

217

Cold-water injection reanalysis for N Reactor  

Science Conference Proceedings (OSTI)

A re-analysis of the cold water injection transient, with a space- dependent, two-dimensional reactor kinetics code TWIGL has been completed. The analyses considered the impact of flux flattening on the consequences of this accident. Separate categories of cold water source were evaluated. Introduction of a sixth steam generator cell, postulated as a significant cold water transient in previous studies, was re-analyzed in greater depth. Activation of the emergency core cooling system (ECCS) and the main steam line break on the secondary side were also evaluated for worst-case condition. In all instances, the results of the analyses confirmed that N Reactor is well protected against the consequences of cold water reactivity transients by appropriate trip settings and by fast acting control systems. Accidents were analyzed for the possibilities that the control rods failed to insert, and safe shutdown was accomplished with the ball backup safety system. All calculations were performed for the flattened core. The flux flattened core did not alter the timing or the severity of the transient. The results of the re-analyses compare favorably with the analysis discussed in N Reactor Updated Safety Analysis Report (NUSAR) (UNI 1978). Total control aspects of cold water injection, a steady-state analysis, are unaffected by the conclusions of this report. The document contains detailed discussion of the computer analyses including the preparation of input, underlying assumptions, code validation discussion, and comparisons to past work. 10 refs., 27 figs.

Toffer, H.; Crowe, R.D.; Fortner, R.L.; Mohr, C.L.

1988-02-01T23:59:59.000Z

218

Property:CoolingTowerWaterUseWinterConsumed | Open Energy Information  

Open Energy Info (EERE)

search Property Name CoolingTowerWaterUseWinterConsumed Property Type Number Description Cooling Tower Water use (winter average) (afday) Consumed. Retrieved from "http:...

219

Property:CoolingTowerWaterUseSummerConsumed | Open Energy Information  

Open Energy Info (EERE)

Name CoolingTowerWaterUseSummerConsumed Property Type Number Description Cooling Tower Water use (summer average) (afday) Consumed. Retrieved from "http:en.openei.orgw...

220

Property:CoolingTowerWaterUseAnnlAvgConsumed | Open Energy Information  

Open Energy Info (EERE)

Name CoolingTowerWaterUseAnnlAvgConsumed Property Type Number Description Cooling Tower Water use (annual average) (afday) Consumed. Retrieved from "http:en.openei.orgw...

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


221

Property:CoolingTowerWaterUseSummerGross | Open Energy Information  

Open Energy Info (EERE)

Property Name CoolingTowerWaterUseSummerGross Property Type Number Description Cooling Tower Water use (summer average) (afday) Gross. Retrieved from "http:en.openei.orgw...

222

Purification of water from cooling towers and other heat exchange systems  

SciTech Connect

The amount of silica in cooling tower water is reduced by passing cooling tower water through a column of silica gel.

Sullivan; Enid J. (Los Alamos, NM), Carlson; Bryan J. (Ojo Caliente, NM), Wingo; Robert M. (Los Alamos, NM), Robison; Thomas W. (Stilwell, KS)

2012-08-07T23:59:59.000Z

223

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

Science Conference Proceedings (OSTI)

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

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

2006-07-01T23:59:59.000Z

224

CALIFORNIA ENERGY COMMISSION STAFF COOLING WATER MANAGEMENT  

E-Print Network (OSTI)

maintenance includes having effective drift eliminators, periodically cleaning the system if appropriate, minimization of process leads into the cooling system that provide nutrients for bacteria, maintenance management plan shall describe how the system will be returned to normal microbial control following an upset

225

Corrosion Product Transport during Boiling Water Reactor and Pressurized Water Reactor Startups  

Science Conference Proceedings (OSTI)

Corrosion product transport to Pressurized Water Reactor (PWR) steam generators and to the Boiling Water Reactor (BWR) reactor vessel during startups is of increased interest due to reductions in feedwater transport rates during normal operation and the recent emphasis on minimizing total transport during the cycle. Reductions in transport will reduce deposition on the fuel and the tendency for hot spot formation in BWRs and reduce surface fouling and the tendency for formation of aggressive chemical sol...

2010-12-17T23:59:59.000Z

226

DUSEL Facility Cooling Water Scaling Issues  

Science Conference Proceedings (OSTI)

Precipitation (crystal growth) in supersaturated solutions is governed by both kenetic and thermodynamic processes. This is an important and evolving field of research, especially for the petroleum industry. There are several types of precipitates including sulfate compounds (ie. barium sulfate) and calcium compounds (ie. calcium carbonate). The chemical makeup of the mine water has relatively large concentrations of sulfate as compared to calcium, so we may expect that sulfate type reactions. The kinetics of calcium sulfate dihydrate (CaSO4 {center_dot} 2H20, gypsum) scale formation on heat exchanger surfaces from aqueous solutions has been studied by a highly reproducible technique. It has been found that gypsum scale formation takes place directly on the surface of the heat exchanger without any bulk or spontaneous precipitation in the reaction cell. The kinetic data also indicate that the rate of scale formation is a function of surface area and the metallurgy of the heat exchanger. As we don't have detailed information about the heat exchanger, we can only infer that this will be an issue for us. Supersaturations of various compounds are affected differently by temperature, pressure and pH. Pressure has only a slight affect on the solubility, whereas temperature is a much more sensitive parameter (Figure 1). The affect of temperature is reversed for calcium carbonate and barium sulfate solubilities. As temperature increases, barium sulfate solubility concentrations increase and scaling decreases. For calcium carbonate, the scaling tendencies increase with increasing temperature. This is all relative, as the temperatures and pressures of the referenced experiments range from 122 to 356 F. Their pressures range from 200 to 4000 psi. Because the cooling water system isn't likely to see pressures above 200 psi, it's unclear if this pressure/scaling relationship will be significant or even apparent. The most common scale minerals found in the oilfield include calcium carbonates (CaCO3, mainly calcite) and alkaline-earth metal sulfates (barite BaSO4, celestite SrSO4, anhydrite CaSO4, hemihydrate CaSO4 1/2H2O, and gypsum CaSO4 2H2O or calcium sulfate). The cause of scaling can be difficult to identify in real oil and gas wells. However, pressure and temperature changes during the flow of fluids are primary reasons for the formation of carbonate scales, because the escape of CO2 and/or H2S gases out of the brine solution, as pressure is lowered, tends to elevate the pH of the brine and result in super-saturation with respect to carbonates. Concerning sulfate scales, the common cause is commingling of different sources of brines either due to breakthrough of injected incompatible waters or mixing of two different brines from different zones of the reservoir formation. A decrease in temperature tends to cause barite to precipitate, opposite of calcite. In addition, pressure drops tend to cause all scale minerals to precipitate due to the pressure dependence of the solubility product. And we can expect that there will be a pressure drop across the heat exchanger. Weather or not this will be offset by the rise in pressure remains to be seen. It's typically left to field testing to prove out. Progress has been made toward the control and treatment of the scale deposits, although most of the reaction mechanisms are still not well understood. Often the most efficient and economic treatment for scale formation is to apply threshold chemical inhibitors. Threshold scale inhibitors are like catalysts and have inhibition efficiency at very low concentrations (commonly less than a few mg/L), far below the stoichiometric concentrations of the crystal lattice ions in solution. There are many chemical classes of inhibitors and even more brands on the market. Based on the water chemistry it is anticipated that there is a high likelihood for sulfate compound precipitation and scaling. This may be dependent on the temperature and pressure, which vary throughout the system. Therefore, various types and amounts of scaling may occur at different

Daily, W D

2011-04-05T23:59:59.000Z

227

Method of detecting leakage of reactor core components of liquid metal cooled fast reactors  

DOE Patents (OSTI)

A method of detecting the failure of a sealed non-fueled core component of a liquid-metal cooled fast reactor having an inert cover gas. A gas mixture is incorporated in the component which includes Xenon-124; under neutron irradiation, Xenon-124 is converted to radioactive Xenon-125. The cover gas is scanned by a radiation detector. The occurrence of 188 Kev gamma radiation and/or other identifying gamma radiation-energy level indicates the presence of Xenon-125 and therefore leakage of a component. Similarly, Xe-126, which transmutes to Xe-127 and Kr-84, which produces Kr-85.sup.m can be used for detection of leakage. Different components are charged with mixtures including different ratios of isotopes other than Xenon-124. On detection of the identifying radiation, the cover gas is subjected to mass spectroscopic analysis to locate the leaking component.

Holt, Fred E. (Richland, WA); Cash, Robert J. (Richland, WA); Schenter, Robert E. (Richland, WA)

1977-01-01T23:59:59.000Z

228

Designing a 'Near Optimum' Cooling-Water System  

E-Print Network (OSTI)

Cooling water is expensive to circulate. Reducing its flow - i.e., hiking exchanger outlet temperatures - can cut tower, pump and piping investment as much as one-third and operating cost almost in half. Heat-exchanger-network optimization has been accomplished in large integrated plants, such as petroleum refineries. In many of the chemical process industries, however, a plant contains several individual processes, and network optimization, except on a limited basis, is not feasible. So far, no one has developed similar procedures for designing and optimizing a cooling-water once through-exchanger system. This article attempts to fill the void by presenting a design basis that will produce a 'near optimum' system. A cooling-water system consists of four major components: heat exchangers, cooling towers, circulation piping and pumps. To optimize such a system, one must define the system interactions and apply these relationships to the simultaneous design of the aforementioned equipment. This article develops criteria that for most applications allow one to ignore system interactions, and still design a 'near optimum' system. Cooling-water systems have long been designed by 'rules of thumb' that call for fixing the cool ant temperature-rise across all heat exchangers (usually 20 F) and setting the coolant inlet temperature to the heat exchanger at the site's wet-bulb temperature plus 8 F. These rules produce a workable cooling system; but, by taking the same coolant rise across all exchangers, regardless of the individual process outlet-temperatures, this cannot result in an optimized design. The design method presented in this article replaces the 'rules of thumb' with criteria that are easy to apply and that take into account the effect that the individual exchanger process outlet- temperatures have on cooling-system economics. Economic analyses of actual process have shown that cooling-system investment can be reduced by one third, and cooling-system operating cost by one half, If the proposed design criteria are used instead of the 'rules of thumb.' It has been found that the controlling economic factor for a cooling system is the quantity of water being circulated. Reducing the flow (raising the coolant outlet temperature of heat exchangers) significantly reduces cooling tower, pump and piping investment, and operating cost, and only moderately increases the heat-exchanger investment. The overriding conclusion to be drawn is that cooling water is very expensive, and its conservation can result in significant savings.

Crozier, R. A., Jr.

1981-01-01T23:59:59.000Z

229

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

E-Print Network (OSTI)

Design Windows for a He Cooled Fusion Reactor* Dai-Kai Sze and Ahmed Hassanein Argonne National Laboratory 9700 South Cass Avenue, Argonne, IL 60439 EQUATIONDERIVATION ABSTRACT A design window concept by this design window concept. INTRODUCTION Helium is an attractive coolant for both fusion and fission reactors

Harilal, S. S.

230

Evaluation of models for predicting evaporative water loss in cooling impoundments  

E-Print Network (OSTI)

Cooling impoundments can offer a number of advantages over cooling towers for condenser water cooling at steam electric power plants. However, a major disadvantage of cooling ponds is a lack of confidence in the ability ...

Helfrich, Karl Richard

1982-01-01T23:59:59.000Z

231

BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES  

DOE Patents (OSTI)

This patent covers the use of injection nozzles for pumping water into the lower ends of reactor fuel tubes in which water is converted directly to steam. Pumping water through fuel tubes of this type of boiling water reactor increases its power. The injection nozzles decrease the size of pump needed, because the pump handles only the water going through the nozzles, additional water being sucked into the tubes by the nozzles independently of the pump from the exterior body of water in which the fuel tubes are immersed. The resulting movement of exterior water along the tubes holds down steam formation, and thus maintains the moderator effectiveness, of the exterior body of water. (AEC)

Treshow, M.

1963-04-30T23:59:59.000Z

232

Computational Flow Predictions for the Lower Plenum of a High-Temperature, Gas-Cooled Reactor  

Science Conference Proceedings (OSTI)

Advanced gas-cooled reactors offer the potential advantage of higher efficiency and enhanced safety over present day nuclear reactors. Accurate simulation models of these Generation IV reactors are necessary for design and licensing. One design under consideration by the Very High Temperature Reactor (VHTR) program is a modular, prismatic gas-cooled reactor. In this reactor, the lower plenum region may experience locally high temperatures that can adversely impact the plant’s structural integrity. Since existing system analysis codes cannot capture the complex flow effects occurring in the lower plenum, computational fluid dynamics (CFD) codes are being employed to model these flows [1]. The goal of the present study is to validate the CFD calculations using experimental data.

Donna Post Guillen

2006-11-01T23:59:59.000Z

233

Overview of Component Testing Requirements for a Small Fluoride Salt-Cooled High Tempreature Reactor  

Science Conference Proceedings (OSTI)

This article summarizes the information necessary to provide reasonable assurance that components for a small fluoride salt-cooled high temperature reactor will meet their functional requirements. In support of the analysis of testing requirements, a simplified, conceptual description of the systems, structures, and components specific to this reactor class was developed. These reactor system elements were divided into major categories based on their functions: (1) reactor core systems, (2) heat transport system, (3) reactor auxiliary cooling system, and (4) instrumentation and controls system. An assessment of technical maturity for each element was made, and a gap analysis was performed to identify specific areas that require further testing. A prioritized list of the testing requirements was then developed. The prioritization was based on both the relative importance of the system to reactor viability, and performance and time requirements to perform the testing.

Cetiner, Mustafa Sacit [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

2010-01-01T23:59:59.000Z

234

An Improved Simple Chilled Water Cooling Coil Model  

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

An Improved Simple Chilled Water Cooling Coil Model An Improved Simple Chilled Water Cooling Coil Model Title An Improved Simple Chilled Water Cooling Coil Model Publication Type Conference Paper LBNL Report Number LBNL-6031E Year of Publication 2012 Authors Wang, Liping, Philip Haves, and Walter F. Buhl Conference Name SimBuild 2012 IBPSA Conference Date Published 08/2012 Abstract The accurate prediction of cooling and dehumidification coil performance is important in model-based fault detection and in the prediction of HVAC system energy consumption for support of both design and operations. It is frequently desirable to use a simple cooling coil model that does not require detailed specification of coil geometry and material properties. The approach adopted is to match the overall UA of the coil to the rating conditions and to estimate the air-side and water-side components of the UA using correlations developed by Holmes (1982). This approach requires some geometrical information about the coil and the paper investigates the sensitivity of the overall performance prediction to uncertainties in this information, including assuming a fixed ratio of air-side to water-side UA at the rating condition. Finally, simulation results from different coil models are compared, and experimental data are used to validate the improved cooling coil model.

235

Improving the Water Efficiency of Cooling Production System  

E-Print Network (OSTI)

For most of the time, cooling towers (CTs) of cooling systems operate under partial load conditions and by regulating the air circulation with a variable frequency drive (VFD), significant reduction in the fan power can be achieved. In Kuwait and other counties of Arabian Peninsula, reduced airflow can lead to reduction in water consumption as well, since during the summer season, the dry bulb temperature of the ambient air is higher than the incoming hot water temperature, and the air undergoes sensible cooling. This paper presents the findings of a study conducted in the Avenues mall, Kuwait. Initially, the CTs operated only at high speed, and on a typical summer day nearly one fourth of the make-up water was used for self cooling of air. The study based on measured data revealed that the use of VFD can reduce the water wastage for self-cooling of air by as much as 75% and overall water consumption by 18.6% while keeping the cooling system performance at design level.

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

2010-01-01T23:59:59.000Z

236

Water-lithium bromide double-effect absorption cooling analysis  

SciTech Connect

A numerical model was developed for the transient simulation of the double-effect, water-lithium bromide absorption cooling machine, and the use of the model to determine the effect of the various design and input variables on the absorption unit performance. The performance parameters considered were coefficient of performance and cooling capacity. The sensitivity analysis was performed by selecting a nominal condition and determining performance sensitivity for each variable with others held constant. The variables considered in the study include source hot water, cooling water, and chilled water temperatures; source hot water, cooling water, and chilled water flow rates; solution circulation rate; heat exchanger areas; pressure drop between evaporator and absorber; solution pump characteristics; and refrigerant flow control methods. The performance sensitivity study indicated in particular that the distribution of heat exchanger area among the various (seven) heat exchange components is a very-important design consideration. Moreover, it indicated that the method of flow control of the first effect refrigerant vapor through the second effect is a critical design feature when absorption units operate over a significant range of cooling capacity. The model was used to predict the performance of the Trane absorption unit with fairly good accuracy.

Vliet, G.C.; Lawson, M.B.; Lithgow, R.A.

1980-12-01T23:59:59.000Z

237

EVALUATION OF REACTOR CORE MATERIALS FOR A GAS-COOLED REACTOR EXPERIMENT  

DOE Green Energy (OSTI)

An evaluation of core materials for a gas-cooled reactor is being made. Work on the ZrH/sub n/ moderator has been confined to the high-hydrogen or delta- phase material. Methods for preparing sound hydride bodies of the highhydrogen composition have been developed. Both solid hydride and hydride powder compacts are being clad by a pressure-bonding technique. The hot hardness, tensile strength, thermal conductivity, thermal-expansion coefficient, and dissociation pressure of the delta-phase material are being determined. Control-material development was directed at rare-earth-oxide dispersions in Ni-chrome V or Co alloys. The reference fuel element is dispersedUO/sub 2/ in stainless steel. Studies include work on fabrication techniques and irradiation damage, and physical- and mechanical-property determinations. Several alternate fuels are being investigated. Gas-coolant studies involve N-metal and NH/sub 4/-metal reactions. Several additives to retard nitriding are being investigated. An in- pile-loop facility for testing reference materials is being constructed for operation in the Battelle Research Reactor. (auth)

Keller, D.L.

1957-07-11T23:59:59.000Z

238

Fuel Summary Report: Shippingport Light Water Breeder Reactor  

SciTech Connect

The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

Illum, D.B.; Olson, G.L.; McCardell, R.K.

1999-01-01T23:59:59.000Z

239

Plateout Phenomena in Direct-Cycle High Temperature Gas-Cooled Reactors  

Science Conference Proceedings (OSTI)

The plateout of condensable radionuclides in the primary coolant circuits of high-temperature gas-cooled reactors (HTGRs) -- particularly direct-cycle HTGRs -- has significant design, operations and maintenance (O&M), and safety implications. This report reviews and evaluates the available international information on plateout phenomena, specifically as it applies to the gas turbine-modular helium reactor (GT-MHR) and the pebble bed modular reactor (PBMR).

2002-06-26T23:59:59.000Z

240

Accident Performance of Light Water Reactor Cladding Materials  

DOE Green Energy (OSTI)

During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200 C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.

Nelson, Andrew T. [Los Alamos National Laboratory

2012-07-24T23:59:59.000Z

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241

Method and apparatus for enhancing reactor air-cooling system performance  

DOE Patents (OSTI)

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

Hunsbedt, A.

1996-03-12T23:59:59.000Z

242

Method and apparatus for enhancing reactor air-cooling system performance  

DOE Patents (OSTI)

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

Hunsbedt, Anstein (Los Gatos, CA)

1996-01-01T23:59:59.000Z

243

Use of reclaimed water for power plant cooling.  

SciTech Connect

Freshwater demands are steadily increasing throughout the United States. As its population increases, more water is needed for domestic use (drinking, cooking, cleaning, etc.) and to supply power and food. In arid parts of the country, existing freshwater supplies are not able to meet the increasing demands for water. New water users are often forced to look to alternative sources of water to meet their needs. Over the past few years, utilities in many locations, including parts of the country not traditionally water-poor (e.g., Georgia, Maryland, Massachusetts, New York, and North Carolina) have needed to reevaluate the availability of water to meet their cooling needs. This trend will only become more extreme with time. Other trends are likely to increase pressure on freshwater supplies, too. For example, as populations increase, they will require more food. This in turn will likely increase demands for water by the agricultural sector. Another example is the recent increased interest in producing biofuels. Additional water will be required to grow more crops to serve as the raw materials for biofuels and to process the raw materials into biofuels. This report provides information about an opportunity to reuse an abundant water source -- treated municipal wastewater, also known as 'reclaimed water' -- for cooling and process water in electric generating facilities. The report was funded by the U.S. Department of Energy's (DOE's) National Energy Technology Laboratory (NETL) Innovations for Existing Plants research program (Feeley 2005). This program initiated an energy-water research effort in 2003 that includes the availability and use of 'nontraditional sources' of water for use at power plants. This report represents a unique reference for information on the use of reclaimed water for power plant cooling. In particular, the database of reclaimed water user facilities described in Chapter 2 is the first comprehensive national effort to identify and catalog those plants that are using reclaimed water for cooling.

Veil, J. A.; Environmental Science Division

2007-10-16T23:59:59.000Z

244

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

SciTech Connect

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

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

2012-02-01T23:59:59.000Z

245

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

DOE Green Energy (OSTI)

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

Ian Mckirdy

2010-12-01T23:59:59.000Z

246

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

E-Print Network (OSTI)

Reuse of cooling tower blow down cannot only provide energy conservation, but can provide water conservation and chemical conservation. To be effective, it is critical that the water treatment program be coordinated with the treatment of the blow down for reuse into the cooling tower system. Several plants have been built and operated with considerable difficulty regarding effective operation of the softener due to improper chemical selection. However, other plants have utilized the proper chemicals which not only improve the softener's performance and operation, but also effectively reduces the size of the softener. Thus, initial capital and operating savings are obtained. Detailed information is provided on guidelines and case histories of operating units.

Puckorius, P. R.

1981-01-01T23:59:59.000Z

247

Water-Cooled Electric Chillers, Purchasing Specifications for Energy-Efficient Products (Fact Sheet)  

Science Conference Proceedings (OSTI)

Energy efficiency purchasing specifications for federal procurements of water-cooled electric chillers.

Not Available

2010-12-01T23:59:59.000Z

248

CIVILIAN POWER REACTOR PROGRAM. PART II. ECONOMIC POTENTIAL AND DEVELOPMENT PROGRAM. HEAVY WATER-MODERATED POWER REACTOR  

SciTech Connect

The reactor design which forms the base for the current economic status of D/sub 2/O-moderated reactors was estimated from developments in several reactor programs. However, since a heavy water-moderated reactor was not operated on natural U fuel at power reactor conditions, considerable improvement from this current status can be foreseen. A summary of improvements is presented concerning the concept which would result solely from operation of succeeding generation plants without a parallel development program, and improvements which would result from the successful completion of the development program as presented. One plant size was used in the evaluation of plant potential, with a 300 Mw/sub e/ nominal rating. The boiling D/sub 2/O-cooled, pressure tube direct cycle plant design was used. The current development program is outlined; this work includes several items leading to the long-range development of the concept. (auth)

Hutton, J.H.; Davis, S.A.; Graves, C.C.; Duffy, J.G. comps.

1960-08-19T23:59:59.000Z

249

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

E-Print Network (OSTI)

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

Pope, Michael A. (Michael Alexander)

2006-01-01T23:59:59.000Z

250

Light Water Reactor Sustainability Technical Documents | Department of  

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

Initiatives » Nuclear Reactor Technologies » Light Water Reactor Initiatives » Nuclear Reactor Technologies » Light Water Reactor Sustainability Program » Light Water Reactor Sustainability Technical Documents Light Water Reactor Sustainability Technical Documents September 30, 2011 Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement

251

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

DOE Green Energy (OSTI)

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

Marshall, A.C.

1989-08-01T23:59:59.000Z

252

Developmental Light-Water Reactor Program  

SciTech Connect

This report summarizes the progress of the Developmental Light-Water Reactor (DLWR) Program at Oak Ridge National Laboratory in FY 1989. It also includes (1) a brief description of the program, (2) definition of goals, (3) earlier achievements, and (4) proposed future activities.

Forsberg, C.W.

1989-12-01T23:59:59.000Z

253

Hydrogen and water reactor safety: proceedings  

DOE Green Energy (OSTI)

Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

Not Available

1982-01-01T23:59:59.000Z

254

Modeling cooling water discharges from the Burrard Generating Station  

E-Print Network (OSTI)

Abstract-A three-dimensional numerical model was applied to examine the impact of the Burrard Generating Station cooling water on the circulation patterns and thermal regime in the receiving water of Port Moody Arm. A key aspect of this study involved properly incorporating the submerged cooling water buoyant jet into the 3D model. To overcome the scale and interface barriers between the near-field and far-field zones of the buoyant jet, a sub-grid scheme was applied, and the coupled system of equations of motion, heat conservation and state are solved with a single modeling procedure over the complete field. Special care was taken with the diffusion and jet entrainment by using a second order turbulence closure model for vertical diffusion and the Smagorinsky formula for horizontal diffusion as well as jet entrainment. The model was calibrated and validated in terms of buoyant jet trajectory, centerline dilution, and temperature and velocity profiles. Extensive modeling experiments without and with the Burrard Generating Station in operation were then carried out to investigate the receiving water circulations and thermal processes under the influence of the cooling water discharge. The model results reveal that under the influence of the cooling water discharge, peak ebb currents are stronger than peak flood currents in the near-surface layer, and the reverse is true in the near-bottom layer. Meanwhile, the model revealed a well-developed eddy at the southeast side of the buoyant jet in the near-surface layer. It is also found that the warmer water released from the cooling water discharge is mainly confined to the upper layer of the Arm, which is largely flushed out of the Arm through tidal mixing processes, and a corresponding inflow of colder water into the Arm occurs within the lower layer. I.

J. Jiang; D. B. Fissel; D. D. Lemon

2002-01-01T23:59:59.000Z

255

100-MW NUCLEAR POWER PLANT UTILIZING A SODIUM COOLED, GRAPHITE MODERATED REACTOR  

SciTech Connect

The conceptual design of a 100 Mw(e) nuclear power plant is described. The plant utilized a sodium-cooled graphite-moderated reactor with stainless- steel clad. slightiy enriched UO/sub 2/ fuel. The reactor is provided with three main coolant circuits, and the steam cycle has three stages of regenerative heating. The plant control system allows automatic operation over the range of 20 to 100% load, or manual operation at all loads. The site, reactor, sodium systems, reactor auxiliaries, fuel handling, instrumentation, turbine-generator, buildings. and safety measures are described. Engineering drawings are included. (W.D.M.)

1958-02-28T23:59:59.000Z

256

Passive containment cooling system  

DOE Patents (OSTI)

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

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

1991-01-01T23:59:59.000Z

257

Keeping Cool, Saving Water and Money | Department of Energy  

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

Keeping Cool, Saving Water and Money Keeping Cool, Saving Water and Money Keeping Cool, Saving Water and Money July 2, 2010 - 2:25pm Addthis The Orlando Science Center has installed a new energy efficient HVAC unit. | Photo courtesy of Orlando Science Center The Orlando Science Center has installed a new energy efficient HVAC unit. | Photo courtesy of Orlando Science Center In the summer of 2009, the Orlando Science Center (OSC) was full of hot air, literally. The museum's heating, ventilation and air conditioning (HVAC) system - which had been an operational challenge for several years - was running at 30 percent capacity. That meant the building's interior temperature was often at a toasty 80 degrees, subjecting patrons to miserable conditions. "To keep visitors happy, the museum had to reduce admission prices and

258

Keeping Cool, Saving Water and Money | Department of Energy  

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

Cool, Saving Water and Money Cool, Saving Water and Money Keeping Cool, Saving Water and Money July 2, 2010 - 2:25pm Addthis The Orlando Science Center has installed a new energy efficient HVAC unit. | Photo courtesy of Orlando Science Center The Orlando Science Center has installed a new energy efficient HVAC unit. | Photo courtesy of Orlando Science Center In the summer of 2009, the Orlando Science Center (OSC) was full of hot air, literally. The museum's heating, ventilation and air conditioning (HVAC) system - which had been an operational challenge for several years - was running at 30 percent capacity. That meant the building's interior temperature was often at a toasty 80 degrees, subjecting patrons to miserable conditions. "To keep visitors happy, the museum had to reduce admission prices and

259

Program on Technology Innovation: Review of Advanced Cooling Tower Technologies with Reduced Cooled Water Temperature and Evaporatio  

Science Conference Proceedings (OSTI)

This report reviews current technologies and solutions for advanced cooling towers with reduced cooled water temperature and evaporation losses. This is the first report for the dew-point cooling tower fill development project, funded by the Electric Power Research Institute (EPRI) Program on Technology Innovation, Water Conservation program. It is prepared by the Gas Technology Institute (GTI).This review is based on a literature and patent survey; it summarizes advancements in cooling ...

2013-03-29T23:59:59.000Z

260

SELF-REGULATING BOILING-WATER NUCLEAR REACTORS  

DOE Patents (OSTI)

A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

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

1960-08-16T23:59:59.000Z

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


261

Light Water Reactor Sustainability Technical Documents | Department of  

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

Reactor Technologies » Light Water Reactor Reactor Technologies » Light Water Reactor Sustainability Program » Light Water Reactor Sustainability Technical Documents Light Water Reactor Sustainability Technical Documents April 30, 2013 LWRS Program and EPRI Long-Term Operations Program - Joint R&D Plan To address the challenges associated with pursuing commercial nuclear power plant operations beyond 60 years, the U.S. Department of Energy's (DOE) Office of Nuclear Energy (NE) and the Electric Power Research Institute (EPRI) have established separate but complementary research and development programs: DOE-NE's Light Water Reactor Sustainability (LWRS) Program and EPRI's Long-Term Operations (LTO) Program. April 30, 2013 Light Water Reactor Sustainability Program - Integrated Program Plan The Light Water Reactor Sustainability (LWRS) Program is a research and

262

Lifetime Test of a Partial Model of a High-Temperature Gas-Cooled Reactor Helium-Helium Heat Exchanger  

Science Conference Proceedings (OSTI)

H. Design Codes and Life Prediction / Status of Metallic Materials Development for Application in Advanced High-Temperature Gas-Cooled Reactor / Material

Masaki Kitagawa; Hiroshi Hattori; Akira Ohtomo; Tetsuo Teramae; Junichi Hamanaka; Hiroshi Ukikusa

263

Experimental and Analytical Simulation of MFCI (Molten Fuel Coolant Interaction) during CDA (Core Disruptive Accident) in Sodium Cooled Fast Reactor.  

E-Print Network (OSTI)

??With increasing demand for understanding Severe Accident Scenario in Sodium Cooled Fast Reactors, there is an urgent need of enhancing numerical and experimental simulation techniques.… (more)

Natarajan, Venkataraman

2011-01-01T23:59:59.000Z

264

Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor  

DOE Green Energy (OSTI)

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

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

2009-10-01T23:59:59.000Z

265

STUDY OF REMOTE MILITARY POWER APPLICATIONS. REPORT NO. 12. EVALUATION AND SELECTION OF APPLICABLE REACTOR CONCEPTS  

SciTech Connect

An evaluation of the reactor concepts under consideration for remote military power plants is presented. The concepts include water-cooled and - moderated reactors, both direct and indirect cycle. organic-cooled and -moderated reactors, heavy-water-cooled and -moderated reactors. gas-cooled reactors, sodium- cooled graphite-moderated reactors, fast reactors, and fluid-fuel reactors. The limitations and advantages, technological status, economics, and future potential of each reactor are reviewed. From the reviews it is concluded that direct-cycle boiling-water and pressurized-water reactors are most suitable for applications requiring power before 1965. (C.J.G.)

1960-01-01T23:59:59.000Z

266

Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.  

DOE Green Energy (OSTI)

As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong 3-D effects result in large heat flux, temperature, and heat transfer variations around the tube wall; (b) there is a large difference in the heat transfer coefficient predicted by turbulence models and heat transfer correlations, and this underscores the need of experimental work to validate the thermal performance of the RCCS; and (c) tests at the NSTF would embody all important fluid flow and heat transfer phenomena in the RCCS, in addition to covering the entire parameter ranges that characterize these phenomena. Additional supporting scaling study results are available in Reference 2. The purpose of this work is to develop a high-level engineering plan for mechanical and instrumentation modifications to NSTF in order to meet the following two technical objectives: (1) provide CFD and system-level code development and validation data for the RCCS under prototypic (full-scale) natural convection flow conditions, and (2) support RCCS design validation and optimization. As background for this work, the report begins by providing a summary of the original NSTF design and operational capabilities. Since the facility has not been actively utilized since the early 1990's, the next step is to assess the current facility status. With this background material in place, the data needs and requirements for the facility are then defined on the basis of supporting analysis activities. With the requirements for the facility established, appropriate mechanical and instrumentation modifications to NSTF are then developed in order to meet the overall project objectives. A cost and schedule for modifying the facility to satisfy the RCCS data needs is then provided.

Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

2005-09-01T23:59:59.000Z

267

Program on Technology Innovation: New Concepts of Water Conservation Cooling and Water Treatment Technologies  

Science Conference Proceedings (OSTI)

This report summarizes 114 proposals received as a result of Electric Power Research Institute (EPRI) worldwide solicitations that were distributed in February 2011 and June 2012. The solicitations sought early-stage concepts for innovative power plant water-conserving technologies for cooling, waste heat utilization, and water treatment. The report also highlights 18 cooling proposals from 2011, including 5 funded projects. In addition, it describes current cooling ...

2012-11-20T23:59:59.000Z

268

Advanced Light Water Reactor - Boiling Water Reactor Degradation Matrix (ALWR BWR DM), Revision 0  

Science Conference Proceedings (OSTI)

The advanced light water reactor–boiling water reactor degradation matrix (ALWR BWR DM) is an essential piece of the Electric Power Research Institute’s (EPRI’s) Advanced Nuclear Technology (ANT) materials management matrix initiative for advanced LWR designs. The materials management matrix provides a tool to assist the industry in proactive identification and consideration of materials issues as well as mitigation and management opportunities from the design phase, through component fabrication and pla...

2009-08-25T23:59:59.000Z

269

Condensate Polishing Guidelines for Pressurized Water Reactor and Boiling Water Reactor Plants - 2004 Revision  

Science Conference Proceedings (OSTI)

Successful condensate polishing allows more reliable operation of nuclear units by maintaining control of ionic and particulate impurity transport to the pressurized water reactor (PWR) steam generators and the boiling water reactor (BWR) and recirculation system. This report presents revisions of EPRI's 1997 nuclear industry consensus guidelines for the design and operation of deep bed and filter demineralizer condensate polishers. These guidelines are consistent with the 2000 revisions of EPRI's "BWR W...

2004-03-16T23:59:59.000Z

270

Cross section generation strategy for high conversion light water reactors  

E-Print Network (OSTI)

High conversion water reactors (HCWR), such as the Resource-renewable Boiling Water Reactor (RBWR), are being designed with axial heterogeneity of alternating fissile and blanket zones to achieve a conversion ratio of ...

Herman, Bryan R. (Bryan Robert)

2011-01-01T23:59:59.000Z

271

HEAVY-WATER-MODERATED POWER REACTORS ENGINEERING AND ECONOMIC EVALUATIONS. VOLUME I. SUMMARY REPORT  

SciTech Connect

Capital investments and the cost of power were estimated for 21 heavy- water-moderated, natural-uraniumfueled power-reactor plants, ranging in capacity from 100 to 460 Mw(e). Comparisons were made of hot- and coldmoderator reactors and of the relative merits of pressuretube and pressure-vessel designs. Reactors cooled with liquid D/sub 2/O, boiling D/sub 2/O, /sub 2/O steam, and helium were evalunted. A cold-moderator pressure-tube reactor cooled with boiling D/sub 2/O shows the most economic promise of the D/sub 2/Omoderated reactor systems studied to date. Reactors of this type have sufficient reactivity to permit satisfactory fuel exposures, but the development of additional technology is a prerequisite for optimum designs. At capacities of 300 and 400 Mw(e), the estimated power costs from the current designs of boiling-D/sub 2/O pressure-tabe reactor plants are 11.3 and 9.8 mills/kwh, respectively. From liquid-D/sub 2/-cooled concepts of comparable capacities the indicated power costs are 7 to 20% higher. With an active development program, a power cost of 8.0 to 8.5 mills/kwh may be attained in a 300 Mw(e) boiling-D/sub 2/O reactor plant within the next decade. (auth)

1960-06-01T23:59:59.000Z

272

POWER PLANT USING A STEAM-COOLED NUCLEAR REACTOR  

SciTech Connect

A method of providing efficient and economic means for obtaining reheat from nuclear heat is described. A steamcooled steam-moderated reactor produces high-pressure, high-temperature steam. A multi-stage steam turbine partially expands the high-pressure steam, which is then withdrawn and reheated, and then further expanded for producing useful power. The saturated steam is superheated by leading it through tubular passages provided in the fuel assemblies of a nuclear reactor, leading the useful part of the superheated steam into a steam turbine in which it expands to a predetermined intermediate pressure, leading the steam at that reduced pressure from the turbine back into the reactor where it is reheated by flowing through other tubular passages in the fuel assemblies, and returning the reheated steam to the turbine for further expansion. (M.C.G.)

Nettel, F.; Nakanishi, T.

1963-10-29T23:59:59.000Z

273

An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor  

SciTech Connect

The existing sodium cooled fast reactors (SFR) have two types of designs – loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphénix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL’s Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed.

Haihua Zhao; Hongbin Zhang

2007-11-01T23:59:59.000Z

274

Technology to Facilitate the Use of Impaired Water in Cooling...  

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

1 Research Circle Niskayuna, NY 12309-1027 518-387-5055 whisenhunt@crd.ge.com Technology To FaciliTaTe The Use oF impaired WaTers in cooling ToWers promisprojecT no.:...

275

Solar heating/cooling and domestic hot-water systems  

Science Conference Proceedings (OSTI)

Increasing awareness of global warming forces policy makers and industries to face two challenges: reducing greenhouse gas emissions and securing stable energy supply against ever-increasing world energy consumption, which is projected to increase by ... Keywords: buildings heating, domestic hot-water, energetical analysis, renewable energy sources, solar cooling technologies, solar energy collection, solar thermal systems

Ioan Sârbu; Marius Adam

2011-02-01T23:59:59.000Z

276

Development and Design of a Cooling Water Intake Structure Database  

Science Conference Proceedings (OSTI)

EPRI contracted Alden Laboratories, Inc. to develop an entrainment and impingement database (EIDB) in response to information needs that were identified from the U.S. Environmental Protection Agency's (EPA's) proposed revisions to Section 316(b) of the Clean Water Act (CWA). The original objective for the use of the EIDB was to determine if various environmental and plant operational factors influence entrainment and impingement of fish at cooling water intake structures (CWISs). It was subsequently dete...

2002-11-14T23:59:59.000Z

277

CONCEPTUAL DESIGN AND ECONOMIC EVALUATION OF A STEAM-COOLED FAST BREEDER REACTOR  

SciTech Connect

A conceptual design and economic evaluation of 300 and 40 MW/.sub e/ steam-cooled fast breeder reactor power plants were performed. A reactor core composed of U-Pu oxide rod-type fuel elements clad with Inconel-X and surrounded by a blanket of depleted UO/sub 2/ fuel was studied in some detail. Reactor breeding ratios of from 1.27 to 1.5 and overall system doubling times of from 20 to 30 years are achievable. For the near term (1967) 300 MW/sub e/ plant, an energy cost of 7.6 mills/kwh is estimated, based on AEC ground rules for privately financed plants and utilities. This cost may go down to 5.7 mills/kwh by 1975. For the 40 MW/sub e/ plant corresponding energy costs are 19.5 and 13.7 mills/kwh, r -spectively. The R&D program required for this reactor concept is estimated at million with an additional million for improvements leading to the 1975 reactor. Investigation of the operational and safety aspects of the reactor indicated that satisfactory procedures can be used for startup, shutdown, and emergency cooling of the reactor. An increase in reactivity upon flooding can be prevented by incorprating small amounts of high resonance absorption material in the core. Preliminary calculations indicate a substantial increase in reactivity upon loss of coolant for the 300 MW/sub e/ PuO/sub 2/ fueled reactor. To obtain designs with satisfactory voiding characteristics it may be necessary to provide high neutron leakage as ib a low L/D core or smaller volume core. Acceptable voiding characteristics appear possible with a Pu fueled 40 MWe reactor cooled with H/O steam, a Pu fueled 300MW/sub w/,reactor cooled with D/O steam, and a 300 MW/sub e/ U/sup 233/-Th fueled 300 MW/sub e/ breeder reactor cooled with H/sub 2/O steam. (auth)

Sofer, G.; Hankel, R.; Goldstein, L.; Birman, G.

1961-11-15T23:59:59.000Z

278

Utilization of Rainwater as a Supplementary Water Source for Cooling Tower Makeup: A Sustainability Strategy for Potable Water Use Reduction.  

E-Print Network (OSTI)

?? The use of rainwater as a supplementary water source for cooling water makeup was explored in an effort to reduce the potable water demand… (more)

Costello, Elizabeth Stassun

2012-01-01T23:59:59.000Z

279

Advanced Light Water Reactor utility requirements document  

SciTech Connect

The ALWR Requirements Document is a primary work product of the EPRI Program. This document is an extensive compilation of the utility requirements for design, construction and performance of advanced light water reactor power plants for the 1990s and beyond. The Requirements Document's primary emphasis is on resolution of significant problems experienced at existing nuclear power plants. It is intended to be used with companion documents, such as utility procurement specifications, which would cover the remaining detailed technical requirements applicable to new plant projects. The ALWR Requirements Document consists of several major parts. This volume is Part I, The Executive Summary. It is intended to serve as a concise, management level synopsis of advanced light water reactors including design objectives and philosophy, overall configuration and features and the steps necessary to proceed from the conceptual design stage to a completed, functioning power plant.

1986-06-01T23:59:59.000Z

280

Boiling Water Reactor Zinc Addition Sourcebook  

Science Conference Proceedings (OSTI)

Boiling water reactors (BWRs) have been injecting zinc into the primary coolant via the feedwater system for over 25 years to control primary system radiation fields. The zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. This evolution included transition from natural zinc oxide to depleted zinc oxide and from active zinc injection skids (pumped systems) to passive injection systems (zinc pellet beds).  Also occurring were various ...

2013-11-15T23:59:59.000Z

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


281

No Chemical, Zero Bleed Cooling Tower Water Treatment Process  

E-Print Network (OSTI)

This paper describes a process to treat cooling tower water by means of a fully automated and chemical free mechanical water treatment process. This is an alternative to conventional chemical treatment. Beginning with a suction pump to draw water out of the tower sump, water goes through a permanent magnetic descaler to increase the water solubility and begin the scale inhibition process. This also descales existing scale build-up in the system. Ozone is manufactured from ambient air and injected into the bypass system through a venturi type injector. This kills algae, slime and bacteria and enhances the magnetic descaling process. The final stage filter separates solids from the water to prevent corrosion from impingement. These solids are automatically purged to the sanitary drain. Clarified water is returned to the sump where the process repeats on a 10%-20% by volume side stream basis.

Coke, A. L.

1992-04-01T23:59:59.000Z

282

Light Water Reactor Sustainability (LWRS) Program | Department of Energy  

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

Light Water Reactor Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program The Light Water Reactor Sustainability (LWRS) Program is developing the scientific basis to extend existing nuclear power plant operating life beyond the current 60-year licensing period and ensure long-term reliability, productivity, safety, and security. The program is conducted in collaboration with national laboratories, universities, industry, and international partners. Idaho National Laboratory serves as the Technical Integration Office and coordinates the research and development (R&D) projects in the following pathways: Materials Aging and Degradation Assessment, Advanced Instrumentation, Information, and Control Systems

283

New approches for high temperature gas cooled reactors (HTGRs)  

Science Conference Proceedings (OSTI)

Several approaches are being evaluated in the US HTR Program to explore designs which might improve the commercial viability of nuclear power. The general approach is to reduce the power level of the reactor and increase ability to use passive methods for removing afterheat energy following extreme accidents. One approach most fully discussed in this paper is represented by modular HTRs for which the unit size and design are constrained such that extreme accidents do not result in significant release of radioactivity from the reactor circuit. Through such an approach, it should be possible to minimize the amount of nuclear grade components required in the balance-of-plant and achieve an economic system. Attaining such performance should provide low investment risk to the owner.

Kasten, P.R.; Cleveland, J.C.; Bowers, H.I.

1984-01-01T23:59:59.000Z

284

GAS-COOLED REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING SEPTEMBER 30, 1961  

SciTech Connect

Progress is reported on investigations in support of the Experimental Gas-Cooled Reactor, the Pebble-Bed Reactor Experiment, Advanced reactor design and development, test facilities, components, and materials. Topics covered include EGCR physics, EGCR performance analyses, structural investigations, EGCR component and materials development and testing, EGCR experimental facilities, PBRE physics and design studies, fueled-graphite investigations, clad fuel development, design studies of advanced power plants, experimental investigations of heat transfer and fluid flow, development of equipment anmd test facilities. and fabrication studies. (M.C.G.)

1962-02-01T23:59:59.000Z

285

Municipal waste water as a source of cooling water for California electric power plants. Final report  

SciTech Connect

This report discusses sources of municipal waste water for potential use as cooling water in California power plants. It notes the major factors which affect this practice. Municipal treatment facilities in California with discharge volumes deemed adequate to supply new power plants are identified. Also included is a summary of the experiences of several utilities in California and other western states with existing or planned applications of municipal waste water in power plant cooling towers.

MacDonald, T.

1980-05-01T23:59:59.000Z

286

Water cooled scavenged crankcase type otto internal combustion engine  

Science Conference Proceedings (OSTI)

In a system for a water cooled scavenged crankcase type two-cycle internal combustion engine comprising: a heat reclaimation system for extracting heat from an engine jacket for heating water supplied form an add-on reservoir via a heat exchanger located within the engine cylinder cooling system, the water being subsequently additionally heated by an exhaust pipe type heat exchanger to a superheated steam state and thence conveyed by a conduit to a steam lubricator for adjustably conveying in variably timed spaced succession intervals of regulated droplets of high viscous oil, fortified with adde graphite and tallow enrichment lubricant ingredients, and thence conveying such by steam at atmospheric pressure into an intake manifold which receives a carbureted air/fuel mixture into the crankcase via a manually operated auxiliary air intake device and way check valve and fire screen, due to suction effect of the piston up stroke action of the piston during engine operation.

Bidwell, H.

1988-10-25T23:59:59.000Z

287

Nuclear reactor with makeup water assist from residual heat removal system  

DOE Patents (OSTI)

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

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

1993-12-07T23:59:59.000Z

288

Nuclear reactor with makeup water assist from residual heat removal system  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

289

Heat pipe cooled reactors for multi-kilowatt space power supplies  

SciTech Connect

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

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

1995-01-01T23:59:59.000Z

290

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

E-Print Network (OSTI)

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

Petroski, Robert C

2008-01-01T23:59:59.000Z

291

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

Science Conference Proceedings (OSTI)

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

Lee Nelson

2011-09-01T23:59:59.000Z

292

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

SciTech Connect

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

Not Available

1980-01-01T23:59:59.000Z

293

Heat exchanger and water tank arrangement for passive cooling system  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

294

Deep burn strategy for the optimized incineration of reactor waste plutonium in pebble bed high temperature gas–cooled reactors / Serfontein D.E.  

E-Print Network (OSTI)

??In this thesis advanced fuel cycles for the incineration, i.e. deep–burn, of weapons–grade plutonium, reactor–grade plutonium from pressurised light water reactors and reactor–grade plutonium +… (more)

Serfontein, Dawid Eduard.

2013-01-01T23:59:59.000Z

295

A SODIUM COOLED, GRAPHITE MODERATED, LOW ENRICHMENT URANIUM REACTOR FOR THE PRODUCTION OF USEFUL POWER  

SciTech Connect

A design study is presented for a sodium cooled, graphite moderated power reactor utilizing low enrichment uranium fuel. The design is characterized by dependence on existing technology and the use of standard, or nearly standard, components. The reactor has a nominal rating of 167 thermal megawatts, and a plant comprising three such reactors for a total output of 500 thermal megawatts is described. Sodium in a secondary, non-radioactive, circulation system carries the heat to a steam generator at 910 deg F and is returned at 420 deg F. Steam conditions at the turbine throttle are 600 psig and 825 deg F. Cost of the complete reactor power plant, consisting of the three reactors and one 150- megawatt turbogenerator, is estimated to be approximately ,165,000. (auth)

Weisner, E.F. ed.

1954-09-15T23:59:59.000Z

296

Application of upspray type water distribution systems in cooling towers  

SciTech Connect

The efficient and uniform distribution of the warm circulating water on to the filling of cooling towers has been the continuing goal of the tower designer. The final element in the water distribution system, the sprayer, plays an important role in achieving this objective. This paper discusses the performance and operational characteristics of a sprayer utilized in counterflow towers that directs the water leaving the sprayer nozzle in an upward direction and briefly compares its performance with that of downward sprayers. The discussion also covers relative tower economics and application data of the sprayer.

Fay, H.P.; Hesse, G.

1985-01-01T23:59:59.000Z

297

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

SciTech Connect

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.

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y. [Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Physics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia)

2012-05-22T23:59:59.000Z

298

High Performance Fuel Desing for Next Generation Pressurized Water Reactors  

SciTech Connect

The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

Mujid S. Kazimi; Pavel Hejzlar

2006-01-31T23:59:59.000Z

299

GAS-COOLED REACTOR PROGRAM SEMIANNUAL PROGRESS REPORT FOR PERIOD ENDING MARCH 31, 1963  

SciTech Connect

Progress is reported on the development of gas-cooled reactors. The report contains eleven sections which are abstracted separately in NSA. These sections are contained in two parts: investigations in support of the Experimental Gas-Cooled Reactor and advanced reactor design and development. The four sections abstracted under Part I are: performance analyses, component development and testing, materials development, and irradiation testing of components and materials. The remaining sections are under Part II and they are: development of fueled-graphite bodies, investigations of fueled-graphite systems, clad fuel development, investigations of moderator materials, studies of advanced systems, experimental investigations of heat transfer and fluid flow, and facilities and equipment. (N.W.R.)

1963-07-23T23:59:59.000Z

300

SWR 1000: The Innovative Boiling Water Reactor  

SciTech Connect

Framatome ANP has developed the boiling water reactor SWR 1000 in close cooperation with German nuclear utilities and with support from various European partners. This advanced reactor design marks a new era in the successful tradition of boiling water reactor technology and, with a gross electric output of between 1290 and 1330 MW, is aimed at assuring competitive power generating costs compared to gas- and coal-fired stations. At the same time, the SWR 1000 meets the highest safety standards, including control of a core melt accident these objectives are met by supplementing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. The SWR 1000 fulfills international nuclear regulatory requirements and has been offered to TVO for the fifth nuclear unit in Finland. (authors)

Brettschuh, Werner [Framatome ANP GmbH, Berlinerstrasse 295, 63067 Offenbach (Germany); Hudson, Greg [Framatome ANP Inc., 400 South Tyron Street, Charlotte, NC 28285 (United States)

2004-07-01T23:59:59.000Z

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


301

USE of mine pool water for power plant cooling.  

Science Conference Proceedings (OSTI)

Water and energy production issues intersect in numerous ways. Water is produced along with oil and gas, water runs off of or accumulates in coal mines, and water is needed to operate steam electric power plants and hydropower generating facilities. However, water and energy are often not in the proper balance. For example, even if water is available in sufficient quantities, it may not have the physical and chemical characteristics suitable for energy or other uses. This report provides preliminary information about an opportunity to reuse an overabundant water source--ground water accumulated in underground coal mines--for cooling and process water in electric generating facilities. The report was funded by the U.S. Department of Energy's (DOE's) National Energy Technology Laboratory (NETL), which has implemented a water/energy research program (Feeley and Ramezan 2003). Among the topics studied under that program is the availability and use of ''non-traditional sources'' of water for use at power plants. This report supports NETL's water/energy research program.

Veil, J. A.; Kupar, J. M .; Puder, M. G.

2006-11-27T23:59:59.000Z

302

Environmental behavior of transuranic nuclides leaked from water cooled nuclear power plants. Final report, August 1, 1977-December 31, 1978  

Science Conference Proceedings (OSTI)

Release data are reported for three coastal water-cooled nuclear reactors: Millstone Point No. 1 and No. 2 (for the period January 1977 through April 1978), and Maine Yankee (for the period 20 June 1977 through 25 March 1978); release samples were analyzed for /sup 55/Fe, /sup 60/Co, /sup 134/Cs, /sup 137/Cs, /sup 238/Pu, /sup 239,240/Pu, /sup 241/Am, /sup 242/Cm, and /sup 244/Cm, but not all nuclides on every sample. Radioiron is a major component of the releases measured; the transuranium nuclides are less significant components than was expected, but levels have occasionally reached microcuries per month. Pulses of this size are adequate for tracer studies. Environmental samples (water, sediments, and biota) have been analyzed from about the two reactor sites noted, and that of the Pilgrim No. 1 reactor. No water samples remote from reactor outflows have unequivocally shown reactor contamination. No sediment samples from near Millstone Point or Pilgrim 1 have shown reactor contamination; this has been clearly evident in several sediment collections from near Maine Yankee. Biota so far measured from near Millstone Point show reactor contamination only when taken from the effluent canal. From the Maine Yankee and Plymouth areas, however, biota samples frequently prove to show slight, but definite, reactor contamination.

Bowen, V.T.

1980-09-01T23:59:59.000Z

303

Light-water breeder reactor (LWBR Development Program)  

DOE Patents (OSTI)

Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

Beaudoin, B.R.; Cohen, J.D.; Jones, D.H.; Marier, L.J. Jr.; Raab, H.F.

1972-06-20T23:59:59.000Z

304

Modelling of a passive reactor cavity cooling system (RCCS) for a nuclear reactor core subject to environmental changes and the optimisation of the RCCS radiation heat shield heat shield.  

E-Print Network (OSTI)

??ENGLISH ABSTRACT: A reactor cavity cooling system (RCCS) is used in the PBMR to protect the concrete citadel surrounding the reactor from direct nuclear radiation… (more)

Verwey, Aldo

2010-01-01T23:59:59.000Z

305

WATER-LITHIUM BROMIDE DOUBLE-EFFECT ABSORPTION COOLING ANALYSIS  

Office of Scientific and Technical Information (OSTI)

WATER-LITHIUM BROMIDE DOUBLE-EFFECT WATER-LITHIUM BROMIDE DOUBLE-EFFECT ABSORPTION COOLING ANALYSIS Gary C . V l i e t , Michael B . Lawson, and Rudolf0 A . Lithgow Center f o r Energy Studies The University of Texas a t Austin December 1980 Final Report f o r Contract: DE AC03-79SF10540 (Mu1 tiple-Effect Absorption Cycle Solar Cooling) with the U.S. Department of Energy DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately

306

Property:CoolingTowerWaterUseAnnlAvgGross | Open Energy Information  

Open Energy Info (EERE)

Property Name CoolingTowerWaterUseAnnlAvgGross Property Type Number Description Cooling Tower Water use (annual average) (afday) Gross. Retrieved from "http:en.openei.orgw...

307

pH Adjustment of Power Plant Cooling Water with Flue Gas/Fly Ash  

to fossil fuel burning power plants to control mineral precipitation in cooling water. Flue gas, which is 10% CO2, could be diverted into a plant’s cooling water

308

Municipal waste water as a source of cooling water for California electric power plants  

SciTech Connect

The results of an investigation of sources of municipal waste water for potential use as cooling water in California power plants and the major factors which affect this practice are presented. Municipal treatment facilities in California with discharge volumes deemed adequate to supply new power plants are identified. Also included is a summary of the experiences of several utilities in California and other western states with existing or planned applications of municipal waste water in power plant cooling towers. Due to limited supplies of high-quality water, municipal waste water is increasingly viewed as an alternative source of supply for a variety of water uses, including electric power plant evaporative cooling. In California, enough municipal effluent is discharged to the ocean to conceivably supply the total projected cooling water needs of new power plants for the next 20 years or more. A number of existing applications of such waste water for power plant cooling, including several California cases, demonstrate the technical feasibility of its use for this purpose. However, a combination of economic, environmental, and geographic factors reduce the likelihood of widespread use of this alternative for meeting anticipated large increases in power plant water requirements in the state. The most important factors are: the long distances involved; the public health concerns; added costs and environmental effects; and unreliability of supply quality.

McDonald, T.

1980-05-01T23:59:59.000Z

309

Cool water demonstration project and its industrial applications  

SciTech Connect

This paper discusses the 100 MW coal gasification combined cycle demonstration project underway at the ''Cool Water'' site of Southern California Edison Company, including the technology, project participants, schedule and opportunities for future industrial users. Industrial applications with multiple product outputs, termed ''polygeneration'', are illustrated with examples for cogeneration and trigeneration. Finally, actions required for planning large-sized gas turbine installations are suggested for today in order to hold open the future options in coal gasification.

Alger, J.; Ahner, D.J.

1982-08-01T23:59:59.000Z

310

Prevention of Flow Restrictions in Generator Stator Water Cooling Circuits  

Science Conference Proceedings (OSTI)

Generator stator water cooling systems are designed to operate with dissolved oxygen (DO) concentrations of either more than 2 parts per million (ppm) or less than 50 parts per billion (ppb). Large- and small-scale experimental studies and literature surveys show that if the system operates with DO concentration in either design range, copper corrosion-product-particle release rates are low and do not lead to plugging of hollow strands or clogging of strainers. In the range between the extremes -- ...

2002-02-06T23:59:59.000Z

311

Potential applications of helium-cooled high-temperature reactors to process heat use  

DOE Green Energy (OSTI)

High-Temperature Gas-Cooled Reactors (HTRs) permit nuclear energy to be applied to a number of processes presently utilizing fossil fuels. Promising applications of HTRs involve cogeneration, thermal energy transport using molten salt systems, steam reforming of methane for production of chemicals, coal and oil shale liquefaction or gasification, and - in the longer term - energy transport using a chemical heat pipe. Further, HTRs might be used in the more distant future as the energy source for thermochemical hydrogen production from water. Preliminary results of ongoing studies indicate that the potential market for Process Heat HTRs by the year 2020 is about 150 to 250 GW(t) for process heat/cogeneration application, plus approximately 150 to 300 GW(t) for application to fossil conversion processes. HTR cogeneration plants appear attractive in the near term for new industrial plants using large amounts of process heat, possibly for present industrial plants in conjunction with molten-salt energy distribution systems, and also for some fossil conversion processes. HTR reformer systems will take longer to develop, but are applicable to chemicals production, a larger number of fossil conversion processes, and to chemical heat pipes.

Gambill, W.R.; Kasten, P.R.

1981-01-01T23:59:59.000Z

312

Quarterly technical report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, April--June 1975  

SciTech Connect

The current water reactor safety activities of ANC are accomplished in four programs. The Semiscale Program consists of small-scale nonnuclear thermal- hydraulic experiments for the generation of experimental data that can be applied to analytical models describing loss-of-coolant accident (LOCA) phenomena in water-cooled nuclear power plants. Emphasis is placed on acquiring system effects data from integral tests that characterize the most significant thermal- hydraulic phenomena during the depressurization (blowdown) and emergency cooling phase of a LOCA. The LOFT Program provides test data to support: (a) assessment and improvement of the analytical methods utilized for predicting the behavior of pressurized water reactors (PWR) under LOCA conditions; (b) evaluation of the performance of PWR engineered safety features (ESF), particularly the emergency core cooling system (ECCS); and (c) assessment of the quantitative margins of safety inherent in the performance of these safety features. The Thermal Fuels Behavior Program is a program designed to provide information on the behavior of reactor fuels under normal, off-normal, and accident conditions. The experimental portion is concentrated on testing of single fuel rods and fuel rod clusters under power-cooling-mismatch (PCM), loss-of-coolant, and reactivity initiated accident conditions. The Reactor Behavior Program encompasses the analytical aspects of predicting the response of nuclear power reactors under normal, abnormal, and accident conditions. The status of each program is reported. (auth)

1975-11-01T23:59:59.000Z

313

New and Underutilized Technology: Water Cooled Oil Free Magnetic Bearing Compressors  

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

The following information outlines key deployment considerations for water cooled oil free magnetic bearing compressors within the Federal sector.

314

IRIS Reactor a Suitable Option to Provide Energy and Water Desalination for the Mexican Northwest Region  

SciTech Connect

The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity. The IRIS reactor offers a very suitable source of energy given its modular size of 300 MWe and it can be coupled with a desalination plant to provide the potable water for human consumption, agriculture and industry. The present paper assess the water and energy requirements for the Northwest region of Mexico and how the deployment of the IRIS reactor can satisfy those necessities. The possible sites for deployment of Nuclear Reactors are considered given the seismic constraints and the closeness of the sea for external cooling. And in the other hand, the size of the desalination plant and the type of desalination process are assessed accordingly with the water deficit of the region.

Alonso, G.; Ramirez, R.; Gomez, C.; Viais, J.

2004-10-03T23:59:59.000Z

315

ENVIRONMENTAL IMPACT OF THE COOLING WATER INTAKE STRUCTURE, TANGUISSON POWER PLANT  

E-Print Network (OSTI)

report for the Environmental Effects of Cooling Water Intake Structures project (contract number/or chemicals) and as impingement (where the cooling water intake traps larger organisms against the intake and impingement of aquatic organisms in cooling water intakes.) For rules such as those outlined above

Schupp, Peter

316

Coagulation chemistries for silica removal from cooling tower water.  

SciTech Connect

The formation of silica scale is a problem for thermoelectric power generating facilities, and this study investigated the potential for removal of silica by means of chemical coagulation from source water before it is subjected to mineral concentration in cooling towers. In Phase I, a screening of many typical as well as novel coagulants was carried out using concentrated cooling tower water, with and without flocculation aids, at concentrations typical for water purification with limited results. In Phase II, it was decided that treatment of source or make up water was more appropriate, and that higher dosing with coagulants delivered promising results. In fact, the less exotic coagulants proved to be more efficacious for reasons not yet fully determined. Some analysis was made of the molecular nature of the precipitated floc, which may aid in process improvements. In Phase III, more detailed study of process conditions for aluminum chloride coagulation was undertaken. Lime-soda water softening and the precipitation of magnesium hydroxide were shown to be too limited in terms of effectiveness, speed, and energy consumption to be considered further for the present application. In Phase IV, sodium aluminate emerged as an effective coagulant for silica, and the most attractive of those tested to date because of its availability, ease of use, and low requirement for additional chemicals. Some process optimization was performed for coagulant concentration and operational pH. It is concluded that silica coagulation with simple aluminum-based agents is effective, simple, and compatible with other industrial processes.

Nyman, May Devan; Altman, Susan Jeanne; Stewart, Tom

2010-02-01T23:59:59.000Z

317

Inherent Prevention and Mitigation of Severe Accident Consequences in Sodium-Cooled Fast Reactors  

SciTech Connect

Safety challenges for sodium-cooled fast reactors include maintaining core temperatures within design limits and assuring the geometry and integrity of the reactor core. Due to the high power density in the reactor core, heat removal requirements encourage the use of high-heat-transfer coolants such as liquid sodium. The variation of power across the core requires ducted assemblies to control fuel and coolant temperatures, which are also used to constrain core geometry. In a fast reactor, the fuel is not in the most neutronically reactive configuration during normal operation. Accidents leading to fuel melting, fuel pin failure, and fuel relocation can result in positive reactivity, increasing power, and possibly resulting in severe accident consequences including recriticalities that could threaten reactor and containment integrity. Inherent safety concepts, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, can be used to increase the level of safety to the point where it is highly unlikely, or perhaps even not credible, for such severe accident consequences to occur.

Roald A. Wigeland; James E. Cahalan

2011-04-01T23:59:59.000Z

318

REVIEW OF THE STATUS OF SUPERCRITICAL WATER REACTOR TECHNOLOGY  

SciTech Connect

Supercritical water-reactor design studies are reviewed. The status of supercritical water technology relative to heat transfer and fluid flow, water chemistry, internal deposition on heated surfaces, plant power cycles, and reactor construction materials is reviewed. The direct cycle was found to offer the highest probability for achieving economic power. (C.J.G.)

Marchaterre, J.F.; Petrick, M.

1960-08-01T23:59:59.000Z

319

Gas-cooled fast breeder reactor steady-state irradiation testing program  

Science Conference Proceedings (OSTI)

The requirements for the gas-cooled fast breeder reactor irradiation program are specified, and an irradiation program plan which satisfies these requirements is presented. The irradiation program plan consists of three parts and includes a schedule and a preliminary cost estimate: (1) a steady-state irradiation program, (2) irradiations in support of the design basis transient test program, and (3) irradiations in support of the GRIST-2 safety test program. Data from the liquid metal fast breeder reactor program are considered, and available irradiation facilities are examined.

Acharya, R.T.; Campana, R.J.; Langer, S.

1980-08-01T23:59:59.000Z

320

Heat exchanger and water tank arrangement for passive cooling system  

DOE Patents (OSTI)

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

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

1993-11-30T23:59:59.000Z

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


321

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

DOE Green Energy (OSTI)

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

Peters, Curtis D. (.)

2006-04-01T23:59:59.000Z

322

Materials Degradation in Light Water Reactors: Life After 60 | Department  

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

Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the

323

Materials Degradation in Light Water Reactors: Life After 60 | Department  

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

Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the

324

Simulating core melt accidents helps improve nuclear reactor...  

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

Did You Know? Almost every commercial reactor today is a light-water reactor.These reactors are cooled by water, which is cheap, easy to get, and well-understood. But many of the...

325

Low-pressure water-cooled inductively coupled plasma torch  

DOE Patents (OSTI)

An inductively coupled plasma torch is provided which comprises an inner tube, including a sample injection port to which the sample to be tested is supplied and comprising an enlarged central portion in which the plasma flame is confined; an outer tube surrounding the inner tube and containing water therein for cooling the inner tube, the outer tube including a water inlet port to which water is supplied and a water outlet port spaced from the water inlet port and from which water is removed after flowing through the outer tube; and an rf induction coil for inducing the plasma in the gas passing into the tube through the sample injection port. The sample injection port comprises a capillary tube including a reduced diameter orifice, projecting into the lower end of the inner tube. The water inlet is located at the lower end of the outer tube and the rf heating coil is disposed around the outer tube above and adjacent to the water inlet.

Seliskar, C.J.; Warner, D.K.

1984-02-16T23:59:59.000Z

326

Low-pressure water-cooled inductively coupled plasma torch  

DOE Patents (OSTI)

An inductively coupled plasma torch is provided which comprises an inner tube, including a sample injection port to which the sample to be tested is supplied and comprising an enlarged central portion in which the plasma flame is confined; an outer tube surrounding the inner tube and containing water therein for cooling the inner tube, the outer tube including a water inlet port to which water is supplied and a water outlet port spaced from the water inlet port and from which water is removed after flowing through the outer tube; and an r.f. induction coil for inducing the plasma in the gas passing into the tube through the sample injection port. The sample injection port comprises a capillary tube including a reduced diameter orifice, projecting into the lower end of the inner tube. The water inlet is located at the lower end of the outer tube and the r.f. heating coil is disposed around the outer tube above and adjacent to the water inlet.

Seliskar, Carl J. (Cincinnati, OH); Warner, David K. (Centerville, OH)

1988-12-27T23:59:59.000Z

327

Light Water Reactor Sustainability Program - Non-Destructive...  

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

Remaining Useful Life of Aging Cables in Nuclear Power Plants Light Water Reactor Sustainability Program - Non-Destructive Evaluation R&D Roadmap for Determining Remaining Useful...

328

Exploration of supercritical water gasification of biomass using batch reactor .  

E-Print Network (OSTI)

??The focus of this study is on gasification of a biomass in supercritical water. Vapor mass yield in a batch reactor after 20 minutes in… (more)

Venkitasamy, Chandrasekar

2011-01-01T23:59:59.000Z

329

The use of dispersants in pressurised water reactor steam generators.  

E-Print Network (OSTI)

??Environmental degradation promoted by the presence of sludge piles in the steam generators of Pressurised Water Reactors (PWR) can pose a threat to their safe… (more)

Tulloch, Sam

2011-01-01T23:59:59.000Z

330

Light Water Reactor Materials for Commercial Nuclear Power ...  

Science Conference Proceedings (OSTI)

Presentation Title, Light Water Reactor Materials for Commercial Nuclear ... First- Principles Theory of Magnetism, Crystal Field and Phonon Spectrum of UO2.

331

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

SciTech Connect

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

Wayne Moe

2013-05-01T23:59:59.000Z

332

Thermionic nuclear reactor with internal heat distribution and multiple duct cooling  

DOE Patents (OSTI)

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.

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

1975-11-01T23:59:59.000Z

333

Heavy Water and Graphite Reactors - Reactors designed/built by...  

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

experiments, necessary to achieve higher precision for the determination of reactor power distribution patterns, effect of non-uniform void distributions, kinetic behavior,...

334

Fluoride-Salt-Cooled High-Temperature Reactor (FHR) with Silicon-Carbide-Matrix Coated-Particle Fuel  

Science Conference Proceedings (OSTI)

The FHR is a new reactor concept that uses coated-particle fuel and a low-pressure liquid-salt coolant. Its neutronics are similar to a high-temperature gas-cooled reactor (HTGR). The power density is 5 to 10 times higher because of the superior cooling properties of liquids versus gases. The leading candidate coolant salt is a mixture of {sup 7}LiF and BeF{sub 2} (FLiBe) possessing a boiling point above 1300 C and the figure of merit {rho}C{sub p} (volumetric heat capacity) for the salt slightly superior to water. Studies are underway to define a near-term base-line concept while understanding longer-term options. Near-term options use graphite-matrix coated-particle fuel where the graphite is both a structural component and the primary neutron moderator. It is the same basic fuel used in HTGRs. The fuel can take several geometric forms with a pebble bed being the leading contender. Recent work on silicon-carbide-matrix (SiCm) coated-particle fuel may create a second longer-term fuel option. SiCm coated-particle fuels are currently being investigated for use in light-water reactors. The replacement of the graphite matrix with a SiCm creates a new family of fuels. The first motivation behind the effort is to take advantage of the superior radiation resistance of SiC compared to graphite in order to provide a stable matrix for hosting coated fuel particles. The second motivation is a much more rugged fuel under accident, repository, and other conditions.

Forsberg, C. W. [Massachusetts Institute of Technology (MIT); Terrani, Kurt A [ORNL; Snead, Lance Lewis [ORNL; Katoh, Yutai [ORNL

2012-01-01T23:59:59.000Z

335

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

Science Conference Proceedings (OSTI)

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

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

2008-04-01T23:59:59.000Z

336

Thermal Hydraulic Challenges of Gas Cooled Fast Reactors with Passive Safety Features  

SciTech Connect

Transient response of a Gas cooled Fast Reactor (GFR) coupled to a recompression supercritical CO2 (S-CO2) power conversion system (PCS) in a direct cycle to Loss of Coolant and Loss of Generator Load Accidents is analyzed using RELAP5-3D. A number of thermal hydraulic challenges for GFR design are pointed out as the designers strive to accommodate cooling of the high power density core of a fast reactor by a gas with its inherently low heat transfer capability, in particular under post LOCA events when system pressure is lost and when reliance on passive decay heat removal is emphasized. Although it is possible to design a S-CO2 cooled GFR that can survive LOCA by cooling the core through natural circulating loops between the core and elevated emergency cooling heat exchangers, it is not an attractive approach because of various bypass paths that can, depending on break location, degrade core cooling. Moreover, natural circulation gas loops can operate in deteriorated heat transfer regimes with substantial reduction of heat transfer coefficient: as low as 30% of forced convection values, and data and correlations in these regimes carry large uncertainties. Therefore, reliable battery powered blowers for post-LOCA decay heat removal (DHR) that provide flow in well defined regimes with low uncertainty, and can be easily over-designed to accommodate bypass flows were selected. The results confirm that a GFR with such a DHR system and negative coolant void worth can withstand LOCA with and without scram as well as loss of electrical load without exceeding core temperature and turbomachinery overspeed limits.

Michael Pope; Jeong-Ik Lee; Pavel Hejzlar; Michael J. Driscoll

2009-05-01T23:59:59.000Z

337

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

E-Print Network (OSTI)

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

Fray, Elliott Shepard

2013-01-01T23:59:59.000Z

338

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

E-Print Network (OSTI)

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

Short, Michael Philip

2010-01-01T23:59:59.000Z

339

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

SciTech Connect

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

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

2012-06-06T23:59:59.000Z

340

An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components  

SciTech Connect

This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

Holcomb, David Eugene [ORNL; Cetiner, Mustafa Sacit [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

2009-11-01T23:59:59.000Z

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


341

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

SciTech Connect

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

Ilas, Dan [ORNL

2012-01-01T23:59:59.000Z

342

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

SciTech Connect

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

Ilas, Dan [ORNL

2013-01-01T23:59:59.000Z

343

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

SciTech Connect

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

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

2013-11-01T23:59:59.000Z

344

Light water reactor lower head failure analysis  

SciTech Connect

This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

1993-10-01T23:59:59.000Z

345

GROUND WATER USE FOR COOLING: ASSOCIATED AQUIFER TEMPERATURE CHANGES  

E-Print Network (OSTI)

expensive or scarce, cooling towers or ponds are employed.~-1), for dry and wet cooling tower systems, respectively.condenser cooling sys terns such as towers or ponds are

Lippmann, Marcelo J.

2012-01-01T23:59:59.000Z

346

Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor  

SciTech Connect

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)

Pope, M. A.; Sen, R. S.; Ougouag, A. M.; Youinou, G. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Boer, B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); SCK-CEN, Boertang 200, BE-2400 Mol (Belgium)

2012-07-01T23:59:59.000Z

347

Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor  

SciTech Connect

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

2012-04-01T23:59:59.000Z

348

Optimum hot water temperature for absorption solar cooling  

SciTech Connect

The hot water temperature that maximizes the overall instantaneous efficiency of a solar cooling facility is determined. A modified characteristic equation model is used and applied to single-effect lithium bromide-water absorption chillers. This model is based on the characteristic temperature difference and serves to empirically calculate the performance of real chillers. This paper provides an explicit equation for the optimum temperature of vapor generation, in terms of only the external temperatures of the chiller. The additional data required are the four performance parameters of the chiller and essentially a modified stagnation temperature from the detailed model of the thermal collector operation. This paper presents and discusses the results for small capacity machines for air conditioning of homes and small buildings. The discussion highlights the influence of the relevant parameters. (author)

Lecuona, A.; Ventas, R.; Venegas, M.; Salgado, R. [Dpto. Ingenieria Termica y de Fluidos, Universidad Carlos III de Madrid, Avda. Universidad 30, 28911 Leganes, Madrid (Spain); Zacarias, A. [ESIME UPA, IPN, Av. de las Granjas 682, Col. Santa Catarina, 02550, D.F. Mexico (Mexico)

2009-10-15T23:59:59.000Z

349

Advanced water-cooled phosphoric acid fuel cell development  

DOE Green Energy (OSTI)

The Advanced Water Cooled Phosphoric Acid Fuel Cell Development program is being conducted by International Fuel Cells Corporation (IFC) to improve the performance and minimize the cost of water-cooled, electric utility phosphoric acid fuel cell stacks. The program adapts the existing on-site Configuration B cell design to electric utility operating conditions and introduces additional new design features. Task 1 consists of the conceptual design of a full-scale electric utility cell stack that meets program objectives. Tasks 2 and 3 develop the materials and processes required to fabricate the components that meet the program objective. The design of the small area and two 10-ft[sup 2] short stacks is conducted in Task 4. The conceptual design also is updated to incorporate the results of material and process developments, as well as results of stack tests conducted in Task 6. Fabrication and assembly of the short stacks are conducted in Task 5 and subsequent tests are conducted in Task 6. The Contractor expects to enter into a contract with the Electric Power Research Institute (EPRI) to assemble and endurance test the second 10-ft[sup 2] short stack. The management and reporting functions of Task 7 provide DOE/METC with program visibility through required documentation and program reviews. This report describes the cell design and development effort that is being conducted to demonstrate, by subscale stack test, the technical achievements made toward the above program objectives.

Not Available

1992-07-01T23:59:59.000Z

350

Utility of Variable Speed Drives for Fish Protection at Cooling Water Intakes  

Science Conference Proceedings (OSTI)

This report reviews the utility of a variable frequency drive (VFD) for reducing cooling water flow and potentially the extent of impingement and entrainment of fish and shellfish at power plant cooling water intakes. Reduction of impingement and entrainment at cooling water intakes is the objective of Clean Water Act 316(b) requirements that are being developed by the U.S. Environmental Protection Agency (EPA).

2010-06-03T23:59:59.000Z

351

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

E-Print Network (OSTI)

Pebble-fueled high temperature gas-cooled reactor (HTGR) technology was first developed by the Federal Republic of Germany in the 1950s. More recently, the design has been embraced by the People's Republic of China and the Republic of South Africa. Unlike light water reactors that generate heat from fuel assemblies comprised of fuel rods, pebble-fueled HTGRs utilize thousands of 60-mm diameter fuel spheres (pebbles) comprised of thousands of TRISO particles. As this reactor type is deployed across the world, adequate methods for safeguarding the reactor must be developed. Current safeguards methods for the pebble-fueled HTGR focus on extensive, redundant containment and surveillance (C/S) measures or a combination of item-type and bulk-type material safeguards measures to deter and detect the diversion of fuel pebbles. The disadvantages to these approaches are the loss of continuity of knowledge (CoK) when C/S systems fail, or are compromised, and the introduction of material unaccounted for (MUF). Either vulnerability can be exploited by an adversary to divert fuel pebbles from the reactor system. It was determined that a solution to maintaining CoK is to develop a system to identify each fuel pebble that is inserted and removed from the reactor. Work was performed to develop and evaluate the use of inert microspheres placed in each fuel pebble, whose random placement could be used as a fingerprint to identify the fuel pebble. Ultrasound imaging of 1 mm zirconium oxide microspheres was identified as a possible imaging system and microsphere material for the new safeguards system concept. The system concept was evaluated, and it was found that a minimum of three microspheres are necessary to create enough random fingerprints for 10,000,000 pebbles. It was also found that, over the lifetime of the reactor, less than 0.01% of fuel pebbles can be expected to have randomly the same microsphere fingerprint. From an MCNP 5.1 model, it was determined that less than fifty microspheres in each pebble will have no impact on the reactivity or temperature coefficient of reactivity of the reactor system. Finally, using an ultrasound system it was found that ultrasound waves can penetrate thin layers of graphite to image the microsphere fingerprint.

Gitau, Ernest Travis Ngure

2011-08-01T23:59:59.000Z

352

INFORMATION MEETING ON GAS-COOLED POWER REACTORS, OAK RIDGE NATIONAL LABORATORY, OCTOBER 21-22, 1958  

SciTech Connect

This meeting is one of a series of Civilian Power Reactor Conferences and was held colncident with an AEC invitation to industry to bid on the construction of a gas-cooled facility. Papers are presented on design studles, hazards, components, costs, materials, and design concepts for specific reactors. (W.D.M.)

1959-10-31T23:59:59.000Z

353

Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis  

SciTech Connect

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

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

2012-08-01T23:59:59.000Z

354

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

DOE Green Energy (OSTI)

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

Chang Oh

2008-02-01T23:59:59.000Z

355

Fuel cycle design and analysis of SABR: subrcritical advanced burner reactor.  

E-Print Network (OSTI)

??Various fuel cycles for a sodium-cooled, subcritical, fast reactor with a fusion neutron source for the transmutation of light water reactor spent fuel have been… (more)

Sommer, Christopher

2008-01-01T23:59:59.000Z

356

Environmentally assisted cracking in light water reactors.  

DOE Green Energy (OSTI)

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni-alloys and welds. A critical review of the ASME Code fatigue design margins and an assessment of the conservation in the current choice of design margins are presented. The existing fatigue {var_epsilon}-N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments. Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 10{sup 21} n x cm{sup -2}. The crack growth rates (CGRs) of the irradiated steels are a factor of {approx}5 higher than the disposition curve proposed in NUREG-0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low-dissolved oxygen (DO) environments. Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 C water on steels irradiated to {approx}3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289 C water. The IASCC susceptibility of SSs that contain >0.003 wt. % S increased drastically. bend tests in inert environments at 23 C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature indicate that IASCC in 289 C water is dominated by a crack-tip grain-boundary process that involves S. An initial IASCC model has been proposed. A crack growth test was completed on mill annealed Alloy 600 in high-purity water at 289 C and 320 C under various environmental and loading conditions. The results from this test are compared with data obtained earlier on several other heats of Alloy 600.

Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.

2007-11-06T23:59:59.000Z

357

Light Water Reactors [Corrosion and Mechanics of Materials] - Nuclear  

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

Light Water Reactors Light Water Reactors Capabilities Materials Testing Environmentally Assisted Cracking (EAC) of Reactor Materials Corrosion Performance/Metal Dusting Overview Light Water Reactors Fatigue Testing of Carbon Steels and Low-Alloy Steels Environmentally Assisted Cracking of Ni-Base Alloys Irradiation-Induced Stress Corrosion Cracking of Austenitic Stainless Steels Steam Generator Tube Integrity Program Air Oxidation Kinetics for Zr-based Alloys Fossil Energy Fusion Energy Metal Dusting Publications List Irradiated Materials Steam Generator Tube Integrity Other Facilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Corrosion and Mechanics of Materials Light Water Reactors Bookmark and Share To continue safe operation of current LWRs, the aging degradation of the

358

Light Water Reactor Sustainability Nondestructive Evaluation for Concrete  

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

Nondestructive Evaluation for Nondestructive Evaluation for Concrete Research and Development Roadmap Light Water Reactor Sustainability Nondestructive Evaluation for Concrete Research and Development Roadmap Materials issues are a key concern for the existing nuclear reactor fleet as material degradation can lead to increased maintenance, increased downtown, and increased risk. Extending reactor life to 60 years and beyond will likely increase susceptibility and severity of known forms of degradation. Additionally, new mechanisms of materials degradation are also possible. The purpose of the US Department of Energy Office of Nuclear Energy's Light Water Reactor Sustainability (LWRS) Program is to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend

359

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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.

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

1959-10-27T23:59:59.000Z

360

ATWS Transients for the 2400 MWt Gas-Cooled Fast Reactor  

SciTech Connect

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

Cheng,L.Y.; Ludewig, H.

2007-08-05T23:59:59.000Z

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


361

A utility assessment of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)  

SciTech Connect

A team of electric utility representatives conducted an in-depth, independent evaluation of the current Modular High Temperature Gas-Cooled Reactor (MHTGR) design. The emphasis was on the fuel design with respect to safety, the licensability of the proposed containment concept, refueling operations and equipment, spent fuel storage capacity, staffing projections, and the economic competitiveness. Specific comments and recommendations are provided as a contribution towards enhancing the MHTGR design, licensability and acceptance from a utility's view. Individual sections have been indexed separately for inclusion on the data base.

Bliss, H.E.; Grier, C.A. (Commonwealth Edison Co., Chicago, IL (USA)); Crews, M.R. (Duke Engineering and Services, Inc., Charlotte, NC (USA)); Fernandez, R.T.; Heard, J.W.; Hinkle, W.D. (Yankee Atomic Electric Co., Framingham, MA (USA)); Pschirer, D.M.; Sharpe, R.O. (Duke Power Co., Charlotte, NC (USA))

1991-01-01T23:59:59.000Z

362

Deep Burn Develpment of Transuranic Fuel for High-Temperature Helium-Cooled Reactors - July 2010  

SciTech Connect

The DB Program Quarterly Progress Report for April - June 2010, ORNL/TM/2010/140, was distributed to program participants on August 4. This report discusses the following: (1) TRU (transuranic elements) HTR (high temperature helium-cooled reactor) Fuel Modeling - (a) Thermochemical Modeling, (b) 5.3 Radiation Damage and Properties; (2) TRU HTR Fuel Qualification - (a) TRU Kernel Development, (b) Coating Development, (c) ZrC Properties and Handbook; and (3) HTR Fuel Recycle - (a) Recycle Processes, (b) Graphite Recycle, (c) Pyrochemical Reprocessing - METROX (metal recovery from oxide fuel) Process Development.

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

2010-08-01T23:59:59.000Z

363

High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics  

DOE Green Energy (OSTI)

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

Larry Demick

2010-08-01T23:59:59.000Z

364

Gas-cooled reactor programs annual progress report for period ending December 31, 1972  

SciTech Connect

Information on the gas-cooled reactor development programs is presented concerning HTGR head-end fuel reprocessing development; fuel microsphere preparation development; fuel fabrication process development; HTGR fuel recycle pilot-plant studies; studies and evaluation of commercial HTGR fuel recycle plants; HTGR fuel element development; HTGR fuel irradiations and postirradiation evaluations; HTGR fuel chemistry, fuel integrity, and fission product behavior; reactions of HTGR core materials with steam; fission product behavior in HTGR coolant circuits; HTGR safety program plan and safety analysis; prestressed concrete pressure vessel development; GCFR irradiation experiments; and GCFR steam generator modeling studies. (DCC)

1974-03-01T23:59:59.000Z

365

Scaling Analysis for the Direct Reactor Auxiliary Cooling System for AHTRs  

Science Conference Proceedings (OSTI)

The Direct Reactor Auxiliary Cooling System (DRACS), shown in Fig. 1 [1], is a passive heat removal system proposed for the Advanced High-Temperature Reactor (AHTR). It features three coupled natural circulation/convection loops completely relying on the buoyancy as the driving force. A prototypic design of the DRACS employed in a 20-MWth AHTR has been discussed in our previous work [2]. The total height of the DRACS is usually more than 10 m, and the required heating power will be large (on the order of 200 kW), both of which make a full-scale experiment not feasible in our laboratory. This therefore motivates us to perform a scaling analysis for the DRACS to obtain a scaled-down model. In this paper, theory and methodology for such a scaling analysis are presented.

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

2011-01-01T23:59:59.000Z

366

2400MWt GAS-COOLED FAST REACTOR DHR STUDIES STATUS UPDATE.  

Science Conference Proceedings (OSTI)

A topical report on demonstrating the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor was published in March 2006. The analysis was performed with the system code RELAP5-3D (version 2.4.1.1a) and the model included the full complement of the power conversion unit (PCU): heat exchange components (recuperator, precooler, intercooler) and rotating machines (turbine, compressor). A re-analysis of the success case in Ref is presented in this report. The case was redone to correct unexpected changes in core heat structure temperatures when the PCU model was first integrated with the reactor model as documented in Ref [1]. Additional information on the modeling of the power conversion unit and the layout of the heat exchange components is provided in Appendix A.

CHENG,L.Y.; LUDEWIG, H.

2007-06-01T23:59:59.000Z

367

Thermal hydraulic considerations in liquid-metal-cooled components of tokamak fusion reactors  

Science Conference Proceedings (OSTI)

The basic considerations of MHD thermal hydraulics for liquid-metal-cooled blankets and first walls of tokamak fusion reactors are discussed. The liquid-metal MHD program of Argonne National Laboratory (ANL) dedicated to analytical and experimental investigations of reactor relevant MHD flows and development of relevant thermal hydraulic design tools is presented. The status of the experimental program and examples of local velocity measurements are given. An account of the MHD codes developed to date at ANL is also presented as is an example of a 3-D thermal hydraulic analysis carried out with such codes. Finally, near term plans for experimental investigations and code development are outlined. 20 refs., 8 figs., 1 tab.

Picologlou, B.F.; Reed, C.B.; Hua, T.Q. (Argonne National Lab., IL (USA))

1989-01-01T23:59:59.000Z

368

Effects of evaporative cooling on the regulation of body water and ...  

Science Conference Proceedings (OSTI)

building, causing air to be drawn through the cooling pads. The study was conducted during two ...... of gut water in living ruminants. Aust J Agric Res 15:

369

pH Adjustment of Power Plant Cooling Water with Flue Gas/ Fly ...  

The discovery represents a cost-effective way to utilize materials indigenous to fossil fuel burning power platns to control mineral precipitation is cooling water.

370

Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors  

Science Conference Proceedings (OSTI)

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (k{sub eff}) is in almost linear relations with the change of the fuel volume to coolant ratio.

Ariani, Menik [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Su'ud, Zaki; Waris, Abdul; Asiah, Nur [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Shafii, M. Ali [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Physics Department, Andalas University, Kampus Limau Manis, Padang, Sumatera Barat (Indonesia); Khairurrijal

2010-12-23T23:59:59.000Z

371

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

DOE Green Energy (OSTI)

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

Gorensek, M.

2011-07-06T23:59:59.000Z

372

Transpiring wall supercritical water oxidation test reactor design report  

Science Conference Proceedings (OSTI)

Sandia National Laboratories is working with GenCorp, Aerojet and Foster Wheeler Development Corporation to develop a transpiring wall supercritical water oxidation reactor. The transpiring wall reactor promises to mitigate problems of salt deposition and corrosion by forming a protective boundary layer of pure supercritical water. A laboratory scale test reactor has been assembled to demonstrate the concept. A 1/4 scale transpiring wall reactor was designed and fabricated by Aerojet using their platelet technology. Sandia`s Engineering Evaluation Reactor serves as a test bed to supply, pressurize and heat the waste; collect, measure and analyze the effluent; and control operation of the system. This report describes the design, test capabilities, and operation of this versatile and unique test system with the transpiring wall reactor.

Haroldsen, B.L.; Ariizumi, D.Y.; Mills, B.E.; Brown, B.G. [Sandia National Labs., Livermore, CA (United States). Engineering for Transportation and Environment Dept.; Rousar, D.C. [GenCorp Aerojet, Sacramento, CA (United States)

1996-02-01T23:59:59.000Z

373

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

DOE Green Energy (OSTI)

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

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

2010-11-01T23:59:59.000Z

374

Light Water Reactor Sustainability Nondestructive Evaluation...  

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

to the safety of LWR plants include the containment building, spent fuel pool, and cooling towers. This use has made its long-term performance crucial for the safe operation...

375

Energy penalty analysis of possible cooling water intake structurerequirements on existing coal-fired power plants.  

SciTech Connect

Section 316(b) of the Clean Water Act requires that cooling water intake structures must reflect the best technology available for minimizing adverse environmental impact. Many existing power plants in the United States utilize once-through cooling systems to condense steam. Once-through systems withdraw large volumes (often hundreds of millions of gallons per day) of water from surface water bodies. As the water is withdrawn, fish and other aquatic organisms can be trapped against the screens or other parts of the intake structure (impingement) or if small enough, can pass through the intake structure and be transported through the cooling system to the condenser (entrainment). Both of these processes can injure or kill the organisms. EPA adopted 316(b) regulations for new facilities (Phase I) on December 18, 2001. Under the final rule, most new facilities could be expected to install recirculating cooling systems, primarily wet cooling towers. The EPA Administrator signed proposed 316(b) regulations for existing facilities (Phase II) on February 28, 2002. The lead option in this proposal would allow most existing facilities to achieve compliance without requiring them to convert once-through cooling systems to recirculating systems. However, one of the alternate options being proposed would require recirculating cooling in selected plants. EPA is considering various options to determine best technology available. Among the options under consideration are wet-cooling towers and dry-cooling towers. Both types of towers are considered to be part of recirculating cooling systems, in which the cooling water is continuously recycled from the condenser, where it absorbs heat by cooling and condensing steam, to the tower, where it rejects heat to the atmosphere before returning to the condenser. Some water is lost to evaporation (wet tower only) and other water is removed from the recirculating system as a blow down stream to control the building up of suspended and dissolved solids. Makeup water is withdrawn, usually from surface water bodies, to replace the lost water. The volume of makeup water is many times smaller than the volume needed to operate a once-through system. Although neither the final new facility rule nor the proposed existing facility rule require dry cooling towers as the national best technology available, the environmental community and several States have supported the use of dry-cooling technology as the appropriate technology for addressing adverse environmental impacts. It is possible that the requirements included in the new facility rule and the ongoing push for dry cooling systems by some stakeholders may have a role in shaping the rule for existing facilities. The temperature of the cooling water entering the condenser affects the performance of the turbine--the cooler the temperature, the better the performance. This is because the cooling water temperature affects the level of vacuum at the discharge of the steam turbine. As cooling water temperatures decrease, a higher vacuum can be produced and additional energy can be extracted. On an annual average, once-through cooling water has a lower temperature than recirculated water from a cooling tower. By switching a once-through cooling system to a cooling tower, less energy can be generated by the power plant from the same amount of fuel. This reduction in energy output is known as the energy penalty. If a switch away from once-through cooling is broadly implemented through a final 316(b) rule or other regulatory initiatives, the energy penalty could result in adverse effects on energy supplies. Therefore, in accordance with the recommendations of the Report of the National Energy Policy Development Group (better known as the May 2001 National Energy Policy), the U.S. Department of Energy (DOE), through its Office of Fossil Energy, National Energy Technology Laboratory (NETL), and Argonne National Laboratory (ANL), has studied the energy penalty resulting from converting plants with once-through cooling to wet towers or indirect-dry towers. Five l

Veil, J. A.; Littleton, D. J.; Gross, R. W.; Smith, D. N.; Parsons, E.L., Jr.; Shelton, W. W.; Feeley, T. J.; McGurl, G. V.

2006-11-27T23:59:59.000Z

376

Light Water Reactor Sustainability Program - Integrated Program Plan |  

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

Light Water Reactor Sustainability Program - Integrated Program Light Water Reactor Sustainability Program - Integrated Program Plan Light Water Reactor Sustainability Program - Integrated Program Plan The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U. S. Department of Energy (DOE), performed in close collaboration and cooperation with related industry R&D programs. The LWRS Program provides technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants, utilizing the unique capabilities of the national laboratory system. Sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer than-initially-licensed lifetime. It has two facets

377

Optimization of hydride fueled pressurized water reactor cores  

E-Print Network (OSTI)

This thesis contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This pursuit involves ...

Shuffler, Carter Alexander

2004-01-01T23:59:59.000Z

378

Corrosion-Product Release in Light Water Reactors  

Science Conference Proceedings (OSTI)

Colbalt released through corrosion is the primary source of radiation fields on out-of-core surfaces in pressurized water reactors. Lowering colbalt impurity levels in Inconel 600, a generator tubing material, could reduce radiation fields.

1984-03-01T23:59:59.000Z

379

Optimization of hybrid-water/air-cooled condenser in an enhanced turbine  

Open Energy Info (EERE)

Optimization of hybrid-water/air-cooled condenser in an enhanced turbine Optimization of hybrid-water/air-cooled condenser in an enhanced turbine geothermal ORC system Geothermal Project Jump to: navigation, search Last modified on July 22, 2011. Project Title Optimization of hybrid-water/air-cooled condenser in an enhanced turbine geothermal ORC system Project Type / Topic 1 Recovery Act: Enhanced Geothermal Systems Component Research and Development/Analysis Project Type / Topic 2 Air-Cooling Project Description The technical approaches are: -UTRC shall develop a lab-based analysis of hybrid-water/air-cooled condensers with minimal water consumption, focusing on combined mist evaporative pre-cooling and mist deluge evaporative cooling technology applied to microchannel heat exchangers. Models to predict evaporative cooling performance will be validated by sub-scale testing. The predicted performance will be compared to that of state-of-the-art commercial evaporative coolers. -UTRC shall analyze the interaction of turbine design and cooling needs and specifically address how an enhanced turbine, which features variable nozzles and diffuser boundary layer suction, would further improve the ORC system performance and enable full utilization of the hybrid-cooled system. UTRC shall design, procure and test the enhanced turbine in an existing 200 kW geothermal ORC system for a technology demonstration. -UTRC shall complete a detailed design of the hybrid-cooled geothermal ORC system with an enhanced turbine that complies with its performance, cost, and quality requirements, and use this system design to prescribe subsystem/component technology requirements and interfaces. UTRC shall optimize UTC's PureCycle® geothermal ORC system integrated with a hybrid-water/air-cooled condenser and an enhanced turbine for net power output, efficiency and water consumption. -UTRC shall analyze the feasibility of addressing pure water supply for hybrid-water/aircooled condenser by using geothermal-driven Liquid-Gap-Membrane-Distillation (LGMD) technology, as an alternative to conventional Reverse Osmosis/De-Ionized treatment.

380

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

Science Conference Proceedings (OSTI)

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

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

2007-09-01T23:59:59.000Z

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


381

Process for treating effluent from a supercritical water oxidation reactor  

DOE Patents (OSTI)

A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor.

Barnes, Charles M. (Idaho Falls, ID); Shapiro, Carolyn (Idaho Falls, ID)

1997-01-01T23:59:59.000Z

382

Process for treating effluent from a supercritical water oxidation reactor  

DOE Patents (OSTI)

A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor. 6 figs.

Barnes, C.M.; Shapiro, C.

1997-11-25T23:59:59.000Z

383

Overcoming Solubility Limitations to Zinc Addition in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Zinc addition to the reactor coolant system (RCS) of a pressurized water reactor (PWR) is being used for dose rate reduction and primary water stress corrosion cracking (PWSCC) mitigation. This report summarizes results of aqueous zinc oxide solubility experiments from 150 to 350 degrees Celsius (302 to 662 degrees Fahrenheit). These experiments were performed to develop quantitative models of solubility and aqueous-phase solute speciation behavior as functions of temperature, pH, and solution compositio...

2001-11-29T23:59:59.000Z

384

CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR  

Science Conference Proceedings (OSTI)

The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

Vinson, Dennis

2010-06-01T23:59:59.000Z

385

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

SciTech Connect

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

Sonat Sen; Gilles Youinou

2013-02-01T23:59:59.000Z

386

The Formation and Fate of Trihalomethanes in Power Plant Cooling Water Systems  

Science Conference Proceedings (OSTI)

Trihalomethanes (THMs) are semi-volatile compounds that form in water when chlorine or bromine reacts with dissolved organic carbon. This report investigates the formation and fate of THM compounds in power plant cooling water systems, THM health risks, a generalized mechanism of THM formation, and the applicability of existing THM research to power plant cooling. The report presents results of a two-site sampling and analytical program designed to identify THM formation potential in cooling towers using...

2004-03-16T23:59:59.000Z

387

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

Science Conference Proceedings (OSTI)

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

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

2007-08-13T23:59:59.000Z

388

Actinide minimization using pressurized water reactors  

E-Print Network (OSTI)

Transuranic actinides dominate the long-term radiotoxity in spent LWR fuel. In an open fuel cycle, they impose a long-term burden on geologic repositories. Transmuting these materials in reactor systems is one way to ease ...

Visosky, Mark Michael

2006-01-01T23:59:59.000Z

389

Rethinking the light water reactor fuel cycle  

E-Print Network (OSTI)

The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to ...

Shwageraus, Evgeni, 1973-

2004-01-01T23:59:59.000Z

390

CHEMICAL ASPECTS OF PELLET-CLADDING INTERACTION IN LIGHT WATER REACTOR FUEL ELEMENTS  

E-Print Network (OSTI)

ANS/ENS Topical Meeting on Reactor Safety Aspects of FuelINTERACTION IN LiaiT-WATER-REACTOR FUEL ELEMENTS by D. R.PCI) in light water reactor fuel elements, the chemical

Olander, D.R.

2010-01-01T23:59:59.000Z

391

Application and numerical simulation on water mist cooling for urban environment regulation  

Science Conference Proceedings (OSTI)

The fine water mist is a type of sustainable and environment-friendly cooling technology. This paper concerns the use of water mist flow to improve the quality of urban environment in summer. According to the survey and analysis on the potential for ... Keywords: numerical simulation, regulation of microclimate, spray cooling, two-phase flow

Junfeng Wang; Xincheng Tu; Zhentao Wang; Jiwei Huang

2010-09-01T23:59:59.000Z

392

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT  

SciTech Connect

Progress on reactor programs and in general engineering research and development programs is summarized. Research and development are reported on water-cooled reactors including EBWR and Borax-V, sodium-cooled reactors including ZPR-III, IV, and IX, Juggernaut, and EBR-I and II. Other work included a review of fast reactor technology, and studies on nuclear superheat, thermal and fast reactor safety, and reactor physics. Effort was also devoted to reactor materials and fuels development, heat engineering, separation processes and advanced reactor concepts. (J.R.D.)

1961-04-01T23:59:59.000Z

393

EPRI Materials Management Matrix Project: Advanced Light-Water Reactor - Pressurized Water Reactor Degradation Matrix - Revision 1  

Science Conference Proceedings (OSTI)

The Advanced Light Water Reactor - Pressurized Water Reactor Degradation Matrix (ALWR PWR DM) is an integral piece of the Electric Power Research Institutes (EPRIs) Materials Management Matrix (MMM) initiative for ALWR designs. The MMM provides a tool to assist the industry in proactive identification and consideration of materials issues and mitigation/management opportunities from the design phase through component fabrication and plant construction to operations and maintenance.

2010-09-22T23:59:59.000Z

394

Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme  

SciTech Connect

In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

Nur Asiah, A.; Su'ud, Zaki [Nuclear Physics and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Ferhat, A. [National Nuclear Energ Agency of Indonesia (BATAN) (Indonesia); Sekimoto, H. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

2010-06-22T23:59:59.000Z

395

REPORT OF THE OBJECTIVES AND PLANS FOR THE AEC'S CIVILIAN POWER GAS COOLED REACTOR PROGRAM  

SciTech Connect

Progress in the U. S. civilian power gas-cooled reactor program is discussed. Gas reactors having technical features of high conversion ratio, high temperature, high fuel burnup, and capability of construction in large sizes make them very attractive as potential producers of economic power in the very near term. The operation of Peach Bottom-HTGR and EGCR in late 1964 and 1965, respectively, will contribute to the successful exploitation of thermal gas- cooled reactors. Since the graphite fuel concept promises very low fuel cycle costs along with reactor coolant conditions that can exceed current practice, it was concluded that the concept provides a long term potential that promises some very exciting possibilities. (auth)

Pahler, R.E.

1963-06-01T23:59:59.000Z

396

Microsoft Word - INL_EXT-10-20208 DOE-Cooling Water Issues & Opportunities-Main Report-Rev.1.docx  

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

A Report to the U.S. Department of Energy A Report to the U.S. Department of Energy Office of Nuclear Energy December 2010 INL/EXT-10-20208 Revision 1 ii iii COOLING WATER ISSUES AND OPPORTUNITIES AT U.S. NUCLEAR POWER PLANTS A Report to the U.S. Department of Energy Office of Nuclear Energy Revision 1 December 2010 iv v PURPOSE This report has been prepared for the Department of Energy, Office of Light Water Reactor Technologies within DOE's Office of Nuclear Energy (DOE-NE), for the purpose of providing a status report on the challenges and opportunities facing the U.S. commercial nuclear energy industry in the area of plant cooling water supply. The report was prompted in part by recent Second Circuit and Supreme Court decisions regarding cooling water system designs at existing thermo-electric power generating facilities

397

Plutonium Recycling in Light Water Reactors at Framatome ANP: Status and Trends  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Use of Alternate Fuels in Light Water Reactors

Dieter Porsch; Walter Stach; Pascal Charmensat; Michel Pasquet

398

MARITIME GAS-COOLED REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING, SEPTEMBER 30, 1960  

SciTech Connect

The feasibility was studied and a cost estimate prepared of an experimental reactor to determine the operating characteristics of beryllia- moderated, gas-cooled systems wtthin a power limit of 10 Mw(t). The heat energy produced by the experimental reactor is to be dissipated in a heat dump. No machinery for production of power was to be provided. Other requirements were that the reactor should be capable of testing core types different from the current MGCR design, and the system should permit use of gases other than helium. It was further directed that the reactor should be designated BORE for Beryllium- Oxide Reactor Experiment. Reactor development work was mainly in connection with the BORE preliminary design. It was established that the most important information which could be provided by a 10 Mw(t) reactor experiment would be on performance of fuel elements and moderator bodies. This required that the experirment duplicate the power density in the fuel and moderator that would exist in the full size reactor and made it advisable to use full length fuel elements. This resulted in an unconventionally shaped core which is roughly cylindrical with the length more than twice its mean diameter. Studies continued on performance of fuel elements, and methods were developed for calculation of thermal stress in BeO-moderated modules. These studies are equally applicable to the MGCR prototype and BORE. A thermal analysis of the MGCR pressure vessel and thermal shields was performed and means of externally cooling the vessel were studied. Some of the components of the experimental control rod drive mechanism were received. Endurance tests of a ball-nut lead screw in hot helium continued. Heat exchanger tests were resumed after an interruption due to leaks in the tube to tubesheet joints in the test unit. Plant control studies were continued with analyses of the system dynamics with a turbomachinery configuration in which the high pressure turbine provides the output power. This arrangement was found to be more sensitive than the low-pressure drive system. Turbomachinery component tests are providing stage performance data. The seal and bearing test rig was completed and tests were begun. Physics calculations were made for the BORE design. Basic physics information on Fermi age and neutron thermalization in BeO was provided during the quarter by experlments in the linear accelerator. Materials work continued with further investigations of the effect of additives on properties of UC/sub 2/--BeO diluted fuel bodies. Hot cell examination of the MTR-31-3 fuel capsule indicated no significant dimensional changes after burnups from 30,000 to 45,000 Mwd/T. An experiment to determine irradiation effects on BeO was inserted in GETR during September. Work was started on development of high density BeO bodies. Structural materials research continued with completion of a second series of self-welding tests of different metal pairs in high temperature helium. A series of galling tests and a series of creep rupture tests of weld specimens in SA302-B steel were completed. Work on site development included establishing requirements of the BORE for buildings, auxiliary facilities, and utilities. (auth)

1960-09-30T23:59:59.000Z

399

Advanced water-cooled phosphoric acid fuel cell development  

DOE Green Energy (OSTI)

This program was conducted to improve the performance and minimize the cost of existing water-cooled phosphoric acid fuel cell stacks for electric utility and on-site applications. The goals for the electric utility stack technology were a power density of at least 175 watts per square foot over a 40,000-hour useful life and a projected one-of-a-kind, full-scale manufactured cost of less than $400 per kilowatt. The program adapted the existing on-site Configuration-B cell design to electric utility operating conditions and introduced additional new design features. Task 1 consisted of the conceptual design of a full-scale electric utility cell stack that meets program objectives. The conceptual design was updated to incorporate the results of material and process developments in Tasks 2 and 3, as well as results of stack tests conducted in Task 6. Tasks 2 and 3 developed the materials and processes required to fabricate the components that meet the program objectives. The design of the small area and 10-ft{sup 2} stacks was conducted in Task 4. Fabrication and assembly of the short stacks were conducted in Task 5 and subsequent tests were conducted in Task 6. The management and reporting functions of Task 7 provided DOE/METC with program visibility through required documentation and program reviews. This report describes the cell design and development effort that was conducted to demonstrate, by subscale stack test, the technical achievements made toward the above program objectives.

Not Available

1992-09-01T23:59:59.000Z

400

Advanced Water-Cooled Phosphoric Acid Fuel Cell Development  

DOE Green Energy (OSTI)

This program is being conducted to improve the performance and minimize the cost of water cooled, electric utility phosphoric acid fuel cell stacks. The program adapts the existing on-site Configuration B cell design to electric utility operating conditions and introduces additional new design features. Task 1 consists of the conceptual design of a full-scale electric utility cell stack that meets program objectives. Tasks 2 and 3 develop the materials and processes requested to fabricate the components that meet the program objective. The design of the small area and two 10-ft[sup 2] short stacks is conducted in Task 4. The conceptual design also is updated to incorporate the results of material and process developments, as well as results of stack tests conducted in Task 6. Fabrication and assembly of the short stacks are conducted in Task 5 and subsequent tests are conducted in Task 6. The Contractor expects to enter into a contract with the Electric Power Research Institute (EPRI) to assemble and endurance test the second 10-ft[sup 2] short stack. The management and reporting functions of Task 7 provide DOE/METC with program visibility through required documentation and program reviews. This report describes the cell design and development effort that is being conducted to demonstrate, by subscale stack test, the technical achievements made toward the above program objectives.

Not Available

1992-05-01T23:59:59.000Z

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While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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401

Cooling Water Issues and Opportunities at U.S. Nuclear Power Plants,  

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

Cooling Water Issues and Opportunities at U.S. Nuclear Power Cooling Water Issues and Opportunities at U.S. Nuclear Power Plants, December 2010 Cooling Water Issues and Opportunities at U.S. Nuclear Power Plants, December 2010 Energy and water are both essential to sustainable development and economic productivity. Ample supplies of water are essential to energy production, and water management is dependent on ample supplies of energy for water treatment and transportation. The critical nexus between energy and water has been recognized in a variety of recent studies, but the policy and regulatory machinery that this nexus depends on is not keeping up with the growing challenges. Population growth and societal demand for improved quality of life will require more clean water for drinking and sanitation, more water for

402

Benchmarking of the MIT High Temperature Gas-Cooled Reactor TRISO-Coated Particle Fuel Performance Model  

E-Print Network (OSTI)

309NUCLEAR ENGINEERING AND TECHNOLOGY, VOL.37 NO.4, AUGUST 2005 A NEW BOOK: "LIGHT-WATER REACTOR the Olander work [16], none of the books in the list of references was intended as a textbook. This usage engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from

403

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

Science Conference Proceedings (OSTI)

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

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

1998-03-01T23:59:59.000Z

404

2013 Interim Review of the Pressurized Water Reactor Secondary Water Chemistry Guideline Revision 7  

Science Conference Proceedings (OSTI)

As required by Nuclear Energy Institute (NEI) 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues), the Electric Power Research Institute (EPRI) periodically updates its Pressurized Water Reactor (PWR) water chemistry guidelines when new information becomes available. An industry review committee meeting in June 2013 determined that a revision of the 2009 version of EPRI’s Pressurized Water Reactor Secondary Water ...

2013-10-18T23:59:59.000Z

405

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

Science Conference Proceedings (OSTI)

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

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

2013-11-01T23:59:59.000Z

406

COOLING WATER ISSUES AND OPPORTUNITIES AT U.S. NUCLEAR POWER PLANTS  

SciTech Connect

This report has been prepared for the Department of Energy, Office of Nuclear Energy (DOE-NE), for the purpose of providing a status report on the challenges and opportunities facing the U.S. commercial nuclear energy industry in the area of plant cooling water supply. The report was prompted in part by recent Second Circuit and Supreme Court decisions regarding cooling water system designs at existing thermo-electric power generating facilities in the U.S. (primarily fossil and nuclear plants). At issue in the courts have been Environmental Protection Agency regulations that define what constitutes “Best Technology Available” for intake structures that withdraw cooling water that is used to transfer and reject heat from the plant’s steam turbine via cooling water systems, while minimizing environmental impacts on aquatic life in nearby water bodies used to supply that cooling water. The report was also prompted by a growing recognition that cooling water availability and societal use conflicts are emerging as strategic energy and environmental issues, and that research and development (R&D) solutions to emerging water shortage issues are needed. In particular, cooling water availability is an important consideration in siting decisions for new nuclear power plants, and is an under-acknowledged issue in evaluating the pros and cons of retrofitting cooling towers at existing nuclear plants. Because of the significant ongoing research on water issues already being performed by industry, the national laboratories and other entities, this report relies heavily on ongoing work. In particular, this report has relied on collaboration with the Electric Power Research Institute (EPRI), including its recent work in the area of EPA regulations governing intake structures in thermoelectric cooling water systems.

Gary Vine

2010-12-01T23:59:59.000Z

407

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

SciTech Connect

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

Phillip Mills

2012-02-01T23:59:59.000Z

408

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

SciTech Connect

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

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

2010-02-23T23:59:59.000Z

409

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

SciTech Connect

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

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

2011-03-01T23:59:59.000Z

410

EFFECTS OF SEISMIC VIBRATIONS ON THE EXPERIMENTAL GAS-COOLED REACTOR  

SciTech Connect

The effects of seismic vibrations on the dynamic behavior of a composite system were analyzed. The equations of motion were derived and soIved with special emphasis on determining the resulting stresses. The method of analysis thus developed was applied to the composite structure consisting of the core, pressure vessel, and supporting skirt of the Experimental Gas-Cooled Reactor (EGCR). A system with three degrees of freedom was considered in order to determine the effects of an earthquake of the maximum intensity expected in the area surrounding Oak Ridge, Tennessee. The system of equations of motion was solved both numerically and analytically, and the resonant frequencies were determined. The seismic effect was shown to be small when the frequency of the seismic disturbance coincided with a natural frequency of the system. In particular, the shear stresses in the graphite core were shown to be negligible. (auth)

Witt, F.J.; Carver, D.R.; Maxwell, R.L.

1962-06-22T23:59:59.000Z

411

Applications for a high temperature gas cooled nuclear reactor in oil shale processing  

SciTech Connect

Results are presented of a study concerning possible applications for a high temperature gas cooled reactor as a process heat source in oil shale retorting and upgrading. Both surface and in situ technologies were evaluated with respect to the applicability and potential benefits of introducing an outside heat source. The primary focus of the study was to determine the fossil resource which might be conserved, or freed for higher uses than furnishing process heat. In addition to evaluating single technologies, a centralized upgrading plant, which would hydrotreat the product from a 400,000 bbl/day regional shale oil industry was also evaluated. The process heat required for hydrogen manufacture via steam reforming, and for whole shale oil hydrotreating would be supplied by an HTGR. Process heat would be supplied where applicable, and electrical power would be generated for the entire industry.

Sinor, J.E.; Roe, D.E.

1980-01-01T23:59:59.000Z

412

High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics  

DOE Green Energy (OSTI)

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

Larry Demick

2011-08-01T23:59:59.000Z

413

Predicted nuclear heating and temperatures in gas-cooled nuclear reactors for process heat applications  

SciTech Connect

The high-temperature gas-cooled nuclear reactor (HTGR) is an attractive potential source of primary energy for many industrial and chemical process applications. Significant modification of current HTGR core design will be required to achieve the required elevations in exit gas temperatures without exceeding the maximum allowable temperature limits for the fuel material. A preliminary evaluation of the effects of various proposed design modifications by predicting the resulting fuel and gas temperatures with computer calculational modeling techniques is reported. The design modifications evaluated are generally those proposed by the General Atomic Company (GAC), a manufacturer of HTGRs, and some developed at the LASL. The GAC modifications do result in predicted fuel and exit gas temperatures which meet the proposed design objectives. (auth)

Cort, G.E.; Vigil, J.C.; Jiacoletti, R.J.

1975-09-01T23:59:59.000Z

414

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

E-Print Network (OSTI)

well known from basic reactor theory, the flux distributionof a fast reactor using the perturbation theory”. In: Atomicbeam theory and are not specific to a nuclear reactor core.

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

415

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

E-Print Network (OSTI)

and A. SESONSKE. Nuclear Reactor Engineering: Third Edition.E. LEWIS. Fundamentals of Nuclear Reactor Physics. Elseviervan DAM. “Physics of nuclear reactor safety”. In: Reports on

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

416

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

E-Print Network (OSTI)

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

Galvez, Cristhian

2011-01-01T23:59:59.000Z

417

Propellant actuated nuclear reactor steam depressurization valve  

DOE Patents (OSTI)

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.

Ehrke, Alan C. (San Jose, CA); Knepp, John B. (San Jose, CA); Skoda, George I. (Santa Clara, CA)

1992-01-01T23:59:59.000Z

418

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

SciTech Connect

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

Louison, R.; Boardman, C.E.

1981-05-29T23:59:59.000Z

419

System Engineering Program Applicability for the High Temperature Gas-Cooled Reactor (HTGR) Component Test Capability (CTC)  

SciTech Connect

This white paper identifies where the technical management and systems engineering processes and activities to be used in establishing the High Temperature Gas-cooled Reactor (HTGR) Component Test Capability (CTC) should be addressed and presents specific considerations for these activities under each CTC alternative

Jeffrey Bryan

2009-06-01T23:59:59.000Z

420

Life of Plant Activity Estimates for a Nominal 1000 MWe Pressurized Water Reactor and Boiling Water Reactor  

Science Conference Proceedings (OSTI)

Decommissioning nuclear power plant and disposal site managers must understand the radioactive source term of a nuclear power plant to effectively manage disposition of these materials. This study estimates the radioactive source term from nominal 1000 MWe pressurized water and boiling water reactors to support decisions related to radioactive waste storage, processing, and disposal through decommissioning.BackgroundThis study examines the radionuclide ...

2012-12-05T23:59:59.000Z

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


421

Scaling Analysis for the Direct Reactor Auxillary Cooling System For AHTRS  

Science Conference Proceedings (OSTI)

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

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

2013-01-01T23:59:59.000Z

422

Clean Water Act Section 316(b) Closed-Cycle Cooling Retrofit Research Program Results Summary  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) has investigated the implications of a potential U.S. Environmental Protection Agency (EPA) Clean Water Act 316(b) rulemaking if it establishes closed-cycle cooling retrofits for facilities with once-through cooling as "best technology available" (BTA) for fish protection. This report provides a summary of the results of five studies that comprise EPRI's Closed-Cycle Cooling Retrofit Research Program. These studies evaluated the cost, both financial and econom...

2011-08-15T23:59:59.000Z

423

Sampling Considerations for Monitoring Corrosion Products in the Reactor Coolant System in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Chemistry sampling of the reactor coolant system (RCS) of pressurized water reactors (PWRs) can provide significant information regarding the health of the primary system. Timely detection of increased corrosion product concentrations will aid in evaluating any risks associated with the onset of an axial offset anomaly, increased shutdown releases, increased out-of-core dose rates, or increased personnel doses. This report provides recommendations for improved RCS sampling.

2006-06-19T23:59:59.000Z

424

FEMP-Designated Product: Water-Cooled Ice Machines | Department of Energy  

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

FEMP-Designated Product: Water-Cooled Ice Machines FEMP-Designated Product: Water-Cooled Ice Machines FEMP-Designated Product: Water-Cooled Ice Machines October 7, 2013 - 11:11am Addthis Federal agencies are required by the National Energy Conservation Policy Act (P.L. 95-619), Executive Order 13423, Executive Order 13514, and Federal Acquisition Regulations (FAR) Subpart 23.2 and 53.223 to specify and buy ENERGY STAR® qualified products or, in categories not included in the ENERGY STAR program, FEMP designated products, which are among the highest 25% of equivalent products for energy efficiency. A PDF version of Water-Cooled Ice Machines is also available. Performance Requirements for Federal Purchases Type Ice Harvest Rate (pounds per 24 hours) Energy Usea (per 100 pounds) Potable Water Useb (per 100 pounds)

425

Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors  

SciTech Connect

Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

2005-10-01T23:59:59.000Z

426

2012 Interim Review of the Pressurized Water Reactor Secondary Water Chemistry Guideline Revision 7  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) periodically updates its Pressurized Water Reactor (PWR) water chemistry guidelines as new information becomes available and as required by Nuclear Energy Institute (NEI) 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). An industry review committee meeting in September 2012 determined that a revision of the 2009 version of EPRI’s Pressurized Water Reactor Secondary ...

2012-12-21T23:59:59.000Z

427

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

E-Print Network (OSTI)

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

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

428

REFLECTOR CONTROL OF A BOILING-WATER REACTOR  

DOE Patents (OSTI)

A line connecting the reactor with a spent steam condenser contains a valve set to open when the pressure in the reactor exceeds a predetermined value and an orifice on the upstream side of the valve. Another line connects the reflector with this line between the orifice and the valve. An excess steam pressure causes the valve to open and the flow of steam through the line draws water out of the reflector. Provision is also made for adding water to the reflector when the steam pressure drops. (AEC)

Treshow, M.

1962-05-22T23:59:59.000Z

429

Boiling Water Reactor (BWR) Zinc Injection Strategy Evaluation  

Science Conference Proceedings (OSTI)

All U.S. boiling water reactors (BWRs) inject depleted zinc oxide (DZO) into the reactor feedwater for the purpose of suppressing drywell shutdown radiation dose rates. Current guidance in BWRVIP-190: BWR Vessel and Internals Project, BWR Water Chemistry Guidelines2008 Revision (EPRI report 1016579) is to inject sufficient zinc to achieve a Co-60(s)/Zn(s) ratio of Utility-specific goals may encourage even lower Co-60(s)/Zn(s) levels. This may be in part because BWR e...

2010-11-24T23:59:59.000Z

430

Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability  

SciTech Connect

Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment. • Reactor pressure vessel • Pumps and piping

Philip E. MacDonald

2003-09-01T23:59:59.000Z

431

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

Science Conference Proceedings (OSTI)

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

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

2010-01-01T23:59:59.000Z

432

Heat Transfer Performance and Piping Strategy Study for Chilled Water Systems at Low Cooling Loads  

E-Print Network (OSTI)

The temperature differential of chilled water is an important factor used for evaluating the performance of a chilled water system. A low delta-T may increase the pumping energy consumption and increase the chiller energy consumption. The system studied in this thesis is the chilled water system at the Dallas/Fort Worth International Airport (DFW Airport). This system has the problem of low delta-T under low cooling loads. When the chilled water flow is much lower than the design conditions at low cooling loads, it may lead to the laminar flow of the chilled water in the cooling coils. The main objective of this thesis is to explain the heat transfer performance of the cooling coils under low cooling loads. The water side and air side heat transfer coefficients at different water and air flow rates are calculated. The coefficients are used to analyze the heat transfer performance of the cooling coils at conditions ranging from very low loads to design conditions. The effectiveness-number of transfer units (NTU) method is utilized to analyze the cooling coil performance under different flow conditions, which also helps to obtain the cooling coil chilled water temperature differential under full load and partial load conditions. When the water flow rate drops to 1ft/s, laminar flow occurs; this further decreases the heat transfer rate on the water side. However, the cooling coil effectiveness increases with the drop of water flow rate, which compensates for the influence of the heat transfer performance under laminar flow conditions. Consequently, the delta-T in the cooling coil decreases in the transitional flow regime but increases in the laminar flow regime. Results of this thesis show that the laminar flow for the chilled water at low flow rate is not the main cause of the low delta-T syndrome in the chilled water system. Possible causes for the piping strategy of the low delta-T syndrome existing in the chilled water system under low flow conditions are studied in this thesis: (1) use of two way control valves; and (2) improper tertiary pump piping strategy.

Li, Nanxi 1986-

2012-12-01T23:59:59.000Z

433

Thermal-Structural Design of a Water Shield For Surface Reactor Missions  

SciTech Connect

Water shielding is an attractive option for an affordable lunar surface fission reactor program. The attractiveness of the water shielding option arises from the relative ease of proto-typing and ground testing, the relatively low development effort needed, as well as the fabrication and operating experience with stainless steel and water. The most significant limitation in using a water shield is temperature: to prevent the formation of voids and the consequent loss of cooling, the water temperature has to be maintained below the saturation temperature corresponding to the shield pressure. This paper examines natural convection for a prototypic water shield design using the computational fluid dynamics (CFD) code CFX-5 as well as analytical modeling. The results show that natural convection is adequate to keep the water well-mixed. The results also show that for the above-ground configuration, shield surface and water temperatures during lunar day conditions are high enough to require shield pressures up to 2.5 atm to prevent void formation. For the buried configuration, a set of ammonia heat pipes attached to the shield outer wall can be used to maintain water temperatures within acceptable limits. Overall the results show that water shielding is feasible for lunar surface applications. The results of the CFD analyses can also be used to guide development of testing plans for shield thermal testing. (authors)

Sadasivan, Pratap; Kapernick, Richard J.; Poston, David I. [D-5 Nuclear Systems Design Group MS K575, Los Alamos National Laboratory, Los Alamos, New Mexico, 87545 (United States)

2006-07-01T23:59:59.000Z

434

A Numerical Modeling Study of Warm Offshore Flow over Cool Water  

Science Conference Proceedings (OSTI)

Numerical simulations of boundary layer evolution in offshore flow of warm air over cool water are conducted and compared with aircraft observations of mean and turbulent fields made at Duck, North Carolina. Two models are used: a two-dimensional,...

Eric D. Skyllingstad; Roger M. Samelson; Larry Mahrt; Phil Barbour

2005-02-01T23:59:59.000Z

435