Science.gov

Sample records for coolants desik slutrapportering

  1. Coolant line hydrometer

    SciTech Connect

    Barber, M.D.; Kipp, W.G.

    1987-03-17

    This patent describes a hydrometer unit for connection in an automobile coolant flow line comprising: a tubular fitting adapted to be connected to the coolant flow line; a coolant receiving chamber means connected to the tubular fitting for receiving coolant from the tubular fitting; and indicating float elements contained within the coolant receiving chamber means and adapted to rise therein individually as a function of the specific gravity of the coolant. The coolant receiving chamber means includes a closure cap which when connected to the tubular fitting forms a coolant receiving chamber, retaining means for retaining the indicating float elements within the coolant receiving chamber, a viewing window member of a substantially clear material through which the float elements can be visually observed within the coolant receiving chamber means, and air venturi means located within the coolant receiving chamber means for automatically removing air which may collect within the coolant chamber means.

  2. Machine coolant waste reduction by optimizing coolant life. Project summary

    SciTech Connect

    Pallansch, J.

    1995-08-01

    The project was designed to study the following: A specific water-soluble coolant (Blasocut 2000 Universal) in use with a variety of machines, tools, and materials; Coolant maintenance practices associated with three types of machines; Health effects of use and handling of recycled coolant; Handling practices for chips and waste coolant; Chip/coolant separation; and Oil/water separation.

  3. NUCLEAR REACTOR COOLANT

    DOEpatents

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolytic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 5% of beryllium or magnesium dispersed in the hydrocarbon.

  4. NUCLEAR REACTOR COOLANT

    DOEpatents

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolitic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 10% of an alkall metal dispersed in the hydrocarbon.

  5. MACHINE COOLANT WASTE REDUCTION BY OPTIMIZING COOLANT LIFE

    EPA Science Inventory

    Machine shops use coolants to improve the life and function of machine tools. hese coolants become contaminated with oils with use, and this contamination can lead to growth of anaerobic bacteria and shortened coolant life. his project investigated methods to extend coolant life ...

  6. Environmentally Friendly Coolant System

    SciTech Connect

    David Jackson Principal Investigator

    2011-11-08

    Energy reduction through the use of the EFCS is most improved by increasing machining productivity. Throughout testing, nearly all machining operations demonstrated less land wear on the tooling when using the EFCS which results in increased tool life. These increases in tool life advance into increased productivity. Increasing productivity reduces cycle times and therefore reduces energy consumption. The average energy savings by using the EFCS in these machining operations with these materials is 9%. The advantage for end milling stays with flood coolant by about 6.6% due to its use of a low pressure pump. Face milling and drilling are both about 17.5% less energy consumption with the EFCS than flood coolant. One additional result of using the EFCS is improved surface finish. Certain machining operations using the EFCS result in a smoother surface finish. Applications where finishing operations are required will be able to take advantage of the improved finish by reducing the time or possibly eliminating completely one or more finishing steps and thereby reduce their energy consumption. Some machining operations on specific materials do not show advantages for the EFCS when compared to flood coolants. More information about these processes will be presented later in the report. A key point to remember though, is that even with equivalent results, the EFCS is replacing petroleum based coolants whose production produces GHG emissions and create unsafe work environments.

  7. Reactor coolant pump flywheel

    SciTech Connect

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  8. Proposed reactor coolant density monitor

    SciTech Connect

    Mackley, A.D.

    1986-01-01

    Until now there has been no feasible method of monitoring coolant density in the environment of an operating reactor core. By analysis of output from self-powered neutron detectors (SPNDs) in the core of the Loss of Fluid Test (LOFT) Reactor, the author has successfully estimated local coolant densities under post-scram conditions during a large break loss of coolant transient. The model used for estimation is not fully explained by published principles on the interaction of gamma rays with SPNDs. However, based on the success of the model, the author proposes employing self powered gamma detectors (SPGDs) to monitor reactor coolant density and discusses areas of experimental work to establish the best conditions for this application. 9 refs., 12 figs.

  9. Treatment of mixed waste coolant

    SciTech Connect

    Kidd, S.; Bowers, J.S.

    1995-09-01

    The primary processes used at Lawrence Livermore National Laboratory (LLNL) for treatment of radioactively contaminated machine coolants are industrial waste treatment and in situ carbon adsorption. These two processes simplify approaches to meetings the sanitary sewer discharge limits and subsequent Land Disposal REstriction criteria for hazardous and mixed wastes (40 CFR 268). Several relatively simple technologies are used in industrial water treatment. These technologies are considered {open_quotes}Best Demonstrated Available Technologies,{close_quotes} or BDAT, by the Environmental Protection Agency. The machine coolants are primarily aqueous and contain water soluble oil consisting of ethanol amine emulsifiers derived from fatty acids, both synthetic and natural. This emulsion carries away metal turnings from a part being machined on a lathe or other machining tool. When the coolant becomes spent, it contains chlorosolvents carried over from other cutting operations as well as a fair amount of tramp oil from machine bearings. This results in a mutiphasic aqueous waste that requires treatment of metal and organic contaminants. During treatment, any dissolved metals are oxidized with hydrogen peroxide. Once oxidized, these metals are flocculated with ferric sulfate and precipitated with sodium hydroxide, and then the precipitate is filtered through diatomaceous earth. The emulsion is broken up by acidifying the coolant. Solvents and oils are adsorbed using powdered carbon. This carbon is easily separated from the remaining coolant by vacuum filtration.

  10. Treatment of mixed waste coolant

    SciTech Connect

    Kidd, S.; Bowers, J.S.

    1995-02-01

    The primary processes used at Lawrence Livermore National Laboratory (LLNL) for treatment of radioactively contaminated machine coolants are industrial waste treatment and in situ carbon adsorption. These two processes simplify approaches to meeting the sanitary sewer discharge limits and subsequent Land Disposal Restriction criteria for hazardous and mixed wastes (40 CFR 268). Several relatively simple technologies are used in industrial water treatment. These technologies are considered Best Demonstrated Available Technologies, or BDAT, by the Environmental Protection Agency. The machine coolants are primarily aqueous and contain water soluble oil consisting of ethanol amine emulsifiers derived from fatty acids, both synthetic and natural. This emulsion carries away metal turnings from a part being machined on a lathe or other machining tool. When the coolant becomes spent, it contains chlorosolvents carried over from other cutting operations as well as a fair amount of tramp oil from machine bearings. This results in a multiphasic aqueous waste that requires treatment of metal and organic contaminants. During treatment, any dissolved metals are oxidized with hydrogen peroxide. Once oxidized, these metals are flocculated with ferric sulfate and precipitated with sodium hydroxide, and then the precipitate is filtered through diatomaceous earth. The emulsion is broken up by acidifying the coolant. Solvents and oils are adsorbed using powdered carbon. This carbon is easily separated from the remaining coolant by vacuum filtration.

  11. 1996 Coolant Flow Management Workshop

    NASA Technical Reports Server (NTRS)

    Hippensteele, Steven A. (Editor)

    1997-01-01

    The following compilation of documents includes a list of the 66 attendees, a copy of the viewgraphs presented, and a summary of the discussions held after each session at the 1996 Coolant Flow Management Workshop held at the Ohio Aerospace Institute, adjacent to the NASA Lewis Research Center, Cleveland, Ohio on December 12-13, 1996. The workshop was organized by H. Joseph Gladden and Steven A. Hippensteele of NASA Lewis Research Center. Participants in this workshop included Coolant Flow Management team members from NASA Lewis, their support service contractors, the turbine engine companies, and the universities. The participants were involved with research projects, contracts and grants relating to: (1) details of turbine internal passages, (2) computational film cooling capabilities, and (3) the effects of heat transfer on both sides. The purpose of the workshop was to assemble the team members, along with others who work in gas turbine cooling research, to discuss needed research and recommend approaches that can be incorporated into the Center's Coolant Flow Management program. The workshop was divided into three sessions: (1) Internal Coolant Passage Presentations, (2) Film Cooling Presentations, and (3) Coolant Flow Integration and Optimization. Following each session there was a group discussion period.

  12. Long life coolant pump technology

    NASA Technical Reports Server (NTRS)

    1976-01-01

    Design concepts were investigated to improve space system coolant pump technology to be suitable for mission durations of two years and greater. These design concepts included an improved bearing system for the pump rotating elements, consisting of pressurized conical bearings. This design was satisfactorily endurance tested as was a new prototype pump built using various other improved design concepts. Based upon an overall assessment of the results of the program it is concluded that reliable coolant pumps can be designed for three year space missions.

  13. Vertical reactor coolant pump instabilities

    NASA Technical Reports Server (NTRS)

    Jones, R. M.

    1985-01-01

    The investigation conducted at the Tennessee Valley Authority's Sequoyah Nuclear Power Plant to determine and correct increasing vibrations in the vertical reactor coolant pumps is described. Diagnostic procedures to determine the vibration causes and evaluate the corrective measures taken are also described.

  14. INHIBITING THE POLYMERIZATION OF NUCLEAR COOLANTS

    DOEpatents

    Colichman, E.L.

    1959-10-20

    >The formation of new reactor coolants which contain an additive tbat suppresses polymerization of the primary dissoclation free radical products of the pyrolytic and radiation decomposition of the organic coolants is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to 5% of a powdered metal hydride chosen from the group consisting of the group IIA and IVA dispersed in the hydrocarbon.

  15. Organic coolant for ARIES-III

    SciTech Connect

    Sze, D.K. ); Sviatoslavsky, I.; Sawan, M. ); Gierszewski, P. ); Hollies, R. ); Sharafat, S. ); Herring, S. )

    1991-04-01

    ARIES-III is a D-He{sub 3} reactor design study. It is found that the organic coolant is well suited for the D-He{sub 3} reactor. This paper discusses the unique features of the D-He{sub 3} reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs.

  16. Cleaning of uranium vs machine coolant formulations

    SciTech Connect

    Cristy, S.S.; Byrd, V.R.; Simandl, R.F.

    1984-10-01

    This study compares methods for cleaning uranium chips and the residues left on chips from alternate machine coolants based on propylene glycol-water mixtures with either borax, ammonium tetraborate, or triethanolamine tetraborate added as a nuclear poison. Residues left on uranium surfaces machined with perchloroethylene-mineral oil coolant and on surfaces machined with the borax-containing alternate coolant were also compared. In comparing machined surfaces, greater chlorine contamination was found on the surface of the perchloroethylene-mineral oil machined surfaces, but slightly greater oxidation was found on the surfaces machined with the alternate borax-containing coolant. Overall, the differences were small and a change to the alternate coolant does not appear to constitute a significant threat to the integrity of machined uranium parts.

  17. Flow boiling test of GDP replacement coolants

    SciTech Connect

    Park, S.H.

    1995-08-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C{sub 4}F{sub 10} and C{sub 4}F{sub 8}, were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C{sub 4}F{sub 10} mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C{sub 4}F{sub 10} weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd.

  18. Bi-coolant flat plate solar collector

    NASA Astrophysics Data System (ADS)

    Chon, W. Y.; Green, L. L.

    The feasibility study of a flat plate solar collector which heats air and water concurrently or separately was carried out. Air flows above the collector absorber plate, while water flows in tubes soldered or brazed beneath the plate. The collector efficiencies computed for the flow of both air and water are compared with those for the flow of a single coolant. The results show that the bi-coolant collector efficiency computed for the entire year in Buffalo, New York is higher than the single-coolant collector efficiency, although the efficiency of the water collector is higher during the warmer months.

  19. Coolant mass flow equalizer for nuclear fuel

    DOEpatents

    Betten, Paul R.

    1978-01-01

    The coolant mass flow distribution in a liquid metal cooled reactor is enhanced by restricting flow in sub-channels defined in part by the peripheral fuel elements of a fuel assembly. This flow restriction, which results in more coolant flow in interior sub-channels, is achieved through the use of a corrugated liner positioned between the bundle of fuel elements and the inner wall of the fuel assembly coolant duct. The corrugated liner is expandable to accommodate irradiation induced growth of fuel assembly components.

  20. Computing Flows Of Coolants In Turbomachines

    NASA Technical Reports Server (NTRS)

    Meitner, P. L.

    1994-01-01

    Coolant Passage Flow (CPF) computer code developed to predict accurately coolant flow and heat transfer inside turbomachinery cooling passages (either radial or axial blading). Computes flow in one-inlet/one-outlet passage of any shape. Calculates rate of flow of coolant, temperature, pressure, velocity, and heat-transfer coefficients along passage. Integrates one-dimensional momentum and energy equations along defined flow path, taking into account change in area, addition or subtraction of mass, pumping, friction, and transfer of heat. Written in FORTRAN IV.

  1. Coolant passage heat transfer with rotation

    NASA Astrophysics Data System (ADS)

    Hajek, T. J.; Higgins, A. W.

    1985-10-01

    The objective is to develop a heat transfer and pressure drop data base, computational fluid dynamic techniques, and correlations for multi-pass rotating coolant passages with and without flow turbulators. The experimental effort is focused on the simulation of configurations and conditions expected in the blades of advanced aircraft high pressure turbines. With the use of this data base, the effects of Coriolis and buoyancy forces on the coolant side flow can be included in the design of turbine blades.

  2. Nuclear reactor loss of coolant protection system

    SciTech Connect

    Loose, R.A.

    1986-03-18

    A pressurized water reactor system is described of a nuclear power plant having a water storage tank for providing emergency coolant water and means provided external to the containment vessel, for use in the event of a primary loss of coolant situation, to circulate emergency water as a coolant by withdrawal through a wall of the containment vessel and return the same back through the wall of the containment vessel and passing the water through a heat exchange means prior to use as a coolant for the reactor core. The improvement described here consists of: an enslosure, the interior of which is sealed to the atmosphere, positioned adjacent to and exterior of a wall of the containment vessel; an inlet conduit, enclosed within a sealed outer casing, communicating between the interior of the containment vessel and the interior of the enclosure; an exhaust conduit, enclosed within a sealed outer casing, communicating between the interior of the enclosure and the interior of the containment vessel; a rupture disc on the inlet conduit within the enclosure, such that failure of the exhaust conduit within the enclosure will produce an increase of the pressure within the enclosure and above a predetermined pressure will fracture the rupture disc, and will circulate the coolant within the enclosure; and means within the interior of the enclosure for pumping coolant from the interior of the containment vessel through the inlet conduit, and back to the interior of the containment vessel through the exhaust conduit; whereby if either of the conduits should fail, coolant will be collected within the enclosure and sealed to the atmosphere.

  3. Molecular Design for Cryogenic Magnetic Coolants.

    PubMed

    Liu, Jun-Liang; Chen, Yan-Cong; Tong, Ming-Liang

    2016-04-01

    The area of molecular magnetic coolants has developed rapidly in recent years. A large number of competitive candidates have been reported, with the cooling performances chasing each other. In this account, four explicit strategies, namely, increasing ground-state spin, reducing magnetic anisotropy, weakening magnetic interactions, and lowering the molecular weight, are proposed from the theoretical viewpoint towards improving the magnetocaloric effect (MCE). According to this guidance, these successful strategies are discussed to pursue excellent magnetic coolants. This is accompanied by a discussion of the representative examples reported by our group. The magnetic entropy change increases from one compound to another, which in the most pronounced cases is suggestive of being the largest MCE in magnetic coolants. PMID:26929130

  4. Lead Coolant Test Facility Development Workshop

    SciTech Connect

    Paul A. Demkowicz

    2005-06-01

    A workshop was held at the Idaho National Laboratory on May 25, 2005, to discuss the development of a next generation lead or lead-alloy coolant test facility. Attendees included representatives from the Generation IV lead-cooled fast reactor (LFR) program, Advanced Fuel Cycle Initiative, and several universities. Several participants gave presentations on coolant technology, existing experimental facilities for lead and lead-alloy research, the current LFR design concept, and a design by Argonne National Laboratory for an integral heavy liquid metal test facility. Discussions were focused on the critical research and development requirements for deployment of an LFR demonstration test reactor, the experimental scope of the proposed coolant test facility, a review of the Argonne National Laboratory test facility design, and a brief assessment of the necessary path forward and schedule for the initial stages of this development project. This report provides a summary of the presentations and roundtable discussions.

  5. On-Line Coolant Chemistry Analysis

    SciTech Connect

    LM Bachman

    2006-07-19

    Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level.

  6. Coolant monitoring apparatus for nuclear reactors

    DOEpatents

    Tokarz, Richard D.

    1983-01-01

    A system for monitoring coolant conditions within a pressurized vessel. A length of tubing extends outward from the vessel from an open end containing a first line restriction at the location to be monitored. The flowing fluid is cooled and condensed before passing through a second line restriction. Measurement of pressure drop at the second line restriction gives an indication of fluid condition at the first line restriction. Multiple lengths of tubing with open ends at incremental elevations can measure coolant level within the vessel.

  7. NGNP Reactor Coolant Chemistry Control Study

    SciTech Connect

    Brian Castle

    2010-11-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  8. Design criteria for Waste Coolant Processing Facility and preliminary proposal 722 for Waste Coolant Processing Facility

    SciTech Connect

    Not Available

    1991-09-27

    This document contains the design criteria to be used by the architect-engineer (A-E) in the performance of Titles 1 and 2 design for the construction of a facility to treat the biodegradable, water soluble, waste machine coolant generated at the Y-12 plant. The purpose of this facility is to reduce the organic loading of coolants prior to final treatment at the proposed West Tank Farm Treatment Facility.

  9. Coolant passage heat transfer with rotation

    NASA Astrophysics Data System (ADS)

    Hajek, T. J.; Wagner, J.; Johnson, B. V.

    1986-10-01

    In current and advanced gas turbine engines, increased speeds, pressures and temperatures are used to reduce specific fuel consumption and increase thrust/weight ratios. Hence, the turbine airfoils are subjected to increased heat loads escalating the cooling requirements to satisfy life goals. The efficient use of cooling air requires that the details of local geometry and flow conditions be adequately modeled to predict local heat loads and the corresponding heat transfer coefficients. The objective of this program is to develop a heat transfer and pressure drop data base, computational fluid dynamic techniques and correlations for multi-pass rotating coolant passages with and without flow turbulators. The experimental effort is focused on the simulation of configurations and conditions expected in the blades of advanced aircraft high pressure turbines. With the use of this data base, the effects of Coriolis and buoyancy forces on the coolant side flow can be included in the design of turbine blades.

  10. Evaluation of engine coolant recycling processes: Part 2

    SciTech Connect

    Bradley, W.H.

    1999-08-01

    Engine coolant recycling continues to provide solutions to both economic and environmental challenges often faced with the disposal of used engine coolant. General Motors` Service Technology Group (STG), in a continuing effort to validate the general practice of recycling engine coolants, has conducted an in-depth study on the capabilities of recycled coolants. Various recycling processes ranging from complex forms of fractional distillation to simple filtration were evaluated in this study to best represent the current state of coolant recycling technology. This study incorporates both lab and (limited) fleet testing to determine the performance capabilities of the recycled coolants tested. While the results suggest the need for additional studies in this area, they reveal the true capabilities of all types of engine coolant recycling technologies.

  11. Cryogenic-coolant He-4-superconductor interaction

    NASA Technical Reports Server (NTRS)

    Caspi, S.; Lee, J. Y.; Kim, Y. I.; Allen, R. J.; Frederking, T. H. K.

    1978-01-01

    The thermodynamic and thermal interaction between a type 2 composite alloy and cryo-coolant He4 was studied with emphasis on post quench phenomena of formvar coated conductors. The latter were investigated using a heater simulation technique. Overall heat transfer coefficients were evaluated for the quench onset point. Heat flux densities were determined for phenomena of thermal switching between a peak and a recovery value. The study covered near saturated liquid, pressurized He4, both above and below the lambda transition, and above and below the thermodynamic critical pressure. In addition, friction coefficients for relative motion between formvar insulated conductors were determined.

  12. Secondary coolant circuit for nuclear-reactors

    SciTech Connect

    Brachet, A.

    1981-10-06

    A secondary coolant circuit for a nuclear-reactor of the liquid metal cooled type is described. The circuit comprises at least one intermediate exchanger mounted in the vessel of said reactor, Also included is a steam-generator for the exchange of calories between the secondary liquid-metal flowing through said secondary circuit and water-steam, at least one pump for circulating said secondary sodium and one tank for storing said secondary liquid-metal andrecovering those products generated by a possible liquid-metal-water reaction in said steam-generator.

  13. Nuclear fuel assembly with coolant conducting tube

    SciTech Connect

    Dunlap, T. G.; Cearley, J. E.; Jameson, W. G. Jr.; Mefford, C. R.; Nelson, H. L.

    1983-12-13

    In a nuclear fuel assembly having a coolant conducting or water tube which also retains the spacers in axial position, the fuel rods experience greater axial growth with exposure than the water tube creating a risk that the water tube might become disengaged from the supporting tie plates. An arrangement for preventing such disengagement is described including lengthened end plug shanks for the water tube, a protective boss surrounding the lower end plug shank to protect it from flow induced vibration, a conical seat for the lower end plug and an arrangement for limiting upward movement of the water tube.

  14. A Heated Tube Facility for Rocket Coolant Channel Research

    NASA Technical Reports Server (NTRS)

    Green, James M.; Pease, Gary M.; Meyer, Michael L.

    1995-01-01

    The capabilities of a heated tube facility used for testing rocket engine coolant channels at the NASA Lewis Research Center are presented. The facility uses high current, low voltage power supplies to resistively heat a test section to outer wall temperatures as high as 730 C (1350 F). Liquid or gaseous nitrogen, gaseous helium, or combustible liquids can be used as the test section coolant. The test section is enclosed in a vacuum chamber to minimize heat loss to the surrounding system. Test section geometry, size, and material; coolant properties; and heating levels can be varied to generate heat transfer and coolant performance data bases.

  15. Transpiration cooling using air as a coolant

    SciTech Connect

    Kikkawa, Shinzo; Senda, Mamoru; Sakagushi, Katsuji; Shibutani, Hideki )

    1993-02-01

    Transpiration cooling is one of the most effective techniques for protecting a surface exposed to a high-temperature gas stream. In the present paper, the transpiration cooling effectiveness was measured under steady state. Air as a coolant was transpired from the surface of a porous plate exposed to hot gas stream, and the transpiration rate was varied in the range of 0.001 [approximately] 0.006. The transpiration cooling effectiveness was evaluated by measuring the temperature of the upper surface of the plate. Also, a theoretical study was performed and it was clarified that the effectiveness increases with increasing transpiration rate and heat-transfer coefficient of the upper surface. Further, the effectiveness was expressed as a function of the blowing parameter only. The agreement between the experimental results and theoretical ones was satisfactory.

  16. Testing of organic acids in engine coolants

    SciTech Connect

    Weir, T.W.

    1999-08-01

    The effectiveness of 30 organic acids as inhibitors in engine coolants is reported. Tests include glassware corrosion of coupled and uncoupled metals. FORD galvanostatic and cyclic polarization electrochemistry for aluminum pitting, and reserve alkalinity (RA) measurements. Details of each test are discussed as well as some general conclusions. For example, benzoic acid inhibits coupled metals well but is ineffective on cast iron when uncoupled. In benzoic acid inhibits coupled metals well but is ineffective on cast iron when uncoupled. In general, the organic acids provide little RA when titrated to a pH of 5.5, titration to a pH of 4.5 can result in precipitation of the acid. Trends with respect to acid chain length are reported also.

  17. Power module assemblies with staggered coolant channels

    DOEpatents

    Herron, Nicholas Hayden; Mann, Brooks S; Korich, Mark D

    2013-07-16

    A manifold is provided for supporting a power module assembly with a plurality of power modules. The manifold includes a first manifold section. The first face of the first manifold section is configured to receive the first power module, and the second face of the first manifold section defines a first cavity with a first baseplate thermally coupled to the first power module. The first face of the second manifold section is configured to receive the second power module, and the second face of the second manifold section defines a second cavity with a second baseplate thermally coupled to the second power module. The second face of the first manifold section and the second face of the second manifold section are coupled together such that the first cavity and the second cavity form a coolant channel. The first cavity is at least partially staggered with respect to second cavity.

  18. 73. View of line of stainless steel coolant storage tanks ...

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

    73. View of line of stainless steel coolant storage tanks for bi-sodium sulfate/water coolant solution at first floor of transmitter building no. 102. - Clear Air Force Station, Ballistic Missile Early Warning System Site II, One mile west of mile marker 293.5 on Parks Highway, 5 miles southwest of Anderson, Anderson, Denali Borough, AK

  19. Automatic coolant flow control device for a nuclear reactor assembly

    DOEpatents

    Hutter, E.

    1984-01-27

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  20. Effects of rotation on coolant passage heat transfer. Volume 1: Coolant passages with smooth walls

    NASA Technical Reports Server (NTRS)

    Hajek, T. J.; Wagner, J. H.; Johnson, B. V.; Higgins, A. W.; Steuber, G. D.

    1991-01-01

    An experimental program was conducted to investigate heat transfer and pressure loss characteristics of rotating multipass passages, for configurations and dimensions typical of modern turbine blades. The immediate objective was the generation of a data base of heat transfer and pressure loss data required to develop heat transfer correlations and to assess computational fluid dynamic techniques for rotating coolant passages. Experiments were conducted in a smooth wall large scale heat transfer model.

  1. Radiant energy receiver having improved coolant flow control means

    DOEpatents

    Hinterberger, H.

    1980-10-29

    An improved coolant flow control for use in radiant energy receivers of the type having parallel flow paths is disclosed. A coolant performs as a temperature dependent valve means, increasing flow in the warmer flow paths of the receiver, and impeding flow in the cooler paths of the receiver. The coolant has a negative temperature coefficient of viscosity which is high enough such that only an insignificant flow through the receiver is experienced at the minimum operating temperature of the receiver, and such that a maximum flow is experienced at the maximum operating temperature of the receiver. The valving is accomplished by changes in viscosity of the coolant in response to the coolant being heated and cooled. No remotely operated valves, comparators or the like are needed.

  2. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

    NASA Technical Reports Server (NTRS)

    Morrison, Russell H.; Holt, Mike

    2005-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

  3. Reclamation and disposal of water-based machining coolants

    SciTech Connect

    Taylor, P.A.

    1982-01-01

    The Oak Ridge Y-12 Plant, which is operated by the Union Carbide Corporation, Nuclear Division for the Department of Energy under US government contract W-7405-eng-26, currently uses about 10{sup 6} L/yr (260,000 gal/yr) of water-based coolants in its machining operations. These coolants are disposed of in a 110,000-L (29,000-gal) activated sludge reactor. The reactor has oxidized an average of 38.6 kg of total organic carbon (TOC) per day with an overall efficiency of 90%. The predominant bacteria in the reactor have been identified once each year for the past three years. Six primary types of water-based coolants are currently used in the machine shops. In order to reduce the coolant usage rate, efforts are being made to introduce one universal coolant into the shops. By using a biocide to limit bacterial deterioration and using a filter and centrifuge system to remove dirt and tramp oils from the coolant, the coolant discard rate can be greatly reduced. 1 tab.

  4. Reactor coolant pump monitoring and diagnostic system

    SciTech Connect

    Singer, R.M.; Gross, K.C.; Walsh, M. ); Humenik, K.E. )

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs.

  5. Corrosion problems with aqueous coolants, final report

    SciTech Connect

    Diegle, R B; Beavers, J A; Clifford, J E

    1980-04-11

    The results of a one year program to characterize corrosion of solar collector alloys in aqueous heat-transfer media are summarized. The program involved a literature review and a laboratory investigation of corrosion in uninhibited solutions. It consisted of three separate tasks, as follows: review of the state-of-the-art of solar collector corrosion processes; study of corrosion in multimetallic systems; and determination of interaction between different waters and chemical antifreeze additives. Task 1 involved a comprehensive review of published literature concerning corrosion under solar collector operating conditions. The reivew also incorporated data from related technologies, specifically, from research performed on automotive cooling systems, cooling towers, and heat exchangers. Task 2 consisted of determining the corrosion behavior of candidate alloys of construction for solar collectors in different types of aqueous coolants containing various concentrations of corrosive ionic species. Task 3 involved measuring the degradation rates of glycol-based heat-transfer media, and also evaluating the effects of degradation on the corrosion behavior of metallic collector materials.

  6. Steam as turbine blade coolant: Experimental data generation

    SciTech Connect

    Wilmsen, B.; Engeda, A.; Lloyd, J.R.

    1995-10-01

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  7. Longer life for glyco-based stationary engine coolants

    SciTech Connect

    Hohlfeld, R.

    1996-07-01

    Large, stationary diesel engines used to compress natural gas that is to be transported down pipelines generate a great deal of heat. Unless this heat is dissipated efficiently, it will eventually cause an expensive breakdown. Whether the coolant uses ethylene glycol or propylene glycol, the two major causes of glycol degradation are heat and oxidation. The paper discusses inhibitors that enhance coolant service life and presents a comprehensive list of do`s and don`ts for users to gain a 20-year coolant life.

  8. Turbomachine injection nozzle including a coolant delivery system

    DOEpatents

    Zuo, Baifang

    2012-02-14

    An injection nozzle for a turbomachine includes a main body having a first end portion that extends to a second end portion defining an exterior wall having an outer surface. A plurality of fluid delivery tubes extend through the main body. Each of the plurality of fluid delivery tubes includes a first fluid inlet for receiving a first fluid, a second fluid inlet for receiving a second fluid and an outlet. The injection nozzle further includes a coolant delivery system arranged within the main body. The coolant delivery system guides a coolant along at least one of a portion of the exterior wall and around the plurality of fluid delivery tubes.

  9. Method for removing cesium from a nuclear reactor coolant

    DOEpatents

    Colburn, Richard P.

    1986-01-01

    A method of and system for removing cesium from a liquid metal reactor coolant including a carbon packing trap in the primary coolant system for absorbing a major portion of the radioactive cesium from the coolant flowing therethrough at a reduced temperature. A regeneration subloop system having a secondary carbon packing trap is selectively connected to the primary system for isolating the main trap therefrom and connecting it to the regeneration system. Increasing the temperature of the sodium flowing through the primary trap diffuses a portion of the cesium

  10. Modeling Reactor Coolant Systems Thermal-Hydraulic Transients

    Energy Science and Technology Software Center (ESTSC)

    1999-10-05

    RELAP5/MOD3.2* is used to model reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents and operational transients such as anticipated transients without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal-hydraulic systems. Control system and secondary system components are included to allow modeling of themore » plant controls, turbines, condensers, and secondary feedwater systems.« less

  11. EVALUATION OF FILTRATION AND DISTILLATION METHODS FOR RECYCLING AUTOMOTIVE COOLANT

    EPA Science Inventory

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive engine coolants at a New Jersey Department of Transportation garage. he specific recycling units evaluated are based on the technologies of filtration and distilla...

  12. Transient two-phase performance of LOFT reactor coolant pumps

    SciTech Connect

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed.

  13. INVESTIGATION OF CLEANER TECHNOLOGIES TO MINIMIZE AUTOMOTIVE COOLANT WASTES

    EPA Science Inventory

    The US Environmental Protection Agency in cooperation with the State of New Jersey evaluated chemical filtration and distillation technologies designed to recycle automotive and heavy-duty engine coolants. These evaluations addressed the product quality, waste reduction and econo...

  14. 14 CFR 23.1063 - Coolant tank tests.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... greater, plus the maximum working pressure of the system; and (b) For a tank with a nonmetallic liner the... specimen liner must be conducted with the coolant at operating temperature. Induction System...

  15. 14 CFR 23.1063 - Coolant tank tests.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... greater, plus the maximum working pressure of the system; and (b) For a tank with a nonmetallic liner the... specimen liner must be conducted with the coolant at operating temperature. Induction System...

  16. 14 CFR 23.1063 - Coolant tank tests.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... greater, plus the maximum working pressure of the system; and (b) For a tank with a nonmetallic liner the... specimen liner must be conducted with the coolant at operating temperature. Induction System...

  17. 14 CFR 23.1063 - Coolant tank tests.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... greater, plus the maximum working pressure of the system; and (b) For a tank with a nonmetallic liner the... specimen liner must be conducted with the coolant at operating temperature. Induction System...

  18. 14 CFR 23.1063 - Coolant tank tests.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... greater, plus the maximum working pressure of the system; and (b) For a tank with a nonmetallic liner the... specimen liner must be conducted with the coolant at operating temperature. Induction System...

  19. Optimized Coolant-Flow Diverter For Increased Bearing Life

    NASA Technical Reports Server (NTRS)

    Subbaraman, Maria R.; Butner, Myles F.

    1995-01-01

    Coolant-flow diverter for rolling-element bearings in cryogenic turbopump designed to enhance cooling power of flow in contact with bearings and thereby reduce bearing wear. Delivers jets of coolant as close as possible to hot spots at points of contact between balls and race. Also imparts swirl that enhances beneficial pumping effect. Used with success in end ball bearing of high-pressure-oxidizer turbopump.

  20. Hydrodynamics of large-scale fuel-coolant interactions. [LMFBR

    SciTech Connect

    Baines, M.; Board, S.J.; Buttery, N.E.

    1980-06-01

    The analogy between thermal reactive and chemical reactive flows suggests that all propagating thermal explosions have a detonation-like (i.e., shock) structure. A vapor detonation model, which allows for thermal disequilibrium in the coolant, is developed. It is suggested that similar nonequilibrium effects may limit the efficiency of UO/sub 2/-sodium system, however, because of high conductivity of the coolant. 34 refs.

  1. Investigation of cleaner technologies to minimize automotive coolant wastes

    SciTech Connect

    Randall, P.M.

    1993-01-01

    The U.S. Environmental Protection Agency in cooperation with the State of New Jersey evaluated chemical filtration and distillation technologies designed to recycle automotive and heavy-duty engine coolants. These evaluations addressed the product quality, waste reduction, and economic issues. In addition, the authors examined the potential for substituting propylene glycol for ethylene glycol based engine coolant formulations. (Copyright (c) 1993 Butterworth-Heinemann Ltd.)

  2. Sealing of a shrouded rotor-stator system with pre-swirl coolant

    NASA Astrophysics Data System (ADS)

    El-Oun, Z. B.; Neller, P. H.; Turner, A. B.

    1987-05-01

    Experimental results for a modeled gas turbine rotor-stator system using both preswirled blade coolant and radially outward flowing disc coolant are presented. Although the preswirled coolant flow is found to have little effect on the pressure distribution below the preswirl nozzles, it is shown that considerable contamination of the preswirled coolant by the frictionally heated disc coolant can occur. A clear pressure inversion effect was found when coolant was provided by the preswirl nozzles alone, while the pressure under the rim seal increased with increasing rotational speed. Blade coolant flow increases the sealing flow requirement, except at the lowest flow rates.

  3. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    DOEpatents

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  4. CFD modeling of turbulent duct flows for coolant channel analysis

    NASA Astrophysics Data System (ADS)

    Ungewitter, Ronald J.; Chan, Daniel C.

    1993-07-01

    The design of modern liquid rocket engines requires the analysis of chamber coolant channels to maximize the heat transfer while minimizing the coolant flow. Coolant channels often do not remain at a constant cross section or at uniform curvature. New designs require higher aspect ratio coolant channels than previously used. To broaden the analysis capability and to complement standard analysis tools an investigation on the accuracy of CFD predictions for coolant channel flow has been initiated. Validation of CFD capabilities for coolant channel analysis will enhance the capabilities for optimizing design parameters without resorting to extensive experimental testing. The eventual goal is to use CFD to determine the flow fields of unique coolant channel designs and therefore determine critical heat transfer coefficients. In this presentation the accuracy of a particular CFD code is evaluated for turbulent flows. The first part of the presentation is a comparison of numerical results to existing cold flow data for square curved ducts (NASA CR-3367, 'Measurements of Laminar and Turbulent Flow in a Curved Duct with Thin Inlet Boundary Layers'). The results of this comparison show good agreement with the relatively coarse experimental data. The second part of the presentation compares two cases of higher aspect ratio channels (AR=2.5,10) to show changes in axial and secondary flow strength. These cases match experimental work presently in progress and will be used for future validation. The comparison shows increased secondary flow strength of the higher aspect ratio case due to the change in radius of curvature. The presentation includes a test case with a heated wall to demonstrate the program's capability. The presentation concludes with an outline of the procedure used to validate the CFD code for future design analysis.

  5. Nuclear criticality safety assessment of the proposed CFC replacement coolants

    SciTech Connect

    Jordan, W.C.; Dyer, H.R.

    1993-12-01

    The neutron multiplication characteristics of refrigerant-114 (R-114) and proposed replacement coolants perfluorobutane (C{sub 4}F{sub 10}) and cycloperfluorobutane C{sub 4}F{sub 8}) have been compared by evaluating the infinite media multiplication factors of UF{sub 6}/H/coolant systems and by replacement calculations considering a 10-MW freezer/sublimer. The results of these comparisons demonstrate that R-114 is a neutron absorber, due to its chlorine content, and that the alternative fluorocarbon coolants are neutron moderators. Estimates of critical spherical geometries considering mixtures of UF{sub 6}/HF/C{sub 4}F{sub 10} indicate that the flourocarbon-moderated systems are large compared with water-moderated systems. The freezer/sublimer calculations indicate that the alternative coolants are more reactive than R-114, but that the reactivity remains significantly below the condition of water in the tubes, which was a limiting condition. Based on these results, the alternative coolants appear to be acceptable; however, several follow-up tasks have been recommended, and additional evaluation will be required on an individual equipment basis.

  6. Stagnation region gas film cooling: Effects of dimensionless coolant temperature

    NASA Technical Reports Server (NTRS)

    Bonnice, M. A.; Lecuyer, M. R.

    1983-01-01

    An experimental investigation was conducted to mode the film cooling performance for a turbine vane leading edge using the stagnation region of a cylinder in cross flow. Experiments were conducted with a single row of spanwise angled (25 deg) coolant holes for a range of the coolant blowing ratio and dimensionless coolant temperature with free stream-to-wall temperature ratio approximately 1.7 and Re sub D = 90000. the cylindrical test surface was instrumented with miniature heat flux gages and wall thermocouples to determine the percentage reduction in the Stanton number as a function of the distance downstream from injection (x/d sub 0) and the location between adjacent holes (z/S). Data from local heat flux measurements are presented for injection from a single row located at 5 deg, 22.9 deg, 40.8 deg, from stagnation using a hole spacing ratio of S/d = 5. The film coolant was injected with T sub c T sub w with a dimensionless coolant temperature in the range 1.18 or equal to theta sub c or equal to 1.56. The data for local Stanton Number Reduction (SNR) showed a significant increase in SNR as theta sub c was increased above 1.0.

  7. Coolant mixing and distribution in a transparent reactor model

    SciTech Connect

    Fanning, M.W.; Haury, G.; Pflug, L.; Rothe, P.H.

    1983-11-01

    Following a small break loss-of-coolant accident in a pressurized water reactor, coolant water may be injected at high pressure to help cool the core. This paper reports the results of tests which determined the mixing and distribution of the coolant in a 1/5-scale transparent model of the reactor. The model components included the reactor vessel, cold leg pipe, pump, and loop seal with steam generator and hot leg simulators completing the flow loop. Tests were conducted for a no-refill condition with constant liquid inventory in the facility and zero flow of the primary water. Salt water, dyed red was used for the coolant water to create prototypical density differences in this atmospheric facility. Steady state fluid distribution was determined from flow and density measurements and complete mass balances. Interpretation of the quantitative results was aided by extensive flow visualization studies which include still photographs and motion picture films for all tests. The test parameters included the fluid density ratio, the flow rate of coolant water, and the flow rate of primary water injected in the vessel downcomer to simulate a natural circulation flow through vent valves between the reactor core and the downcomer. Four locations of the small break were tested.

  8. Reactor coolant seal testing under station blackout conditions

    SciTech Connect

    Marsi, J.A.

    1988-01-01

    Failures of reactor coolant pump (RCP) seals that could result in a significant loss-of-coolant inventory are of current concern to the US Nuclear Regulatory Commission. Particular attention is being focused on seal behavior during station blackout conditions, when failure of on-site emergency diesel generators occurs simultaneously with loss of all off-site alternating current power. Under these conditions, both seal injection flow and component cooling water flow are lost, and the RCP seals are exposed to full reactor coolant temperature. Overheating of elastomeric components and flashing of coolant across the sealing faces can cause unacceptably high leakage rates, with potential catastrophic consequences. A test program has been conducted that subjects full-scale seal cartridges to typical pressurized water reactor (PWR) coolant system steady-state and transient operation conditions including associated dynamic shaft motions. A special test segment was developed to evaluate seal operation under station blackout conditions. The test program successfully mirrored the severity of an actual loss-of-seal cooling water event under station blackout conditions, and the Byron Jackson{reg sign} N-9000 seal cartridge maintained its integrity.

  9. Diesel engine coolant analysis, new application for established instrumentation

    SciTech Connect

    Anderson, D.P.; Lukas, M.; Lynch, B.K.

    1998-09-01

    Rotating disk electrode (RDE) arc emission spectrometers are used in many commercial, industrial and military laboratories throughout the world to analyze millions of oil and fuel samples each year. In fact, RDE spectrometers have been used exclusively for oil and fuel analysis for so long, that most practitioners have probably forgotten that when RDE spectrometers were first introduced more than 40 years ago, they were also routinely used for aqueous samples. This paper describes recent work to calibrate and modify RDE arc emission spectrometers for the analysis of engine coolant samples; a mixture of approximately 50% water and 50% glycol. The technique has been shown to be effective for the analysis of wear metals, contamination and supplemental coolant additives in ethylene and propylene glycol. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and RDE spectrometers will be presented. The data correlates extremely well on new and relatively clean coolants. However, not surprisingly, RDE results are sometimes higher for samples containing particles larger than a few micrometers. This paper suggests that RDE spectrometers are appropriate, and sometimes preferred, for most types of coolants and certain types of aqueous samples. Actual field data is be presented to support the arguments.

  10. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    SciTech Connect

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  11. Actively controlling coolant-cooled cold plate configuration

    SciTech Connect

    Chainer, Timothy J.; Parida, Pritish R.

    2015-07-28

    A method is provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The method includes: monitoring a variable associated with at least one of the coolant-cooled cold plate or one or more electronic components being cooled by the cold plate; and dynamically varying, based on the monitored variable, a physical configuration of the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the one or more electronic components, and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the coolant-cooled cold plate, the positioning of which may be adjusted based on the monitored variable.

  12. Method for removing cesium from a nuclear reactor coolant

    DOEpatents

    Colburn, R.P.

    1983-08-10

    A method of and system for removing cesium from a liquid metal reactor coolant including a carbon packing trap in the primary coolant system for absorbing a major portion of the radioactive cesium from the coolant flowing therethrough at a reduced temperature. A regeneration subloop system having a secondary carbon packing trap is selectively connected to the primary system for isolating the main trap therefrom and connecting it to the regeneration system. Increasing the temperature of the sodium flowing through the primary trap diffuses a portion of the cesium inventory thereof further into the carbon matrix while simultaneously redispersing a portion into the regeneration system for absorption at a reduced temperature by the secondary trap.

  13. Spectrophotometric Procedure for Fast Reactor Advanced Coolant Manufacture Control

    NASA Astrophysics Data System (ADS)

    Andrienko, O. S.; Egorov, N. B.; Zherin, I. I.; Indyk, D. V.

    2016-01-01

    The paper describes a spectrophotometric procedure for fast reactor advanced coolant manufacture control. The molar absorption coefficient of dimethyllead dibromide with dithizone was defined as equal to 68864 ± 795 l·mole-1·cm-1, limit of detection as equal to 0.583 · 10-6 g/ml. The spectrophotometric procedure application range was found to be equal to 37.88 - 196.3 g. of dimethyllead dibromide in the sample. The procedure was used within the framework of the development of the method of synthesis of the advanced coolant for fast reactors.

  14. Code System to Calculate Reactor Coolant System Leak Rate.

    SciTech Connect

    Bell, Pat

    1999-10-19

    Version 00 RCSLK9 was developed to analyze the leak tightness of the primary coolant system for any pressurized water reactor (PWR). From given system conditions, water levels in tanks, and certain system design parameters, RCSLK9 calculates the loss of water from the reactor coolant system (RCS) and the increase of water in the leakage collection system during an arbitrary time interval. The program determines the system leak rates and displays or prints a report of the results. During the initial application to a specific reactor, RCSLK9 creates a file of system parameters and saves it for future use.

  15. Code System to Calculate Reactor Coolant System Leak Rate.

    Energy Science and Technology Software Center (ESTSC)

    1999-10-19

    Version 00 RCSLK9 was developed to analyze the leak tightness of the primary coolant system for any pressurized water reactor (PWR). From given system conditions, water levels in tanks, and certain system design parameters, RCSLK9 calculates the loss of water from the reactor coolant system (RCS) and the increase of water in the leakage collection system during an arbitrary time interval. The program determines the system leak rates and displays or prints a report ofmore » the results. During the initial application to a specific reactor, RCSLK9 creates a file of system parameters and saves it for future use.« less

  16. Evaluation of engine coolants under flow boiling conditions

    SciTech Connect

    McAssey, E.V. Jr.; Stinson, C.; Gollin, M.

    1995-12-31

    An experimental program has been conducted to evaluate the heat transfer performance of two engine coolant mixtures, propylene-glycol/water and ethylene-glycol/water. In each mixture, the concentration was 50-50 by volume. Performance in this situation is defined as the ability to maintain a lower surface temperature for a given flux. The heat transfer regimes considered covered the range from single phase forced convection through saturated flow boiling. Results show that both coolants perform satisfactorily. However, in single phase convection, ethylene-glycol/water is slightly more effective. Conversely, for sub-cooled nucleate boiling and saturated boiling, propylene-glycol/water results in slightly lower metal temperatures.

  17. Development of Figure of Merits (FOMs) for Intermediate Coolant Characterization and Selection

    SciTech Connect

    Eung Soo Kim; Piyush Sabharwall; Nolan Anderson

    2011-06-01

    This paper focuses on characterization of several coolant performances in the IHTL. There are lots of choices available for the IHTL coolants; gases, liquid metals, molten salts, and etc. Traditionally, the selection of coolants is highly dependent on engineer's experience and decisions. In this decision, the following parameters are generally considered: melting point, vapor pressure, density, thermal conductivity, heat capacity, viscosity, and coolant chemistry. The followings are general thermal-hydraulic requirements for the coolant in the IHTL: (1) High heat transfer performance - The IHTL coolant should exhibit high heat transfer performance to achieve high efficiency and economics; (2) Low pumping power - The IHTL coolant requires low pumping power to improve economics through less stringent pump requirements; (3) Low amount of coolant volume - The IHTL coolant requires less coolant volume for better economics; (4) Low amount of structural materials - The IHTL coolant requires less structural material volume for better economics; (5) Low heat loss - The IHTL requires less heat loss for high efficiency; and (6) Low temperature drop - The IHTL should allow less temperature drop for high efficiency. Typically, heat transfer coolants are selected based on various fluid properties such as melting point, vapor pressure, density, thermal conductivity, heat capacity, viscosity, and coolant chemistry. However, the selection process & results are highly dependent on the engineer's personal experience and skills. In the coolant selection, if a certain coolant shows superior properties with respect to the others, the decision will be very straightforward. However, generally, each coolant material exhibits good characteristics for some properties but poor for the others. Therefore, it will be very useful to have some figures of merits (FOMs), which can represent and quantify various coolant thermal performances in the system of interest. The study summarized in this

  18. AUTOMOTIVE AND HEAVY-DUTY ENGINE COOLANT RECYCLING BY FILTRATION

    EPA Science Inventory

    This evaluation addresses the product quality, waste reduction and economic issues involved in recycling automotive and heavy-duty engine coolants. he specific recycling units evaluated are a fleet-size unit and a portable unit, both based on the technology of chemical filtration...

  19. PIPING FOR COOLANT WATER IS INSTALLED INSIDE REACTOR STRUCTURE PRIOR ...

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

    PIPING FOR COOLANT WATER IS INSTALLED INSIDE REACTOR STRUCTURE PRIOR TO EMBEDMENT IN CONCRETE. HIGHER PIPE IS INLET; THE OTHER, THE OUTLET LOOP. INLET PIPE WILL CONNECT TO TOP SECTION OF REACTOR VESSEL. INL NEGATIVE NO. 1287. Unknown Photographer, 1/18/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  20. Coolant Characteristics and Control in Direct Chill Casting

    SciTech Connect

    2001-10-01

    This project focuses on understanding the fundamentals of coolant behavior and developing strategies to control the cooling rate of DC casting of aluminum ingots. Project partners will conduct a fundamental study to identify various parameters affecting critical heat flux and boiling transition and evaluate the effects of various additives (impurity particulates, sodium and calcium salts, carbonates, bicarbonates, surfactants, etc.).

  1. Coolants with selective optical filtering characteristics for ruby laser applications

    NASA Technical Reports Server (NTRS)

    Mc Devitt, F. R.; Rasquin, J. R.

    1968-01-01

    Coolant-filtering medium developed consists of a solution of copper sulfate in a 4-1 volumetric mixture of ethanol and methanol. This solution should be a useful addition to ruby laser systems, particularily in large pulse or Q switching applications.

  2. Corrosion of structural materials by lead-based reactor coolants.

    SciTech Connect

    Abraham, D. P.; Leibowitz, L.; Maroni, V. A.; McDeavitt, S. M.; Raraz, A. G.

    2000-11-16

    Advanced nuclear reactor design has, in recent years, focused increasingly on the use of heavy-liquid-metal coolants, such as lead and lead-bismuth eutectic. Similarly, programs on accelerator-based transmutation systems have also considered the use of such coolants. Russian experience with heavy-metal coolants for nuclear reactors has lent credence to the validity of this approach. Of significant concern is the compatibility of structural materials with these coolants. We have used a thermal convection-based test method to allow exposure of candidate materials to molten lead and lead-bismuth flowing under a temperature gradient. The gradient was deemed essential in evaluating the behavior of the test materials in that should preferential dissolution of components of the test material occur we would expect dissolution in the hotter regions and deposition in the colder regions, thus promoting material transport. Results from the interactions of a Si-rich mild steel alloy, AISI S5, and a ferritic-martensitic stainless steel, HT-9, with the molten lead-bismuth are presented.

  3. Integral coolant channels supply made by melt-out method

    NASA Technical Reports Server (NTRS)

    Escher, W. J. D.

    1964-01-01

    Melt-out method of constructing strong, pressure-tight fluid coolant channels for chambers is accomplished by cementing pins to the surface and by depositing a melt-out material on the surface followed by two layers of epoxy-resin impregnated glass fibers. The structure is heated to melt out the low-melting alloy.

  4. Fuels, Lubricants, and Coolants. FOS: Fundamentals of Service.

    ERIC Educational Resources Information Center

    John Deere Co., Moline, IL.

    This manual on fuels, lubricants, and coolants is one of a series of power mechanics tests and visual aids on automotive and off-the-road agricultural and construction equipment. Materials present basic information with illustrations for use by vocational students and teachers as well as shop servicemen and laymen. Focusing on fuels, the first of…

  5. AUTOMOTIVE AND HEAVY-DUTY ENGINE COOLANT RECYCLING BY DISTILLATION

    EPA Science Inventory

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants for a facility such as the New Jersey Department of Transportation garage in Ewing, New Jersey. he specific recycling evaluated is b...

  6. EVALUATION OF FILTRATION AND DISTILLATION METHODS FOR RECYCLING AUTOMOTIVE COOLANT.

    EPA Science Inventory

    This evaluation addresses the product quality, waste reduction, and economic issues involved in recycling automotive and heavy-duty engine coolants at a New Jersey Department of Transportation garage. The specific recycling units evaluated are based on the technologies of filtrat...

  7. 37. Upper level, chromate tanks (formerly provided coolant to missile ...

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

    37. Upper level, chromate tanks (formerly provided coolant to missile guidance section, retractor cables for lock pin in front of ladder at left - Ellsworth Air Force Base, Delta Flight, Launch Facility, On County Road T512, south of Exit 116 off I-90, Interior, Jackson County, SD

  8. Antimony tartrate corrosion inhibitive composition for coolant systems

    SciTech Connect

    Payerle, N.E.

    1987-08-11

    An automobile coolant concentrate is described comprising (a) a liquid polyhydric alcohol chosen from the group consisting of ethylene glycol, propylene glycol, diethylene glycol and mixtures thereof, and (b) corrosion inhibitors in a corrosion inhibitory amount with respect to corrosion of lead-containing solders, the corrosion inhibitors comprising (i) an alkali metal antimony tartrate, and (ii) an azole compound.

  9. Thermal explosions resulting from fuel-coolant interactions

    SciTech Connect

    Bartusiak, R.D.; Caram, H.S.

    1988-08-01

    High-speed photographic data and pressure traces of thermal explosions from the contact of single drops of ion oxide with water were analyzed according to models describing underwater chemical explosion and cavitation bubbles. The objective was to develop a simple method for analyzing the microscale hydrodynamics of fuel-coolant interactions (FCI). For a given external pressure and liquid density essentially all the features of the radial motion of the explosion bubble, including the total energy release, are uniquely determined by a single parameter - the bubble period. Nearly all of the heat transfer from fuel to coolant occurs during the 10/sup -5/ to 10/sup -4/ sec timespan of coolant vapor film collapse during which the fuel fragments. The features of the resulting explosion bubble are not significantly affected by the degree of heat transfer from vapor to coolant liquid and the bubble can be modeled as an empty cavity. The method developed during this study should facilitate investigations on FCI by simplifying the analyses of thermal explosion data. Further attention can be given to experiments on the effects of fuel parameters, e.g., surface tension and viscosity, on fragmentation, heat transfer, and explosive yield.

  10. INVESTIGATION OF CLEANER TECHNOLOGIES TO MINIMIZE AUTOMOTIVE COOLANT WASTES

    EPA Science Inventory

    The U.S. Environmental Protection Agency in cooperation with the State of New Jersey evaluated chemical filtration and distillation technologies designed to recycle automotive and heavy-duty engine coolants. hese evaluations addressed the product quality, waste reduction, and eco...

  11. Directly connected heat exchanger tube section and coolant-cooled structure

    DOEpatents

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

    2014-04-01

    A cooling apparatus for an electronics rack is provided which includes an air-to-liquid heat exchanger, one or more coolant-cooled structures and a tube. The heat exchanger, which is associated with the electronics rack and disposed to cool air passing through the rack, includes a plurality of distinct, coolant-carrying tube sections, each tube section having a coolant inlet and a coolant outlet, one of which is coupled in fluid communication with a coolant loop to facilitate flow of coolant through the tube section. The coolant-cooled structure(s) is in thermal contact with an electronic component(s) of the rack, and facilitates transfer of heat from the component(s) to the coolant. The tube connects in fluid communication one coolant-cooled structure and the other of the coolant inlet or outlet of the one tube section, and facilitates flow of coolant directly between that coolant-carrying tube section of the heat exchanger and the coolant-cooled structure.

  12. Experimental interaction of magma and “dirty” coolants

    NASA Astrophysics Data System (ADS)

    Schipper, C. Ian; White, James D. L.; Zimanowski, Bernd; Büttner, Ralf; Sonder, Ingo; Schmid, Andrea

    2011-03-01

    The presence of water at volcanic vents can have dramatic effects on fragmentation and eruption dynamics, but little is known about how the presence of particulate matter in external water will further alter eruptions. Volcanic edifices are inherently “dirty” places, where particulate matter of multiple origins and grainsizes typically abounds. We present the results of experiments designed to simulate non-explosive interactions between molten basalt and various “coolants,” ranging from homogeneous suspensions of 0 to 30 mass% bentonite clay in pure water, to heterogeneous and/or stratified suspensions including bentonite, sand, synthetic glass beads and/or naturally-sorted pumice. Four types of data are used to characterise the interactions: (1) visual/video observations; (2) grainsize and morphology of resulting particles; (3) heat-transfer data from a network of eight thermocouples; and (4) acoustic data from three force sensors. In homogeneous coolants with <~10% bentonite, heat transfer is by convection, and the melt is efficiently fragmented into blocky particles through multiple thermal granulation events which produce associated acoustic signals. For all coolants with >~20% sediment, heat transfer is by forced convection and conduction, and thermal granulation is less efficient, resulting in fewer blocky particles, larger grainsizes, and weaker acoustic signals. Many particles are droplet-shaped or/and “vesicular,” containing bubbles filled with coolant. Both of these particle types indicate significant hydrodynamic magma-coolant mingling, and many of them are rewelded into compound particles. The addition of coarse material to heterogeneous suspensions further slows heat transfer thus reducing thermal granulation, and variable interlocking of large particles prevents efficient hydrodynamic mingling. This results primarily in rewelded melt piles and inefficient distribution of melt and heat throughout the coolant volume. Our results indicate

  13. Modular Porous Plate Sublimator /MPPS/ requires only water supply for coolant

    NASA Technical Reports Server (NTRS)

    Rathbun, R. J.

    1966-01-01

    Modular porous plate sublimators, provided for each location where heat must be dissipated, conserve the battery power of a space vehicle by eliminating the coolant pump. The sublimator requires only a water supply for coolant.

  14. Numerical simulation of PWR response to a small break LOCA (loss-of-coolant accident) with reactor coolant pumps operating

    SciTech Connect

    Adams, J.P.; Dobbe, C.A.; Bayless, P.D.

    1986-01-01

    Calculations have been made of the response of pressurized water reactors (PWRs) during a small-break, loss-of-coolant accident with the reactor coolant pumps (RCPs) operating. This study was conducted, as part of a comprehensive project, to assess the relationship between measurable RCP parameters, such as motor power or current, and fluid density, both local (at the RCP inlet) and global (average reactor coolant system). Additionally, the efficacy of using these RCP parameters, together with fluid temperature, to identify an off-nominal transient as either a LOCA, a heatup transient, or a cooldown transient and to follow recovery from the transient was assessed. The RELAP4 and RELAP5 computer codes were used with three independent sets of RCP, two-phase degradation multipliers. These multipliers were based on data obtained in two-phase flow conditions for the Semiscale, LOFT, and Creare/Combustion Engineering (CE)/Electric Power Research Institute (EPRI) pumps, respectively. Two reference PWRs were used in this study: Zion, a four-loop, 1100-MWe, Westinghouse plant operated by Commonwealth Edison Co. in Zion, Illinois and Bellefonte, a two-by-four loop, 1213 MWe, Babcock and Wilcox designed plant being built by the Tennessee Valley Authority in Scottsboro, Alabama. The results from this study showed that RCP operation resulted in an approximately homogeneous reactor coolant system and that this result was independent of reference plant, computer code, or two-phase RCP head degradation multiplier used in the calculation.

  15. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems...

  16. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems...

  17. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems...

  18. Directly connected heat exchanger tube section and coolant-cooled structure

    DOEpatents

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

    2015-09-15

    A method is provided for fabricating a cooling apparatus for cooling an electronics rack, which includes an air-to-liquid heat exchanger, one or more coolant-cooled structures, and a tube. The heat exchanger is associated with the electronics rack and disposed to cool air passing through the rack, includes a plurality of coolant-carrying tube sections, each tube section having a coolant inlet and outlet, one of which is coupled in fluid communication with a coolant loop to facilitate flow of coolant through the tube section. The coolant-cooled structure(s) is in thermal contact with an electronic component(s) of the rack, and facilitates transfer of heat from the component(s) to the coolant. The tube connects in fluid communication one coolant-cooled structure and the other of the coolant inlet or outlet of the one tube section, and facilitates flow of coolant directly between that coolant-carrying tube section of the heat exchanger and the coolant-cooled structure.

  19. Health physics aspects of processing EBR-I coolant

    SciTech Connect

    Burke, L.L.; Thalgott, J.O.; Poston, J.W. Jr.

    1998-12-31

    The sodium-potassium reactor coolant removed from the Experimental Breeder Reactor Number One after a partial reactor core meltdown had been stored at the Idaho National Engineering and Environmental Laboratory for 40 years. The State of Idaho considered this waste the most hazardous waste stored in the state and required its processing. The reactor coolant was processed in three phases. The first phase converted the alkali metal into a liquid sodium-potassium hydroxide. The second phase converted this caustic to a liquid sodium-potassium carbonate. The third phase solidified the sodium-potassium carbonate into a form acceptable for land disposal. Health physics aspects and dose received during each phase of the processing are discussed.

  20. Glycol coolants improve heat transfer and corrosion control

    SciTech Connect

    Holfield, R.

    1995-03-01

    Various liquids from plain water to exotic fluids have been used as coolants in large stationary diesel engines that drive compressors on natural gas pipeline distribution systems. Although water is an efficient heat transfer medium, its drawbacks of freezing at {minus}32 F and boiling at 212 F seriously limit its usefulness. Special glycol-based heat transfer fluids are available and refined specifically for long-term needs of gas compressor engines. Appropriate corrosion inhibitors have been formulated for metallurgy and operating conditions encountered with these engines. Propylene glycol was developed as an alternative for use in environmentally sensitive areas. Glycol-based fluids must be specifically inhibited for industrial applications because uninhibited or improperly inhibited coolants can seriously damage reciprocating engines.

  1. Copper-triazole interaction and coolant inhibitor depletion

    SciTech Connect

    Bartley, L.S.; Fritz, P.O.; Pellet, R.J.; Taylor, S.A.; Van de Ven, P.

    1999-08-01

    To a large extent, the depletion of tolyltriazole (TTZ) observed in several field tests may be attributed to the formation of a protective copper-triazole layer. Laboratory aging studies, shown to correlate with field experience, reveal that copper-TTZ layer formation depletes coolant TTZ levels in a fashion analogous to changes observed in the field. XPS and TPD-MS characterization of the complex formed indicates a strong chemical bond between copper and the adsorbed TTZ which can be desorbed thermally only at elevated temperatures. Electrochemical polarization experiments indicate that the layer provides good copper protection even when TTZ is absent from the coolant phase. Examination of copper cooling system components obtained after extensive field use reveals the presence of a similar protective layer.

  2. Expert system for online surveillance of nuclear reactor coolant pumps

    DOEpatents

    Gross, Kenny C.; Singer, Ralph M.; Humenik, Keith E.

    1993-01-01

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

  3. Experimental investigations of thermal interaction between corium and coolants

    NASA Astrophysics Data System (ADS)

    Zagorul'ko, Yu. I.; Zhmurin, V. G.; Volov, A. N.; Kovalev, Yu. P.

    2008-03-01

    We present a generalized analysis of the experimental results from investigations of thermal interaction in corium simulators (melts of thermite mixtures U + Mo3 and Zr + Fe2O3)-coolant (Na and H2O) systems. We also present the results from experimental assessments of the kinematic characteristics pertinent to the displacement of materials during the thermal interaction process and the coefficients for conversion of the corium thermal energy into mechanical work.

  4. 92. View of transmitter building no. 102 first floor coolant ...

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

    92. View of transmitter building no. 102 first floor coolant process water tanks (sodium bisulfate solution), stainless steel, for electronic systems cooling in transmitter and MIP rooms. RCA Services Company 29 September, 1960, official photograph BMEWS Project by unknown photograph, Photographic Services, Riverton, NJ, BMEWS, clear as negative no. A-1226 - Clear Air Force Station, Ballistic Missile Early Warning System Site II, One mile west of mile marker 293.5 on Parks Highway, 5 miles southwest of Anderson, Anderson, Denali Borough, AK

  5. Circulation within the primary system at TMI-2 with lowered coolant level and at atmospheric conditions

    SciTech Connect

    Baston, V.F.; Hofstetter, K.J.; Malinauskas, A.P.

    1985-06-01

    Experiments were performed with the Three Mile Island reactor coolant system open to the atmosphere and the coolant lowered to a level above the fuel (a condition required for defueling) to ascertain the extent of coolant mixing. A principal concern for coolant decontamination during defueling is the radionuclides released and their distribution within the primary system. Analyses of radionuclide, temperature, and dissolved oxygen data taken during these experiments confirm mixing in the primary system under forced coolant flow conditions with only minimal mixing occurring during static periods.

  6. Turbulent Dispersion of Film Coolant in a Turbine Vane Cascade

    NASA Astrophysics Data System (ADS)

    Yapa, Sayuri; Elkins, Christopher; Eaton, John

    2012-11-01

    Gas turbine engines operate at peak temperatures in excess of the material limits because the high pressure turbine nozzles and buckets are film cooled. The nozzle vanes of the first stage turbine use the most cooling air because they are exposed directly to the high temperature combustor exhaust. Existing turbine analysis assumes a uniform temperature at the rotor inlet. However, the coolant does not mix completely with the mainstream flow before impinging on the turbine rotor, and the coolant streaks create variations in temperature along the leading edge of the downstream turbine blades. 3D velocity and concentration measurements are made using magnetic resonance (MR) imaging techniques to study turbulent mixing in a realistic film-cooled nozzle vane cascade. A scalar mixing analogy for thermal diffusion is used in which a chemical contaminant plays the role of temperature. In a typical experiment, the mainstream flow is water and the film coolant is a copper sulfate solution. The concentration of copper sulfate measured anywhere in the flow is a surrogate for normalized temperature. The turbulent scalar diffusivity in the scalar transport equation can be estimated from the MR data and used to improve computational fluid dynamics models. Army Research Office.

  7. Crack stability analysis of low alloy steel primary coolant pipe

    SciTech Connect

    Tanaka, T.; Kameyama, M.; Urabe, Y.

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  8. Coolant Design System for Liquid Propellant Aerospike Engines

    NASA Astrophysics Data System (ADS)

    McConnell, Miranda; Branam, Richard

    2015-11-01

    Liquid propellant rocket engines burn at incredibly high temperatures making it difficult to design an effective coolant system. These particular engines prove to be extremely useful by powering the rocket with a variable thrust that is ideal for space travel. When combined with aerospike engine nozzles, which provide maximum thrust efficiency, this class of rockets offers a promising future for rocketry. In order to troubleshoot the problems that high combustion chamber temperatures pose, this research took a computational approach to heat analysis. Chambers milled into the combustion chamber walls, lined by a copper cover, were tested for their efficiency in cooling the hot copper wall. Various aspect ratios and coolants were explored for the maximum wall temperature by developing our own MATLAB code. The code uses a nodal temperature analysis with conduction and convection equations and assumes no internal heat generation. This heat transfer research will show oxygen is a better coolant than water, and higher aspect ratios are less efficient at cooling. This project funded by NSF REU Grant 1358991.

  9. Acceptance criteria for reactor coolant pumps and valves

    SciTech Connect

    Gupta, N.K.; Miller, R.F.; Sindelar, R.L.

    1993-05-01

    Each of the six primary coolant loop systems of the Savannah River Site (SRS) production reactors contains one reactor coolant pump, one PUMP suction side motor operated valve, and other smaller valves. The pumps me double suction, double volute, and radially split type pumps. The valves are different size shutoff and control valves rated from ANSI B16.5 construction class 150 to class 300. The reactor coolant system components, also known as the process water system (PWS), are classified as nuclear Safety Class I components. These components were constructed in the 1950`s in accordance with the then prevailing industry practices. No uniform construction codes were used for design and analysis of these components. However, no pressure boundary failures or bolting failures have ever been recorded throughout their operating history. Over the years, the in-service inspection (ISI) was limited to visual inspection of the pressure boundaries, and surface and volumetric examination of the pressure retaining bolts. Efforts are now underway to implement ISI requirements similar to the ASME Section XI requirements for pumps and valves. This report discusses the new ISI requirements which also call for volumetric examination of the pump casing and valve body welds.

  10. Turbine vane coolant flow variations and calculated effects on metal temperatures

    NASA Technical Reports Server (NTRS)

    Yeh, F. C.; Meitner, P. L.; Russell, L. M.

    1975-01-01

    Seventy-two air-cooled turbine vanes were tested to determine coolant flow variations among the vanes. Calculations were made to estimate the effect of measured coolant flow variations on local vane metal temperatures. The calculations were based on the following assumed operating conditions: turbine inlet temperature, 1700 K (2600 F); turbine inlet pressure, 31 N/sq cm (45 psia); coolant inlet temperature, 811 K (1000 F); and total coolant to gas flow ratio, 0.065. Variations of total coolant flow were not large (about 10 percent from the arithmetic mean) for all 72 vanes, but variations in local coolant flows were large. The local coolant flow variations ranged from 8 to 75 percent, and calculated metal temperature variations ranged from 8 to 60 K (15 to 180 F).

  11. Heat transfer characteristics for some coolant additives used for water cooled engines

    SciTech Connect

    Abou-Ziyan, H.Z.; Helali, A.H.B.

    1996-12-31

    Engine coolants contain certain additives to prevent engine overheating or coolant freezing in cold environments. Coolants, also, contain corrosion and rust inhibitors, among other additives. As most engines are using engine cooling solutions, it is of interest to evaluate the effect of engine coolants on the boiling heat transfer coefficient. This has its direct impact on radiator size and environment. This paper describes the apparatus and the measurement techniques. Also, it presents the obtained boiling heat transfer results at different parameters. Three types of engine coolants and their mixtures in distilled water are evaluated, under sub-cooled and saturated boiling conditions. A profound effect of the presence of additives in the coolant, on heat transfer, was clear since changes of heat transfer for different coolants were likely to occur. The results showed that up to 180% improvement of boiling heat transfer coefficient is experienced with some types of coolants. However, at certain concentrations other coolants provide deterioration or not enhancement in the boiling heat transfer characteristics. This investigation proved that there are limitations, which are to be taken into consideration, for the composition of engine coolants in different environments. In warm climates, ethylene glycol should be kept at the minimum concentration required for dissolving other components, whereas borax is beneficial to the enhancement of the heat transfer characteristics.

  12. Effect of Coolant Temperature and Mass Flow on Film Cooling of Turbine Blades

    NASA Technical Reports Server (NTRS)

    Garg, Vijay K.; Gaugler, Raymond E.

    1997-01-01

    A three-dimensional Navier Stokes code has been used to study the effect of coolant temperature, and coolant to mainstream mass flow ratio on the adiabatic effectiveness of a film-cooled turbine blade. The blade chosen is the VKI rotor with six rows of cooling holes including three rows on the shower head. The mainstream is akin to that under real engine conditions with stagnation temperature = 1900 K and stagnation pressure = 3 MPa. Generally, the adiabatic effectiveness is lower for a higher coolant temperature due to nonlinear effects via the compressibility of air. However, over the suction side of shower-head holes, the effectiveness is higher for a higher coolant temperature than that for a lower coolant temperature when the coolant to mainstream mass flow ratio is 5% or more. For a fixed coolant temperature, the effectiveness passes through a minima on the suction side of shower-head holes as the coolant to mainstream mass flow, ratio increases, while on the pressure side of shower-head holes, the effectiveness decreases with increase in coolant mass flow due to coolant jet lift-off. In all cases, the adiabatic effectiveness is highly three-dimensional.

  13. Cryogenic-coolant He4-superconductor dynamic and static interactions

    NASA Technical Reports Server (NTRS)

    Caspi, S.; Chuang, C.; Kim, Y. I.; Allen, R. J.; Frederking, T. H. E.

    1980-01-01

    A composite superconducting material (NbTi-Cu) was evaluated with emphasis on post quench solid cooling interaction regimes. The quasi-steady runs confirm the existence of a thermodynamic limiting thickness for insulating coatings. Two distinctly different post quench regimes of coated composites are shown to relate to the limiting thickness. Only one regime,, from quench onset to the peak value, revealed favorable coolant states, in particular in He2. Transient recovery shows favorable recovery times from this post quench regime (not drastically different from bare conductors) for both single coated specimens and a coated conductor bundle.

  14. Vegetable oils: liquid coolants for solar heating and cooling applications

    SciTech Connect

    Ingley, H A

    1980-02-01

    It has been proposed that vegetable oils, renewable byproducts of agriculture processes, be investigated for possible use as liquid coolants. The major thrust of the project was to investigate several thermophysical properties of the four vegetable oils selected. Vapor pressures, specific heat, viscosity, density, and thermal conductivity were determined over a range of temperatures for corn, soybean, peanut, and cottonseed oil. ASTM standard methods were used for these determinations. In addition, chemical analyses were performed on samples of each oil. The samples were collected before and after each experiment so that any changes in composition could be noted. The tests included iodine number, fatty acid, and moisture content determination. (MHR)

  15. System Study: High-Pressure Coolant Injection 1998–2013

    SciTech Connect

    Schroeder, John Alton

    2015-01-31

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

  16. System Study: High-Pressure Coolant Injection 1998-2014

    SciTech Connect

    Schroeder, John Alton

    2015-12-01

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 25 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

  17. System Study: High-Pressure Coolant Injection 1998-2012

    SciTech Connect

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the high-pressure coolant injection system (HPCI) at 69 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

  18. Aerodynamic effect of coolant ejection in the rear part of transonic rotor blades

    NASA Astrophysics Data System (ADS)

    Kost, F. H.; Holmes, A. T.

    1985-09-01

    An investigation of transonic turbine blades designed by Rolls-Royce/Bristol concerning the aerodynamic penalties of coolant flow for two alternative cooling configurations is discussed. Rolls-Royce designed a blade with a thick trailing edge where the coolant is ejected through slots in the trailing edge and a second blade with a thin trailing edge where coolant is ejected through a row of holes on the pressure side and a row of holes on the suction side. Tests were performed in a plane cascade wind tunnel. The results indicate the sensitivity of the blade performance to cooling configuration and coolant flow rate. By combining measured data from blade surface and wake traverses it was possible to separate the various loss mechanisms. Therefore, the separate losses due to the momentum of the coolant, change of base pressure, and change of blade friction could be determined quantitatively as a function of coolant flow rate.

  19. Load following capability of CANDLE reactor by adjusting coolant operation condition

    SciTech Connect

    Sekimoto, Hiroshi; Nakayama, Sinsuke

    2012-06-06

    The load following capability of CANDLE reactor is investigated in the condition that the control rods are unavailable. Both sodium cooled metallic fuel fast reactor (SFR) and {sup 208}Pb cooled metallic fuel fast reactor (LFR) are investigated for their performance in power rate changing by changing its coolant operation condition; either coolant flow rate or coolant inlet temperature. The change by coolant flow rate is difficult especially for SFR because the maximum temperature criteria on cladding material may be violated. The power rate can be changed for its full range easily by changing the coolant temperature at the core inlet. LFR can reduce the same amount of power rate by smaller change of temperature than SFR. However, the coolant output temperature is generally decreased for this method and the thermal efficiency becomes worse.

  20. Load following capability of CANDLE reactor by adjusting coolant operation condition

    NASA Astrophysics Data System (ADS)

    Sekimoto, Hiroshi; Nakayama, Sinsuke

    2012-06-01

    The load following capability of CANDLE reactor is investigated in the condition that the control rods are unavailable. Both sodium cooled metallic fuel fast reactor (SFR) and 208Pb cooled metallic fuel fast reactor (LFR) are investigated for their performance in power rate changing by changing its coolant operation condition; either coolant flow rate or coolant inlet temperature. The change by coolant flow rate is difficult especially for SFR because the maximum temperature criteria on cladding material may be violated. The power rate can be changed for its full range easily by changing the coolant temperature at the core inlet. LFR can reduce the same amount of power rate by smaller change of temperature than SFR. However, the coolant output temperature is generally decreased for this method and the thermal efficiency becomes worse.

  1. Optimization of the water chemistry of the primary coolant at nuclear power plants with VVER

    SciTech Connect

    Barmin, L. F.; Kruglova, T. K.; Sinitsyn, V. P.

    2005-01-15

    Results of the use of automatic hydrogen-content meter for controlling the parameter of 'hydrogen' in the primary coolant circuit of the Kola nuclear power plant are presented. It is shown that the correlation between the 'hydrogen' parameter in the coolant and the 'hydrazine' parameter in the makeup water can be used for controlling the water chemistry of the primary coolant system, which should make it possible to optimize the water chemistry at different power levels.

  2. Analysis of automobile radiator performance with ethylene glycol/water and propylene glycol/water coolants

    SciTech Connect

    Gollin, M.; Bjork, D.

    1996-12-31

    The heat transfer and hydraulic performance of the following coolants was examined in five automobile radiators in a wind tunnel: 100% water; 100% propylene glycol; 70/30 propylene glycol/water (volume); 50/50 propylene glycol/water (volume); 70/30 ethylene glycol/water (volume); 50/50 ethylene glycol water (volume). The results of these studies are presented to demonstrate the relative performance of these coolant mixtures in terms of heat transfer, coolant pressure drop and radiator effectiveness for a range of coolant and air flowrates. It is concluded that the most effective of the coolants in transferring heat in the test radiators was water, followed by 50/50 ethylene glycol/water, 50/50 propylene glycol/water, 70/30 ethylene glycol/water, 70/30 propylene glycol and, finally, 100% propylene glycol. There will be a negligible differences between the performance of a radiator using a 50/50 propylene glycol/water coolant and a 50/50 ethylene glycol/water coolant. It is estimated that, with 50/50 propylene glycol coolant replacing 50/50 ethylene glycol/water, the temperature of the coolant throughout the cooling loop will increase by approximately 5%. The effect that the flow regime (fully turbulent/transition/laminar) has upon the performance of a given radiator/coolant combination was found to be significant. The design of the coolant passages in radiators can affect the onset of fully turbulent flow in the coolant passages in a radiator.

  3. Emergency cooling analysis for the loss of coolant malfunction

    NASA Technical Reports Server (NTRS)

    Peoples, J. A.

    1972-01-01

    This report examines the dynamic response of a conceptual space power fast-spectrum lithium cooled reactor to the loss of coolant malfunction and several emergency cooling concepts. The results show that, following the loss of primary coolant, the peak temperatures of the center most 73 fuel elements can range from 2556 K to the region of the fuel melting point of 3122 K within 3600 seconds after the start of the accident. Two types of emergency aftercooling concepts were examined: (1) full core open loop cooling and (2) partial core closed loop cooling. The full core open loop concept is a one pass method of supplying lithium to the 247 fuel pins. This method can maintain fuel temperature below the 1611 K transient damage limit but requires a sizable 22,680-kilogram auxiliary lithium supply. The second concept utilizes a redundant internal closed loop to supply lithium to only the central area of each hexagonal fuel array. By using this method and supplying lithium to only the triflute region, fuel temperatures can be held well below the transient damage limit.

  4. Viscosity of alumina nanoparticles dispersed in car engine coolant

    SciTech Connect

    Kole, Madhusree; Dey, T.K.

    2010-09-15

    The present paper, describes our experimental results on the viscosity of the nanofluid prepared by dispersing alumina nanoparticles (<50 nm) in commercial car coolant. The nanofluid prepared with calculated amount of oleic acid (surfactant) was tested to be stable for more than 80 days. The viscosity of the nanofluids is measured both as a function of alumina volume fraction and temperature between 10 and 50 C. While the pure base fluid display Newtonian behavior over the measured temperature, it transforms to a non-Newtonian fluid with addition of a small amount of alumina nanoparticles. Our results show that viscosity of the nanofluid increases with increasing nanoparticle concentration and decreases with increase in temperature. Most of the frequently used classical models severely under predict the measured viscosity. Volume fraction dependence of the nanofluid viscosity, however, is predicted fairly well on the basis of a recently reported theoretical model for nanofluids that takes into account the effect of Brownian motion of nanoparticles in the nanofluid. The temperature dependence of the viscosity of engine coolant based alumina nanofluids obeys the empirical correlation of the type: log ({mu}{sub nf}) = A exp(BT), proposed earlier by Namburu et al. (author)

  5. Development of mobile, on-site engine coolant recycling utilizing reverse-osmosis technology

    SciTech Connect

    Kughn, W.; Eaton, E.R.

    1999-08-01

    This paper presents the history of the development of self-contained, mobile, high-volume, engine coolant recycling by reverse osmosis (R/O). It explains the motivations, created by government regulatory agencies, to minimize the liability of waste generators who produce waste engine coolant by providing an engine coolant recycling service at the customer`s location. Recycling the used engine coolant at the point of origin minimizes the generators` exposure to documentation requirements, liability, and financial burdens by greatly reducing the volume of used coolant that must be hauled from the generator`s property. It describes the inherent difficulties of recycling such a highly contaminated, inconsistent input stream, such as used engine coolant, by reverse osmosis. The paper reports how the difficulties were addressed, and documents the state of the art in mobile R/O technology. Reverse osmosis provides a purified intermediate fluid that is reinhibited for use in automotive cooling systems. The paper offers a review of experiences in various automotive applications, including light-duty, medium-duty and heavy-duty vehicles operating on many types of fuel. The authors conclude that mobile embodiments of R/O coolant recycling technology provide finished coolants that perform equivalently to new coolants as demonstrated by their ability to protect vehicles from freezing, corrosion damage, and other cooling system related problems.

  6. Zero waste machine coolant management strategy at Los Alamos National Laboratory

    SciTech Connect

    Carlson, B.; Algarra, F.; Wilburn, D.

    1998-12-01

    Machine coolants are used in machining equipment including lathes, grinders, saws and drills. The purpose of coolants is to wash away machinery debris in the form of metal fines, lubricate, and disperse heat between the part and the machine tool. An effective coolant prolongs tool life and protects against part rejection, commonly due to scoring or scorching. Traditionally, coolants have a very short effective life in the machine, often times being disposed of as frequently as once per week. The cause of coolant degradation is primarily due to the effects of bacteria, which thrive in the organic rich coolant environment. Bacteria in this environment reproduce at a logarithmic rate, destroying the coolant desirable aspects and causing potential worker health risks associated with the use of biocides to control the bacteria. The strategy described in this paper has effectively controlled bacterial activity without the use of biocides, avoided disposal of a hazardous waste, and has extend ed coolant life indefinitely. The Machine Coolant Management Strategy employed a combination of filtration, heavy lubricating oil removal, and aeration, which maintained the coolant peak performance without the use of biocides. In FY96, the Laboratory generated and disposed of 19,880 kg of coolants from 9 separate sites at a cost of $145K. The single largest generator was the main machine shop producing an average 14,000 kg annually. However, in FY97, the waste generation for the main machine shop dropped to 4,000 kg after the implementation of the zero waste strategy. It is expected that this value will be further reduced in FY98.

  7. Zero Waste Machine Coolant Management Strategy at Los Alamos National Laboratory

    SciTech Connect

    Carlson, B.; Algarra, F.; Wilburn, D.

    1998-06-01

    Machine coolants are used in machining equipment including lathes, grinders, saws and drills. The purpose of coolants is to wash away machinery debris in the form of metal fines, lubricate, and disperse heat between the part and the machine tool. An effective coolant prolongs tool life and protects against part rejection, commonly due to scoring or scorching. Traditionally, coolants have a very short effective life in the machine, often times being disposed of as frequently as once per week. The cause of coolant degradation is primarily due to the effects of bacteria, which thrive in the organic rich coolant environment. Bacteria in this environment reproduce at a logarithmic rate, destroying the coolant desirable aspects and causing potential worker health risks associated with the use of biocides to control the bacteria. The strategy described in this paper has effectively controlled bacterial activity without the use of biocides, avoided disposal of a hazardous waste, and has extend ed coolant life indefinitely. The Machine Coolant Management Strategy employed a combination of filtration, heavy lubricating oil removal, and aeration, which maintained the coolant peak performance without the use of biocides. In FY96, the Laboratory generated and disposed of 19,880 kg of coolants from 9 separate sites at a cost of $145K. The single largest generator was the main machine shop producing an average 14,000 kg annually. However, in FY97, the waste generation for the main machine shop dropped to 4,000 kg after the implementation of the zero waste strategy. It is expected that this value will be further reduced in FY98.

  8. Effect of coolant flow ejection on aerodynamic performance of low-aspect-ratio vanes. 2: Performance with coolant flow ejection at temperature ratios up to 2

    NASA Technical Reports Server (NTRS)

    Hass, J. E.; Kofskey, M. G.

    1977-01-01

    The aerodynamic performance of a 0.5 aspect ratio turbine vane configuration with coolant flow ejection was experimentally determined in a full annular cascade. The vanes were tested at a nominal mean section ideal critical velocity ratio of 0.890 over a range of primary to coolant total temperature ratio from 1.0 to 2.08 and a range of coolant to primary total pressure ratio from 1.0 to 1.4 which corresponded to coolant flows from 3.0 to 10.7 percent of the primary flow. The variations in primary and thermodynamic efficiency and exit flow conditions with circumferential and radial position were obtained.

  9. Robotic inspection of PWR coolant pump casing welds

    SciTech Connect

    Pratt, W.R.; Alford, J.W.; Davis, J.B.

    1997-12-01

    As of January 1, 1995, the Swedish Nuclear Inspectorate began requiring more thorough inspections of cast stainless-steel components in nuclear power plants, including pressurized water reactor (PWR) reactor coolant pump (RCP) casings. The examination requirements are established by fracture mechanics analyses of component weldments and demonstrated test system detection capabilities. This may include full volumetric inspection or some portion thereof. Ringhals station is a four-unit nuclear power plant, owned and operated by the Swedish State Power Board, Vattenfall. Unit 1 is a boiling water reactor. Units 2, 3, and 4 are Westinghouse-designed PWRs, ranging in size from 795 to 925 MW. The RCP casings at the PWR units are made of cast stainless steel and contain four circumferential welds that require inspection. Due to the thickness of the casings at the weld locations and configuration and surface conditions on the outside diameter of the casings, remote inspection from the inside diameter of the pump casing was mandated.

  10. Actively controlling coolant-cooled cold plate configuration

    DOEpatents

    Chainer, Timothy J.; Parida, Pritish R.

    2016-04-26

    Cooling apparatuses are provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The cooling apparatus includes the cold plate and a controller. The cold plate couples to one or more electronic components to be cooled, and includes an adjustable physical configuration. The controller dynamically varies the adjustable physical configuration of the cold plate based on a monitored variable associated with the cold plate or the electronic component(s) being cooled by the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the electronic component(s), and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the cold plate, the positioning of which may be adjusted based on the monitored variable.

  11. Leak rate analysis of the Westinghouse Reactor Coolant Pump

    SciTech Connect

    Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.

    1985-07-01

    An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs.

  12. Loss of coolant analysis for the tower shielding reactor 2

    SciTech Connect

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs.

  13. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    SciTech Connect

    DeMuth, J. A.; Meier, W. R.; Jolodosky, A.; Frantoni, M.; Reyes, S.

    2015-10-02

    The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

  14. Electrochemical pitting evaluation of aluminium alloy 7075 in machining coolant

    SciTech Connect

    Stanaland, V.A.; Dillon, J.J.

    1984-08-24

    The corrosion rate of aluminum alloy 7075 in Trim Sol with a Tris-Nitro biocide addition is satisfactory. Both deaeration and increasing the nitrite addition decreased the stability of the passive film. Chloride contamination below 500 ppM does not cause pitting corrosion of aluminum alloy 7075 in the Trim Sol environment. The limit for chloride contamination is between 500 and 1000 ppM. The potentiodynamic, fast-scan-rate technique is satisfactory for evaluating the pitting tendency of the aluminum alloy 7075 in a Trim Sol environment. Consequently, the potentiodynamic, fast-scan-rate technique is recommended for use in conjunction with reverse scans to evaluate the quality of in-use machining coolants, that are suspected of causing contamination.

  15. Investigation of an anion exchange resin for cleanup of a coolant used to machine nuclear materials

    SciTech Connect

    Hinton, E.R. Jr.; Tucker, H.L.; Asbury, W.L.

    1986-01-01

    This article describes the interaction of Dowex SBR-P, which is a strongly basic anion exchange resin, with ions found in a used machining coolant. The coolant is used in machining enriched uranium and contains uranium, chloride, nitrite, borate ions, water, and propylene glycol.

  16. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.46a Acceptance criteria for reactor coolant...

  17. Using automatic particle counting to monitor aluminum cold mill coolant{copyright}

    SciTech Connect

    Adkins, D.L.

    1995-08-01

    A comprehensive program of testing and evaluation of aluminum cold rolling coolant conditions has been conducted using an automatic particle counting instrument. The project had three objectives. First, there was a need to know at what level of coolant particle contamination is surface cleanliness of an aluminum sheet affected during the rolling process. Secondly, is application of particle counting technology a reliable tool for troubleshooting coolant filtration systems and finally, what are the advantages of analyzing rolling coolants for contamination levels? A testing program was designed and performed over a two-year period. The test results revealed that mineral seal and synthetic-type coolants can begin to affect aluminum sheet surface cleanliness levels when particle sizes greater than five microns are in excess of 10,000 particles power 100 milliliters of rolling coolant. After performing over 3,000 separate tests, it was very clean that particle count levels are direct indicators of how well a filtration facility is performing. Through the application of particle counting, a number of conditions in coolant filtration facilities can be readily detected. Such items as defective filter valving, torn or fractured filter cloth, damaged filter parts, improper equipment operation and many other factors will directly impact the operation of aluminum cold rolling coolant filters. 11 figs.

  18. Some peculiarities of checking the Topaz-2 system coolant filling quality

    SciTech Connect

    Ogloblin, B.G.; Svishchev, A.M.; Shalaev, A.I.

    1996-03-01

    This paper contains the analysis of validity of methods used for checking the Topaz-2 system coolant filling quality by a metering tank according to the mathematical model developed. A number of criteria is proposed for detecting occluded gas in the coolant loop. {copyright} {ital 1996 American Institute of Physics.}

  19. A method for measuring cooling air flow in base coolant passages of rotating turbine blades

    NASA Technical Reports Server (NTRS)

    Liebert, C. H.; Pollack, F. G.

    1975-01-01

    Method accurately determines actual coolant mass flow rate in cooling passages of rotating turbine blades. Total and static pressures are measured in blade base coolant passages. Mass flow rates are calculated from these measurements of pressure, measured temperature and known area.

  20. Coolant flows in prismatic fuel and particle bed nuclear reactors for rocket applications

    NASA Astrophysics Data System (ADS)

    Bohachevsky, Ihor O.

    1993-01-01

    Semiempirical expressions for pressure losses in prismatic and particle bed reactors for nuclear propulsion are combined with the geometric characteristics of core configurations and coolant flow patterns. The results are used to illustrate a limitation on the coolant velocity and to develop a unified approach to a quantitative comparison of merits and demerits of different reactor core concepts intended for space applications.

  1. Investigations of ice formation in the Space Shuttle Main Engine 0209 main injector coolant cavity

    NASA Technical Reports Server (NTRS)

    Richards, D. R.; Charklwick, D. M.

    1991-01-01

    Severe main combustion chamber wall and main injector baffle element deterioration occurred during tests of Space Shuttle Main Engine 0209. One of the possible causes considered is ice formation and blockage of coolant to these components, resulting from the mixing of leaking hot turbine exhaust gas (hydrogen rich steam) and hydrogen coolant in the injector coolant cavity. The plausibility of ice blockage is investigated through simple mixing calculations for hot gas and hydrogen, investigation of condensation and water droplet formation, calculation of the freezing times for droplets, and the prediction of ice layer thicknesses. It is concluded that condensation and droplet formation can occur, and small water droplets that form can freeze very quickly when in contact with the cold coolant cavity surfaces. Copnservative analysis predicts, however, that the maximum thickness of the ice layers formed is too small to result in significant blockage of the coolant flow.

  2. Engine coolant compatibility with the nonmetals found in automotive cooling systems

    SciTech Connect

    Greaney, J.P.; Smith, R.A.

    1999-08-01

    High temperature, short term immersion testing was used to determine the impact of propylene and ethylene glycol base coolants on the physical properties of a variety of elastomeric and thermoplastic materials found in automotive cooling systems. The materials tested are typically used in cooling system hoses, radiator end tanks, and water pump seals. Traditional phosphate or borate-buffered silicated coolants as well as extended-life organic acid formulations were included. A modified ASTM protocol was used to carry out the testing both in the laboratory and at an independent testing facility. Post-test fluid chemistry including an analysis of any solids which may have formed is also reported. Coolant impact on elastomer integrity as well as elastomer-induced changes in fluid chemistry were found to be independent of the coolant`s glycol base.

  3. Solar receiver protection means and method for loss of coolant flow

    DOEpatents

    Glasgow, L.E.

    1980-11-24

    An apparatus and method are disclosed for preventing a solar receiver utilizing a flowing coolant liquid for removing heat energy therefrom from overheating after a loss of coolant flow. Solar energy is directed to the solar receiver by a plurality of reflectors which rotate so that they direct solar energy to the receiver as the earth rotates. The apparatus disclosed includes a first storage tank for containing a first predetermined volume of the coolant and a first predetermined volume of gas at a first predetermined pressure. The first storage tank includes an inlet and outlet through which the coolant can enter and exit. The apparatus also includes a second storage tank for containing a second predetermined volume of the coolant and a second predetermined volume of the gas at a second predetermined pressure, the second storage tank having an inlet through which the coolant can enter. The first and second storage tanks are in fluid communication with each other through the solar receiver. The first and second predetermined coolant volumes, the first and second gas volumes, and the first and second predetermined pressures are chosen so that a predetermined volume of the coolant liquid at a predetermined rate profile will flow from the first storage tank through the solar receiver and into the second storage tank. Thus, in the event of a power failure so that coolant flow ceases and the solar reflectors stop rotating, a flow rate maintained by the pressure differential between the first and second storage tanks will be sufficient to maintain the coolant in the receiver below a predetermined upper temperature until the solar reflectors become defocused with respect to the solar receiver due to the earth's rotation.

  4. Solar receiver protection means and method for loss of coolant flow

    DOEpatents

    Glasgow, Lyle E.

    1983-01-01

    An apparatus and method for preventing a solar receiver (12) utilizing a flowing coolant liquid for removing heat energy therefrom from overheating after a loss of coolant flow. Solar energy is directed to the solar receiver (12) by a plurality of reflectors (16) which rotate so that they direct solar energy to the receiver (12) as the earth rotates. The apparatus disclosed includes a first storage tank (30) for containing a first predetermined volume of the coolant and a first predetermined volume of gas at a first predetermined pressure. The first storage tank (30) includes an inlet and outlet through which the coolant can enter and exit. The apparatus also includes a second storage tank (34) for containing a second predetermined volume of the coolant and a second predetermined volume of the gas at a second predetermined pressure, the second storage tank (34) having an inlet through which the coolant can enter. The first and second storage tanks (30) and (34) are in fluid communication with each other through the solar receiver (12). The first and second predetermined coolant volumes, the first and second gas volumes, and the first and second predetermined pressures are chosen so that a predetermined volume of the coolant liquid at a predetermined rate profile will flow from the first storage tank (30) through the solar receiver (12) and into the second storage tank (34). Thus, in the event of a power failure so that coolant flow ceases and the solar reflectors (16) stop rotating, a flow rate maintained by the pressure differential between the first and second storage tanks (30) and (34) will be sufficient to maintain the coolant in the receiver (12) below a predetermined upper temperature until the solar reflectors (16) become defocused with respect to the solar receiver (12) due to the earth's rotation.

  5. Characterization of uranium surfaces machined with aqueous propylene glycol-borax or perchloroethylene-mineral oil coolants

    SciTech Connect

    Cristy, S.S.; Bennett, R.K. Jr.; Dillon, J.J.; Richards, H.L.; Seals, R.D.; Byrd, V.R.

    1986-12-31

    The use of perchloroethylene (perc) as an ingredient in coolants for machining enriched uranium at the Oak Ridge Y-12 Plant has been discontinued because of environmental concerns. A new coolant was substituted in December 1985, which consists of an aqueous solution of propylene glycol with borax (sodium tetraborate) added as a nuclear poison and with a nitrite added as a corrosion inhibitor. Uranium surfaces machined using the two coolants were compared with respects to residual contamination, corrosion or corrosion potential, and with the aqueous propylene glycol-borax coolant was found to be better than that of enriched uranium machined with the perc-mineral oil coolant. The boron residues on the final-finished parts machined with the borax-containing coolant were not sufficient to cause problems in further processing. All evidence indicated that the enriched uranium surfaces machined with the borax-containing coolant will be as satisfactory as those machined with the perc coolant.

  6. Flow in serpentine coolant passages with trip strips

    NASA Technical Reports Server (NTRS)

    Tse, D. G.-N.

    1995-01-01

    Under the subject contract, an effort is being conducted at Scientific Research Associates, Inc. (SRA) to obtain flow field measurements in the coolant passage of a rotating turbine blade with ribbed walls, both in the stationary and rotating frames. The data obtained will be used for validation of computational tools and assessment of turbine blade cooling strategies. The configuration of the turbine blade passage model is given, and the measuring plane locations are given. The model has a four-pass passage with three 180 turns. This geometry was chosen to allow analyses of the velocity measurements corresponding to the heat transfer results obtained by Wagner. Two passes of the passage have a rectangular cross-section of 1.0 in x 0.5 in. Another two passes have a square cross-section of 0.5 in x 0.5 in. Trips with a streamwise pitch to trip height (P/e) = 5 and trip height to coolant passage width (e/Z) = 0.1, were machined along the leading and trailing walls. These dimensions are typical of those used in turbine blade coolant passages. The trips on these walls are staggered by the half-pitch. The trips are skewed at +/- 45 deg, and this allows the effect of trip orientation to be examined. Experiments will be conducted with flow entering the model through the 1.0 in x 0.5 in rectangular passage (Configuration C) and the 0.5 in x 0. 5 in square passage (Configuration D) to examine the effect of passage aspect ratio. Velocity measurements were obtained with a Reynolds number (Re) of 25,000, based on the hydraulic diameter of and bulk mean velocity in the half inch square passage. The coordinate system used in presenting the results for configurations C and D, respectively, is shown. The first, second and third passes of the passage will be referred to as the first, second and third passages, respectively, in later discussion. Streamwise distance (x) from the entrance is normalized by the hydraulic diameter (D). Vertical (y) and tangential (z) distances are

  7. Purification of liquid metal systems with sodium coolant from oxygen using getters

    NASA Astrophysics Data System (ADS)

    Kozlov, F. A.; Konovalov, M. A.; Sorokin, A. P.

    2016-05-01

    For increasing the safety and economic parameters of nuclear power stations (NPSs) with sodium coolant, it was decided to install all systems contacting radioactive sodium, including purification systems of circuit I, in the reactor vessel. The performance and capacity of cold traps (CTs) (conventional element of coolant purification systems) in these conditions are limited by their volume. It was proposed to use hot traps (HTs) in circuit I for coolant purification from oxygen. It was demonstrated that, at rated parameters of the installation when the temperature of the coolant streamlining the getter (gas absorber) is equal to 550°C, the hot trap can provide the required coolant purity. In shutdown modes at 250-300°C, the performance of the hot trap is reduced by four orders of magnitude. Possible HT operation regimes for shutdown modes and while reaching rated parameters were proposed and analyzed. Basic attention was paid to purification modes at power rise after commissioning and accidental contamination of the coolant when the initial oxygen concentration in it reached 25 mln-1. It was demonstrated that the efficiency of purification systems can be increased using HTs with the getter in the form of a foil or granules. The possibility of implementing the "fast purification" mode in which the coolant is purified simultaneously with passing over from the shutdown mode to the rated parameters was substantiated.

  8. Polymer electrolyte membrane fuel cell performance degradation by coolant leakage and recovery

    NASA Astrophysics Data System (ADS)

    Jung, Ju Hae; Kim, Se Hoon; Hur, Seung Hyun; Joo, Sang Hoon; Choi, Won Mook; Kim, Junbom

    2013-03-01

    Coolant leakage leads to decrease in performance during the operation of electric vehicles which make use of polymer electrolyte membrane fuel cells (PEMFC). This study examines the effects of various coolant leak conditions in 3-cell stack and single cell. The experimental results show that an irreversible reduction in performance occurs after coolant injection into the anode side of the stack. Poisoning of carbon monoxide (CO) on the platinum (Pt) catalyst is caused by electro-oxidation reaction of EG. Water cleaning is selected because CO poisoning is desorbed to reaction with water molecules. Performance is quickly reduced when the interval between coolant injections is short. Performance reduction is indicated by the experimental results for the gas diffusion layer (GDL) and the membrane electrode assembly (MEA). It shows that performance of the MEA with the GDL exposed to coolant decreased, but it is recovered after water cleaning. In contrast, results for performance of the MEA exposed to coolant for long time could not be reversed after water cleaning. Therefore, we propose that performance degradation of coolant leak on the Pt catalyst surface and GDL can be recovered by the water cleaning simply without disassembly of stack.

  9. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    SciTech Connect

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  10. Selection of an Alternate Biocide for the ISS Internal Thermal Control System Coolant, Phase 2

    NASA Technical Reports Server (NTRS)

    Wilson, Mark E.; Cole, Harold; Weir, Natalee; Oehler, Bill; Steele, John; Varsik, Jerry; Lukens, Clark

    2004-01-01

    The ISS (International Space Station) ITCS (Internal Thermal Control System) includes two internal coolant loops that utilize an aqueous based coolant for heat transfer. A silver salt biocide had previously been utilized as an additive in the coolant formulation to control the growth and proliferation of microorganisms within the coolant loops. Ground-based and in-flight testing demonstrated that the silver salt was rapidly depleted, and did not act as an effective long-term biocide. Efforts to select an optimal alternate biocide for the ITCS coolant application have been underway and are now in the final stages. An extensive evaluation of biocides was conducted to down-select to several candidates for test trials and was reported on previously. Criteria for that down-select included: the need for safe, non-intrusive implementation and operation in a functioning system; the ability to control existing planktonic and biofilm residing microorganisms; a negligible impact on system-wetted materials of construction; and a negligible reactivity with existing coolant additives. Candidate testing to provide data for the selection of an optimal alternate biocide is now in the final stages. That testing has included rapid biocide effectiveness screening using Biolog MT2 plates to determine minimum inhibitory concentration (amount that will inhibit visible growth of microorganisms), time kill studies to determine the exposure time required to completely eliminate organism growth, materials compatibility exposure evaluations, coolant compatibility studies, and bench-top simulated coolant testing. This paper reports the current status of the effort to select an alternate biocide for the ISS ITCS coolant. The results of various test results to select the optimal candidate are presented.

  11. Cracked shaft detection on large vertical nuclear reactor coolant pump

    NASA Technical Reports Server (NTRS)

    Jenkins, L. S.

    1985-01-01

    Due to difficulty and radiation exposure associated with examination of the internals of large commercial nuclear reactor coolant pumps, it is necessary to be able to diagnose the cause of an excessive vibration problem quickly without resorting to extensive trial and error efforts. Consequently, it is necessary to make maximum use of all available data to develop a consistent theory which locates the problem area in the machine. This type of approach was taken at Three Mile Island, Unit #1, in February 1984 to identify and locate the cause of a continuously climbing vibration level of the pump shaft. The data gathered necessitated some in-depth knowledge of the pump internals to provide proper interpretation and avoid misleading conclusions. Therefore, the raw data included more than just the vibration characteristics. Pertinent details of the data gathered is shown and is necessary and sufficient to show that the cause of the observed vibration problem could logically only be a cracked pump shaft in the shaft overhang below the pump bearing.

  12. Tables of thermodynamic properties of helium magnet coolant. Revision A

    SciTech Connect

    McAshan, M.

    1992-07-01

    The most complete treatment of the thermodynamic properties of helium at the present time is the monograph by McCarty: ``Thermodynamic Properties of Helium 4 from 2 to 1500 K at Pressures to 10{sup 8} Pa``, Robert D. McCarty, Journal of Physical and Chemical Reference Data, Vol. 2, page 923--1040 (1973). In this work the complete range of data on helium is examined and the P-V-T surface is described by an equation of state consisting of three functions P(r,T) covering different regions together with rules for making the transition from one region to another. From this thermodynamic compilation together with correlations of the transport properties of helium was published the well-known NBS Technical Note: ``Thermophysical Properties of Helium 4 from 2 to 1500 K with pressures to 1000 Atmospheres``, Robert D. McCarty, US Department of Commerce, National Bureau of Standards Technical Note 631 (1972). This is the standard reference for helium cryogenics. The NBS 631 tables cover a wide range of temperature and pressure, and as a consequence, the number of points tabulated in the region of the single phase coolant for the SSC magnets are relatively few. The present work sets out to cover the range of interest in more detail in a way that is consistent with NBS 631. This new table is essentially identical to the older one and can be used as an auxiliary to it.

  13. Tables of thermodynamic properties of helium magnet coolant

    SciTech Connect

    McAshan, M.

    1992-07-01

    The most complete treatment of the thermodynamic properties of helium at the present time is the monograph by McCarty: Thermodynamic Properties of Helium 4 from 2 to 1500 K at Pressures to 10{sup 8} Pa'', Robert D. McCarty, Journal of Physical and Chemical Reference Data, Vol. 2, page 923--1040 (1973). In this work the complete range of data on helium is examined and the P-V-T surface is described by an equation of state consisting of three functions P(r,T) covering different regions together with rules for making the transition from one region to another. From this thermodynamic compilation together with correlations of the transport properties of helium was published the well-known NBS Technical Note: Thermophysical Properties of Helium 4 from 2 to 1500 K with pressures to 1000 Atmospheres'', Robert D. McCarty, US Department of Commerce, National Bureau of Standards Technical Note 631 (1972). This is the standard reference for helium cryogenics. The NBS 631 tables cover a wide range of temperature and pressure, and as a consequence, the number of points tabulated in the region of the single phase coolant for the SSC magnets are relatively few. The present work sets out to cover the range of interest in more detail in a way that is consistent with NBS 631. This new table is essentially identical to the older one and can be used as an auxiliary to it.

  14. Tables of thermodynamic properties of helium magnet coolant, revision A

    NASA Astrophysics Data System (ADS)

    McAshan, M.

    1992-07-01

    The most complete treatment of the thermodynamic properties of helium at the present time is the monograph by McCarty: 'Thermodynamic Properties of Helium 4 from 2 to 1500 K at Pressures to 10(exp 8) Pa', Robert D. McCarty, Journal of Physical and Chemical Reference Data, Vol. 2, page 923-1040 (1973). In this work the complete range of data on helium is examined and the P-V-T surface is described by an equation of state consisting of three functions P(r,T) covering different regions together with rules for making the transition from one region to another. From this thermodynamic compilation together with correlations of the transport properties of helium was published the well-known NBS Technical Note: 'Thermophysical Properties of Helium 4 from 2 to 1500 K with pressures to 1000 Atmospheres', Robert D. McCarty, US Department of Commerce, National Bureau of Standards Technical Note 631 (1972). This is the standard reference for helium cryogenics. The NBS 631 tables cover a wide range of temperature and pressure, and as a consequence, the number of points tabulated in the region of the single phase coolant for the SSC magnets are relatively few. The present work sets out to cover the range of interest in more detail in a way that is consistent with NBS 631. This new table is essentially identical to the older one and can be used as an auxiliary to it.

  15. Numerical study: Iron corrosion-resistance in lead-bismuth eutectic coolant by molecular dynamics method

    SciTech Connect

    Arkundato, Artoto; Su'ud, Zaki; Abdullah, Mikrajuddin; Widayani,; Celino, Massimo

    2012-06-06

    In this present work, we report numerical results of iron (cladding) corrosion study in interaction with lead-bismuth eutectic coolant of advanced nuclear reactors. The goal of this work is to study how the oxygen can be used to reduce the corrosion rate of cladding. The molecular dynamics method was applied to simulate corrosion process. By evaluating the diffusion coefficients, RDF functions, MSD curves of the iron and also observed the crystal structure of iron before and after oxygen injection to the coolant then we concluded that a significant and effective reduction can be achieved by issuing about 2% number of oxygen atoms to lead-bismuth eutectic coolant.

  16. Effects of rotation on coolant passage heat transfer. Volume 2: Coolant passages with trips normal and skewed to the flow

    NASA Technical Reports Server (NTRS)

    Johnson, B. V.; Wagner, J. H.; Steuber, G. D.

    1993-01-01

    An experimental program was conducted to investigate heat transfer and pressure loss characteristics of rotating multipass passages, for configurations and dimensions typical of modem turbine blades. This experimental program is one part of the NASA Hot Section Technology (HOST) Initiative, which has as its overall objective the development and verification of improved analysis methods that will form the basis for a design system that will produce turbine components with improved durability. The objective of this program was the generation of a data base of heat transfer and pressure loss data required to develop heat transfer correlations and to assess computational fluid dynamic techniques for rotating coolant passages. The experimental work was broken down into two phases. Phase 1 consists of experiments conducted in a smooth wall large scale heat transfer model. A detailed discussion of these results was presented in volume 1 of a NASA Report. In Phase 2 the large scale model was modified to investigate the effects of skewed and normal passage turbulators. The results of Phase 2 along with comparison to Phase 1 is the subject of this Volume 2 NASA Report.

  17. Computer code for predicting coolant flow and heat transfer in turbomachinery

    NASA Technical Reports Server (NTRS)

    Meitner, Peter L.

    1990-01-01

    A computer code was developed to analyze any turbomachinery coolant flow path geometry that consist of a single flow passage with a unique inlet and exit. Flow can be bled off for tip-cap impingement cooling, and a flow bypass can be specified in which coolant flow is taken off at one point in the flow channel and reintroduced at a point farther downstream in the same channel. The user may either choose the coolant flow rate or let the program determine the flow rate from specified inlet and exit conditions. The computer code integrates the 1-D momentum and energy equations along a defined flow path and calculates the coolant's flow rate, temperature, pressure, and velocity and the heat transfer coefficients along the passage. The equations account for area change, mass addition or subtraction, pumping, friction, and heat transfer.

  18. Sodium coolant purification systems for a nuclear power station equipped with a BN-1200 reactor

    NASA Astrophysics Data System (ADS)

    Alekseev, V. V.; Kovalev, Yu. P.; Kalyakin, S. G.; Kozlov, F. A.; Kumaev, V. Ya.; Kondrat'ev, A. S.; Matyukhin, V. V.; Pirogov, E. P.; Sergeev, G. P.; Sorokin, A. P.; Torbenkova, I. Yu.

    2013-05-01

    Both traditional coolant purification methods (by means of traps and sorbents for removing cesium), the use of which supported successful operation of nuclear power installations equipped with fast-neutron reactors with a sodium coolant, and the possibility of removing oxygen from sodium through the use of hot traps are analyzed in substantiating the purification system for a nuclear power station equipped with a BN-1200 reactor. It is shown that a cold trap built into the reactor vessel must be a mandatory component of the reactor plant primary coolant circuit's purification system. The use of hot traps allows oxygen to be removed from the sodium coolant down to permissible concentrations when the nuclear power station operates in its rated mode. The main lines of works aimed at improving the performance characteristics of cold traps are suggested based on the results of performed investigations.

  19. Hot-Gas-Slide and Coolant-Side Heat Transfer in Liquid Rocket Engine Combustors

    NASA Technical Reports Server (NTRS)

    Wang, Ten-See; Luong, Van

    1994-01-01

    The objectives of this article are to develop a multidisciplinary, computational methodology to predict the hot-gas-side and coolant-side heat transfer in film cooling assisted, regeneratively cooled liquid rocket engine combustors, and to use it in parametric studies to recommend optimized design of the coolant channels for a developmental combustor. An integrated numerical model which incorporates computational fluid dynamics (CFD) for the hot-gas thermal environment, and thermal analysis for the liner and coolant channels, was developed. This integrated CFD/thermal model was validated by comparing predicted heat fluxes with those of hot-firing test and industrial design methods for a 40-k calorimeter thrust chamber and the Space Shuttle Main Engine main combustion chamber. Parametric studies were performed for the advanced main combustion chamber to find a strategy for a proposed coolant channel design.

  20. Thermal transfer structures coupling electronics card(s) to coolant-cooled structure(s)

    DOEpatents

    David, Milnes P; Graybill, David P; Iyengar, Madhusudan K; Kamath, Vinod; Kochuparambil, Bejoy J; Parida, Pritish R; Schmidt, Roger R

    2014-12-16

    Cooling apparatuses and coolant-cooled electronic systems are provided which include thermal transfer structures configured to engage with a spring force one or more electronics cards with docking of the electronics card(s) within a respective socket(s) of the electronic system. A thermal transfer structure of the cooling apparatus includes a thermal spreader having a first thermal conduction surface, and a thermally conductive spring assembly coupled to the conduction surface of the thermal spreader and positioned and configured to reside between and physically couple a first surface of an electronics card to the first surface of the thermal spreader with docking of the electronics card within a socket of the electronic system. The thermal transfer structure is, in one embodiment, metallurgically bonded to a coolant-cooled structure and facilitates transfer of heat from the electronics card to coolant flowing through the coolant-cooled structure.

  1. 120. COOLANT LINES TO SIS HEAT EXCHANGER No.1 IN AUXILIARY ...

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

    120. COOLANT LINES TO SIS HEAT EXCHANGER No.1 IN AUXILIARY CHAMBER, NOVEMBER 1, 1976 - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

  2. The experience in handling of lead-bismuth coolant contaminated by Polonium-210

    SciTech Connect

    Pankratov, D.V.; Gromov, B.F.; Solodjankin, M.A.

    1993-12-31

    During exploitation of lead-bismuth cooled reactors a wide experience in handling of radioactive coolant containing polonium has been gained. By 1990 total time of this reactor operation has reached approximately 60 reactor years.

  3. Reactor coolant pump testing using motor current signatures analysis

    SciTech Connect

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  4. Correct numerical simulation of a two-phase coolant

    NASA Astrophysics Data System (ADS)

    Kroshilin, A. E.; Kroshilin, V. E.

    2016-02-01

    Different models used in calculating flows of a two-phase coolant are analyzed. A system of differential equations describing the flow is presented; the hyperbolicity and stability of stationary solutions of the system is studied. The correctness of the Cauchy problem is considered. The models' ability to describe the following flows is analyzed: stable bubble and gas-droplet flows; stable flow with a level such that the bubble and gas-droplet flows are observed under and above it, respectively; and propagation of a perturbation of the phase concentration for the bubble and gas-droplet media. The solution of the problem about the breakdown of an arbitrary discontinuity has been constructed. Characteristic times of the development of an instability at different parameters of the flow are presented. Conditions at which the instability does not make it possible to perform the calculation are determined. The Riemann invariants for the nonlinear problem under consideration have been constructed. Numerical calculations have been performed for different conditions. The influence of viscosity on the structure of the discontinuity front is studied. Advantages of divergent equations are demonstrated. It is proven that a model used in almost all known investigating thermohydraulic programs, both in Russia and abroad, has significant disadvantages; in particular, it can lead to unstable solutions, which makes it necessary to introduce smoothing mechanisms and a very small step for describing regimes with a level. This does not allow one to use efficient numerical schemes for calculating the flow of two-phase currents. A possible model free from the abovementioned disadvantages is proposed.

  5. Cladding embrittlement during postulated loss-of-coolant accidents.

    SciTech Connect

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  6. Analysis of Loss-of-Coolant Accidents in the NBSR

    SciTech Connect

    Baek J. S.; Cheng L.; Diamond, D.

    2014-05-23

    This report documents calculations of the fuel cladding temperature during loss-of-coolant accidents in the NBSR. The probability of a pipe failure is small and procedures exist to minimize the loss of water and assure emergency cooling water flows into the reactor core during such an event. Analysis in the past has shown that the emergency cooling water would provide adequate cooling if the water filled the flow channels within the fuel elements. The present analysis is to determine if there is adequate cooling if the water drains from the flow channels. Based on photographs of how the emergency water flows into the fuel elements from the distribution pan, it can be assumed that this water does not distribute uniformly across the flow channels but rather results in a liquid film flowing downward on the inside of one of the side plates in each fuel element and only wets the edges of the fuel plates. An analysis of guillotine breaks shows the cladding temperature remains below the blister temperature in fuel plates in the upper section of the fuel element. In the lower section, the fuel plates are also cooled by water outside the element that is present due to the hold-up pan and temperatures are lower than in the upper section. For small breaks, the simulation results show that the fuel elements are always cooled on the outside even in the upper section and the cladding temperature cannot be higher than the blister temperature. The above results are predicated on assumptions that are examined in the study to see their influence on fuel temperature.

  7. Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

    SciTech Connect

    Chen, N.C.J.; Yoder, G.L. ); Wendel, M.W. )

    1991-01-01

    Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs.

  8. Measurement of the Coolant Channel Temperatures and Pressures of a Cooled Radial-Inflow Turbine

    NASA Technical Reports Server (NTRS)

    Dicicco, L. Danielle; Nowlin, Brent C.; Tirres, Lizet

    1994-01-01

    Instrumentation has been installed on the surface of a cooled radial-inflow turbine. Thermocouples and miniature integrated sensor pressure transducers were installed to measure steady state coolant temperatures, blade wall temperatures, and coolant pressures. These measurements will eventually be used to determine the heat transfer characteristics of the rotor. This paper will describe the procedures used to install and calibrate the instrumentation and the testing methods followed. A limited amount of data will compare the measured values to the predicted values.

  9. Influence of coolant tube curvature on film cooling effectiveness as detected by infrared imagery

    NASA Technical Reports Server (NTRS)

    Papell, S. S.; Graham, R. W.; Cageao, R. P.

    1979-01-01

    Thermal film cooling footprints observed by infrared imagery from straight, curved, and looped coolant tube geometries are compared. It was hypothesized that the differences in secondary flow and in the turbulence structure of flow through these three tubes should influence the mixing properties between the coolant and the main stream. A flow visualization tunnel, an infrared camera and detector, and a Hilsch tube were employed to test the hypothesis.

  10. Impact of the propylene glycol-water-borax coolant on material recovery operations

    SciTech Connect

    Duerksen, W.K.; Taylor, P.A.

    1983-05-01

    The reaction of the propylene glycol-water-borax coolant with nitric acid has now been studied in some detail. This document is intended to provide a summary of the results. Findings are summarized under nine headings. Tests have also been conducted to determine if the new coolant would have any adverse effects on the uranium recycle systems. Experiments were scientifically designed after observation of the production operations so that accurate response to the immediate production concerns could be provided. Conclusions from these studies are: formation of glycol nitrates is very improbable; the reaction of concentrated (70%) nitric acid with pure propylene glycol is very violent and hazardous; dilution of the nitric acid-glycol mixture causes a drastic decrease in the rate and intensity of the reaction; the mechanism of the nitric acid propylene glycol reaction is autocatalytic in nitrous acid; no reaction is observed between coolant and 30% nitric acid unless the solution is heated; the coolant reacts fairly vigorously with 55% nitric acid after a concentration-dependent induction time; experiments showed that the dissolution of uranium chips that had been soaked in coolant proceeded at about the same rate as if the chips had not previously contacted glycol; thermodynamic calculations show that the enthalpy change (heat liberated) by the reaction of nitric acid (30%) with propylene glycol is smaller than if the same amount of nitric acid reacted with uranium. Each of these conclusions is briefly discussed. The effect of new coolant on uranium recycle operations is then briefly discussed.

  11. Modeling electrodeposition of charged nanoparticles onto fuel cell coolant flow channel walls

    NASA Astrophysics Data System (ADS)

    Cheng, J.-T.

    2008-10-01

    To cool down the stack system in polymer electrolyte fuel cells (PEFCs), a coolant is needed that must be electrically nonconductive. In the specialized coolant that is modeled by us, charged nanoparticles are added into the flow to neutralize the ion contamination that otherwise gradually degrades the coolant until shunt currents become significant. A computational fluid dynamics (CFD) physicochemical model of the multiphase coolant flow with charged nanoparticles has been formulated and coded using COMSOL Multiphysics and MEMS. Electrochemistry, fluid mechanics, steric stabilization, and heat transfer are coupled in this model. For nanoparticles in the fluid, electrokinetic force, electrical double layer (EDL) force, hydrodynamic force, and buoyancy force have been taken into account for the prediction of the electrodeposition rate onto channel walls. The overall goal of the model is to provide a fundamental first principles-based design tool for a specialized coolant to enable operations in a fuel cell stack for 2-3 years without the need for frequent replacement or filtering of the coolant.

  12. Analysis of a water-coolant leak into a very high-temperature vitrification chamber.

    SciTech Connect

    Felicione, F. S.

    1998-06-11

    A coolant-leakage incident occurred during non-radioactive operation of the Plasma Hearth Process waste-vitrification development system at Argonne National Laboratory when a stray electric arc ruptured az water-cooling jacket. Rapid evaporation of the coolant that entered the very high-temperature chamber pressurized the normally sub-atmospheric system above ambient pressure for over 13 minutes. Any positive pressurization, and particularly a lengthy one, is a safety concern since this can cause leakage of contaminants from the system. A model of the thermal phenomena that describe coolant/hot-material interactions was developed to better understand the characteristics of this type of incident. The model is described and results for a variety of hypothetical coolant-leak incidents are presented. It is shown that coolant leak rates above a certain threshold will cause coolant to accumulate in the chamber, and evaporation from this pool can maintain positive pressure in the system long after the leak has been stopped. Application of the model resulted in reasonably good agreement with the duration of the pressure measured during the incident. A closed-form analytic solution is shown to be applicable to the initial leak period in which the peak pressures are generated, and is presented and discussed.

  13. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project -2006 Update

    NASA Technical Reports Server (NTRS)

    Morrison, Russell H.; Holt, Mike

    2006-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. This paper presents a status of the coolant stability over the past year as well as results from destructive analyses of hardware removed from the on-orbit system and the current approach to coolant remediation.

  14. Thermal criteria to compare fast reactors coolants for the intermediate loop

    SciTech Connect

    Saez, Manuel; Rodriguez, Gilles

    2007-07-01

    Fast Breeder Reactors (FBR) are typically using a liquid metal as the primary coolant. Up to now, sodium is the referenced coolant for all large-scale FBR, but lead and sodium-potassium alloy have both also been used successfully for smaller rigs. The French Atomic Energy Commission (CEA) has an extensive experience and significant expertise in Sodium cooled Fast Reactors (SFR) over the past 40 years of R and D and feedback experiments. Some improvements are needed on the SFR to meet the Generation IV goals, and in particular the safety and the reliability through the intermediate loop coolant. As sodium reacts exo-thermically with air and water and to eliminate the drawback of the water-sodium interaction when a steam generator tube is ruptured, CEA is involved in a substantial effort in order to investigate the interest to use an alternative coolant than sodium in the intermediate loop. This paper presents the main thermal criteria to compare Fast Reactors coolants for the intermediate loop under natural and forced convection. Neutronics considerations are not taken into account for the intermediate loop coolant. Transport, transfer and energetic criteria are analysed in the field of turbulent flows. Criteria are applied to the following potential coolant candidates: sodium, lithium, tin, bismuth, lead, lead-bismuth alloy, lead-lithium alloy, gallium, indium, potassium and sodium-potassium alloy. According to this thermal analysis, the gallium as heat transfer agent for the intermediate loop is considered as a promising candidate. For the discussion of the applicability of the gallium as heat transfer agent for the intermediate loop, a limited thermal hydraulic pre-sizing of a steam generator is undertaken using simple engineering methods implemented in COPERNIC code, a CEA tool dedicated to reactor systems pre-sizing. (authors)

  15. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    SciTech Connect

    Ruger, C.J.; Higgins, J.C.

    1993-11-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970`s and early 1980`s raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants.

  16. Coolant-side heat-transfer rates for a hydrogen-oxygen rocket and a new technique for data correlation

    NASA Technical Reports Server (NTRS)

    Schacht, R. L.; Quentmeyer, R. J.

    1973-01-01

    An experimental investigation was conducted to determine the coolant-side, heat transfer coefficients for a liquid cooled, hydrogen-oxygen rocket thrust chamber. Heat transfer rates were determined from measurements of local hot gas wall temperature, local coolant temperature, and local coolant pressure. A correlation incorporating an integration technique for the transport properties needed near the pseudocritical temperature of liquid hydrogen gives a satisfactory prediction of hot gas wall temperatures.

  17. Numerical analysis of the hot-gas-side and coolant-side heat transfer in liquid rocket engine combustors

    NASA Technical Reports Server (NTRS)

    Wang, T. S.; Luong, V.

    1992-01-01

    The objectives of this paper are to develop computational methods to predict the hot-gas-side and coolant-side heat transfer, and to use these methods in parametric studies to recommend optimized design of the coolant channels for regeneratively cooled liquid rocket engine combustors. An integrated numerical model which incorporates computational fluid dynamics (CFD) for the hot-gas thermal environment, and thermal analysis for the coolant channels, was developed. The mode was validated by comparing predicted heat fluxes with those of hot-firing test and industrial design methods. Parametric studies were performed to find a strategy for optimized combustion chamber coolant channel design.

  18. Measurement of Coolant in a Flat Heat Pipe Using Neutron Radiography

    NASA Astrophysics Data System (ADS)

    Mizuta, Kei; Saito, Yasushi; Goshima, Takashi; Tsutsui, Toshio

    A newly developed flat heat pipe FGHPTM (Morex Kiire Co.) was experimentally investigated by using neutron radiography. The test sample of the FGHP heat spreader was 65 × 65 × 2 mm3 composed of several etched copper plates and pure water was used as the coolant. Neutron radiography was performed at the E-2 port of the Kyoto University Research Reactor (KUR). The coolant distributions in the wick area of the FGHP and its heat transfer characteristics were measured at heating conditions. Experimental results show that the coolant distributions depend slightly on its installation posture and that the liquid thickness in the wick region remains constant with increasing heat input to the FGHP. In addition, it is found that the wick surface does not dry out even in the vertical posture at present experimental conditions.

  19. System and method for determining coolant level and flow velocity in a nuclear reactor

    DOEpatents

    Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

    2013-09-10

    A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

  20. Effects of coolant parameters on steady state temperature distribution in phospheric-acid fuel cell electrode

    NASA Technical Reports Server (NTRS)

    Alkasab, K. A.; Abdul-Aziz, A.

    1991-01-01

    The influence of thermophysical properties and flow rate on the steady-state temperature distribution in a phosphoric-acid fuel cell electrode plate was experimentally investigated. An experimental setup that simulates the operating conditions prevailing in a phosphoric-acid fuel cell stack was used. The fuel cell cooling system utilized three types of coolants to remove excess heat generated in the cell electrode and to maintain a reasonably uniform temperature distribution in the electrode plate. The coolants used were water, engine oil, and air. These coolants were circulated at Reynolds number ranging from 1165 to 6165 for water; 3070 to 6864 for air; and 15 to 79 for oil. Experimental results are presented.

  1. Nanomaterials for efficiently lowering the freezing point of anti-freeze coolants.

    PubMed

    Hong, Haiping; Zheng, Yingsong; Roy, Walter

    2007-09-01

    In this paper, we report, for the first time, the effect of the lowered freezing point in a 50% water/50% anti-freeze coolant (PAC) or 50% water/50% ethylene glycol (EG) solution by the addition of carbon nanotubes and other particles. The experimental results indicated that the nano materials are much more efficient (hundreds fold) in lowering the freezing point than the regular ionic materials (e.g., NaCl). The possible explanation for this interesting phenomenon is the colligative property of fluid and relative small size of nano material. It is quite certain that the carbon nanotubes and metal oxide nano particles could be a wonderful candidate for the nano coolant application because they could not only increase the thermal conductivity, but also efficiently lower the freezing point of traditional coolants. PMID:18019146

  2. New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

    NASA Astrophysics Data System (ADS)

    Rezania, A.; Rosendahl, L. A.

    2012-06-01

    The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The three-dimensional governing equations for the fluid flow and the heat transfer are solved using the finite-volume method for a wide range of pressure drop laminar flows along the heat sink. The temperature and the mass flow rate distribution in the heat sink are discussed. The results, which are in good agreement with previous computational studies, show that using suggested heat sink configurations reduces the coolant pumping power in the system.

  3. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

    SciTech Connect

    Su'ud, Zaki; Anshari, Rio

    2012-06-06

    Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

  4. Preliminary analysis of loss-of-coolant accident in Fukushima nuclear accident

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Anshari, Rio

    2012-06-01

    Loss-of-Coolant Accident (LOCA) in Boiling Water Reactor (BWR) especially on Fukushima Nuclear Accident will be discussed in this paper. The Tohoku earthquake triggered the shutdown of nuclear power reactors at Fukushima Nuclear Power station. Though shutdown process has been completely performed, cooling process, at much smaller level than in normal operation, is needed to remove decay heat from the reactor core until the reactor reach cold-shutdown condition. If LOCA happen at this condition, it will cause the increase of reactor fuel and other core temperatures and can lead to reactor core meltdown and exposure of radioactive material to the environment such as in the Fukushima Dai Ichi nuclear accident case. In this study numerical simulation has been performed to calculate pressure composition, water level and temperature distribution on reactor during this accident. There are two coolant regulating system that operational on reactor unit 1 at this accident, Isolation Condensers (IC) system and Safety Relief Valves (SRV) system. Average mass flow of steam to the IC system in this event is 10 kg/s and could keep reactor core from uncovered about 3,2 hours and fully uncovered in 4,7 hours later. There are two coolant regulating system at operational on reactor unit 2, Reactor Core Isolation Condenser (RCIC) System and Safety Relief Valves (SRV). Average mass flow of coolant that correspond this event is 20 kg/s and could keep reactor core from uncovered about 73 hours and fully uncovered in 75 hours later. There are three coolant regulating system at operational on reactor unit 3, Reactor Core Isolation Condenser (RCIC) system, High Pressure Coolant Injection (HPCI) system and Safety Relief Valves (SRV). Average mass flow of water that correspond this event is 15 kg/s and could keep reactor core from uncovered about 37 hours and fully uncovered in 40 hours later.

  5. Development of a cleaning process for uranium chips machined with a glycol-water-borax coolant

    SciTech Connect

    Taylor, P.A.

    1984-12-01

    A chip-cleaning process has been developed to remove the new glycol-water-borax coolant from oralloy chips. The process involves storing the freshly cut chips in Freon-TDF until they are cleaned, washing with water, and displacing the water with Freon-TDF. The wash water can be reused many times and still yield clean chips and then be added to the coolant to make up for evaporative losses. The Freon-TDF will be cycled by evaporation. The cleaning facility is currently being designed and should be operational by April 1985.

  6. Flow boiling with enhancement devices for cold plate coolant channel design

    NASA Technical Reports Server (NTRS)

    Boyd, Ronald D., Sr.

    1989-01-01

    A research program to study the effect of enhancement devices on flow boiling heat transfer in coolant channels, which are heated either from the top side or uniformly, is discussed. Freon 11 is the working fluid involved. The specific objectives are: (1) examine the variations in both the mean and local (axial and circumferential) heat transfer coefficients for a circular coolant channel with either smooth walls or with both a twisted tape and spiral finned walls, (2) examine the effect channel diameter (and the length-to-diameter aspect ratio) variations for the smooth wall channel, and (3) develop an improved data reduction analysis.

  7. TRAC loss-of-coolant accident analyses of the Savannah River production reactors

    SciTech Connect

    Lime, J.F.; Motley, F.E. )

    1990-06-01

    TRAC loss-of-coolant accident (LOCA) analyses were performed as part of the independent safety review of the US Department of Energy's Savannah River (SR) production reactors. The double-ended guillotine break in a coolant loop is a design-basis LOCA for the SR reactors. Three break locations were analyzed to determine the worst break location: (1) at the pump-suction flange; (2) at the pump discharge flange; or (3) at the plenum inlet. The plenum-inlet break was shown to be the most severe in terms of minimum flow delivered to each fuel assembly in the reactor core.

  8. Probabilistic analyses of failure in reactor coolant piping. [Double-ended guillotine break

    SciTech Connect

    Holman, G.S.

    1984-07-20

    LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB).

  9. Loss-of-coolant accident experiment at the AVR (Arbeitsgemeinschaft Versuchsreaktor) gas-cooled reactor

    SciTech Connect

    Krueger, K. ); Cleveland, J. )

    1989-11-01

    Loss of coolant is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into modular gas-cooled reactor designs, loss-of-coolant accident (LOCA) simulation tests were conducted with the 15-MW(electric), 46-MW(thermal), pebble-bed, high-temperature Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany (FRG). This is the only nuclear power plant ever to have been intentionally subjected to LOCa conditions. Oak Ridge National Laboratory participation in the preparation and conduct of the tests was carried out within the U.S./FRG Agreement for Cooperation in Gas-Cooled Reactor Development.

  10. Coolant and ambient temperature control for chillerless liquid cooled data centers

    DOEpatents

    Chainer, Timothy J.; David, Milnes P.; Iyengar, Madhusudan K.; Parida, Pritish R.; Simons, Robert E.

    2016-02-02

    Cooling control methods include measuring a temperature of air provided to a plurality of nodes by an air-to-liquid heat exchanger, measuring a temperature of at least one component of the plurality of nodes and finding a maximum component temperature across all such nodes, comparing the maximum component temperature to a first and second component threshold and comparing the air temperature to a first and second air threshold, and controlling a proportion of coolant flow and a coolant flow rate to the air-to-liquid heat exchanger and the plurality of nodes based on the comparisons.

  11. Design, manufacture, and test of coolant pump-motor assembly for Brayton power conversion system

    NASA Technical Reports Server (NTRS)

    Gabacz, L. E.

    1973-01-01

    The design, development, fabrication, and testing of seven coolant circulating pump-motor assemblies are discussed. The pump-motor assembly is driven by the nominal 44.4-volt, 400-Hz, 3-phase output of a nominal 56-volt dc input inverter. The pump-motor assembly will be used to circulate Dow Corning 200 liquid coolant for use in a Brayton cycle space power system. The pump-motor assembly develops a nominal head of 70 psi at 3.7 gpm with an over-all efficiency of 26 percent. The design description, drawings, photographs, reliability results, and developmental and acceptance test results are included.

  12. Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

    SciTech Connect

    Young, Michael F.

    1999-05-01

    The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks.

  13. Integrated Fuel-Coolant Interaction (IFCI 6.0) code. User`s manual

    SciTech Connect

    Davis, F.J.; Young, M.F.

    1994-04-01

    The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User`s Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks.

  14. Polonium release from an ATW burner system with liquid lead-bismuth coolant

    SciTech Connect

    Li, N.; Yefimov, E.; Pankratov, D.

    1998-04-01

    The authors analyzed polonium release hazards in a conceptual pool-type ATW burner with liquid lead-bismuth eutectic (LBE) coolant. Simplified quantitative models are used based on experiments and real NPP experience. They found little Po contamination outside the burner under normal operating conditions with nominal leakage from the gas system. In sudden gas leak and/or coolant spill accidents, the P contamination level can reach above the regulation limit but short exposure would not lead to severe health consequences. They are evaluating and developing mitigation methods.

  15. In-core coolant flow monitoring of pressurized water reactors using temperature and neutron noise

    SciTech Connect

    Sweeney, F.J.; Upadhyaya, B.R.; Shieh, D.J.

    1984-01-01

    Noise measurements were performed at the Loss-of-Fluid-Test (LOFT) and Sequoyah-1 pressurized water reactors (PWRs) in order to investigate the possibility of inferring in-core coolant velocities from cross-power spectral density (CPSD) phases of core-exit thermocouple and in-core neutron detector signals. These noise measurements were used to investigate the effects of inlet coolant temperature, core flow, reactor power, and random heat transfer fluctuations on the noise-inferred coolant velocities. The effect on the inferred velocities of varying in-core neutron detector and core-exit thermocouple locations was also investigated. Theoretical models of temperature noise were developed, and the results were used to interpret the experimental measurements. Results of these studies indicate that the neutron detector/thermocouple phase is useful for monitoring core flow in PWRs. Results show that the interpretation of the phase between these signals depends on the source of temperature noise, the response times and locations of the sensors, and the neutron dynamics of the reactor. At Sequoyah-1 we found that the in-core neutron detector/core-exit thermocouple phase can be used to infer in-core coolant velocities, provided that the measurements are corrected for the thermocouple response time.

  16. MTR, TRA603. SUBBASEMENT FLOOR PLAN. INLET/OUTLET TUNNELS FOR COOLANT WATER ...

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

    MTR, TRA-603. SUB-BASEMENT FLOOR PLAN. INLET/OUTLET TUNNELS FOR COOLANT WATER (NORTH SIDE) AND AIR (SOUTH SIDE). RABBIT CANAL AND BULKHEADS. SUMPS AND DRAINS. BLAW-KNOX 3150-3-7, 3/1950. INL INDEX NO. 531-0603-00-098-100006, REV. 4. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  17. Method and apparatus for removing iodine from a nuclear reactor coolant

    DOEpatents

    Cooper, Martin H.

    1980-01-01

    A method and apparatus for removing iodine-131 and iodine-125 from a liquid sodium reactor coolant. Non-radioactive iodine is dissolved in hot liquid sodium to increase the total iodine concentration. Subsequent precipitation of the iodine in a cold trap removes both the radioactive iodine isotopes as well as the non-radioactive iodine.

  18. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... operated from the control room. (b) The design of the vents and associated controls, instruments and power sources must conform to appendix A and appendix B of this part. (c) The vent system must be designed to... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with...

  19. Inverse design of coolant flow passage shapes with partially fixed internal geometries

    NASA Astrophysics Data System (ADS)

    Kennon, S. R.; Dulikravich, G. S.

    1985-03-01

    A method has been developed for the design and analysis of complex coolant flow passage shapes in internally cooled turbine engines. The method is particularly applicable to turbine airfoil cascade inverse design but may also be applied to the design of other nonadiabatic boundaries such as missile cone tips and internal combustion engine cylinder. The method makes it possible to specify and fix the temperature or the heat flux at the turbine airfoil outer surface, together with the desired temperature at the coolant/blade interfaces. Coupled with an appropriate flow solver and stress analysis code, the method provides accurate estimates of the blade surface temperature and heat flux distribution in the coolant passage. A first order panel method is used to solve the Laplace equations for steady heat conduction within the solid portions of the hollow blade. In order to illustrate the efficiency of the method, numerical results are presented for the case of a turbine airfoil having three coolant holes. The discretized temperature contours for the inner and outer parts of the blade are illustrated in graphic form.

  20. Loss-of-coolant accident experiment at the AVR gas-cooled reactor

    SciTech Connect

    Cleveland, J.; Krueger, K.; Kernforschungsanlage Juelich G.m.b.H. . Arbeitsgemeinschaft Versuchsreaktor)

    1989-01-01

    Loss-of-coolant is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into small High-Temperature Gas Cooled Reactor (HTGR) designs, loss-of-coolant accident (LOCA) simulation tests have been conducted with the German pebble-bed High-Temperature Reactor AVR. The AVR is the only nuclear power plant ever to have been intentionally subjected to LOCA conditions. The LOCA test was planned to create conditions that would exist if a rapid LOCA occurred with the reactor operating at full power. The tests demonstrated this reactor's safe response to an accident in which the coolant escapes from the reactor core and no emergency system is available to provide coolant flow to the core. The test is of special interest because it demonstrates the inherent safety features incorporated into modular HTGR designs. The main LOCA test lasted for 5 d. After the test began, core temperatures increased for {approximately}13 h and then gradually and continually decreased as the rate of heat dissipation from the core exceeded accident levels of decay power. Throughout the test, temperatures remained below limiting values for the core and other reactor components. 3 refs., 9 figs., 1 tab.

  1. Loss-of-coolant accident experiment at the AVR gas-cooled reactor

    SciTech Connect

    Krueger, K. ); Cleveland, J. )

    1990-01-01

    A landmark safety test has been conducted at the AVR-reactor, a high-temperature gas-cooled reactor (HTGR) in the Federal Republic of Germany owned by the Arbeitsgemeinschaft Versuchsreaktor, AVR in Juelich. The 46-MW(t), 15-MW(e) AVR reactor was subjected to a simulated loss-of-coolant accident (LOCA), a very severe occurrence in which the coolant escapes from the reactor core and no emergency system provides coolant flow to the core. The test, which demonstrated the inherently safe response of this reactor to a LOCA, marked the first time ever that a reactor has been intentionally subjected to loss-of-coolant conditions without emergency cooling. Oak Ridge National Laboratory (ORNL) and General Atomics participated in the test by working with AVR staff by jointly performing the analyses needed to obtain the license to conduct the test and by performing post test analyses. This participation was carried out under the cooperative AVR Subprogram which is conducted within the US/FRG Agreement for Cooperation in Gas-Cooled Reactor Development. 7 figs.

  2. Method of and apparatus for removing silicon from a high temperature sodium coolant

    DOEpatents

    Yunker, Wayne H.; Christiansen, David W.

    1987-01-01

    A method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

  3. Method of and apparatus for removing silicon from a high temperature sodium coolant

    DOEpatents

    Yunker, Wayne H.; Christiansen, David W.

    1987-05-05

    A method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

  4. Modelling the activity of 129I in the primary coolant of a CANDU reactor

    NASA Astrophysics Data System (ADS)

    Lewis, Brent J.; Husain, Aamir

    2003-01-01

    A mathematical treatment has been developed to describe the activity levels of 129I as a function of time in the primary heat transport system during constant power operation and for a reactor shutdown situation. The model accounts for a release of fission-product iodine from defective fuel rods and tramp uranium contamination on in-core surfaces. The physical transport constants of the model are derived from a coolant activity analysis of the short-lived radioiodine species. An estimate of 3×10 -9 has been determined for the coolant activity ratio of 129I/ 131I in a CANDU Nuclear Generating Station (NGS), which is in reasonable agreement with that observed in the primary coolant and for plant test resin columns from pressurized and boiling water reactor plants. The model has been further applied to a CANDU NGS, by fitting it to the observed short-lived iodine and long-lived cesium data, to yield a coolant activity ratio of ˜2×10 -8 for 129I/ 137Cs. This ratio can be used to estimate the levels of 129I in reactor waste based on a measurement of the activity of 137Cs.

  5. TRITIUM LABORATORY, TRA666, INTERIOR. COOLANT LOOP PIPING DETAIL AND CONTROL ...

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

    TRITIUM LABORATORY, TRA-666, INTERIOR. COOLANT LOOP PIPING DETAIL AND CONTROL VALVE EQUIPMENT ALONG EAST WALL. INL NEGATIVE NO. HD30-2-2. Mike Crane, Photographer, 6/2001 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  6. A New Application of Support Vector Machine Method: Condition Monitoring and Analysis of Reactor Coolant Pump

    NASA Astrophysics Data System (ADS)

    Meng, Qinghu; Meng, Qingfeng; Feng, Wuwei

    2012-05-01

    Fukushima nuclear power plant accident caused huge losses and pollution and it showed that the reactor coolant pump is very important in a nuclear power plant. Therefore, to keep the safety and reliability, the condition of the coolant pump needs to be online condition monitored and fault analyzed. In this paper, condition monitoring and analysis based on support vector machine (SVM) is proposed. This method is just to aim at the small sample studies such as reactor coolant pump. Both experiment data and field data are analyzed. In order to eliminate the noise and useless frequency, these data are disposed through a multi-band FIR filter. After that, a fault feature selection method based on principal component analysis is proposed. The related variable quantity is changed into unrelated variable quantity, and the dimension is descended. Then the SVM method is used to separate different fault characteristics. Firstly, this method is used as a two-kind classifier to separate each two different running conditions. Then the SVM is used as a multiple classifier to separate all of the different condition types. The SVM could separate these conditions successfully. After that, software based on SVM was designed for reactor coolant pump condition analysis. This software is installed on the reactor plant control system of Qinshan nuclear power plant in China. It could monitor the online data and find the pump mechanical fault automatically.

  7. ETR COOLING TOWER PUMP HOUSE, TRA645. FOUR SECONDARY COOLANT PUMPS ...

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

    ETR COOLING TOWER PUMP HOUSE, TRA-645. FOUR SECONDARY COOLANT PUMPS ARE ARRANGED IN A ROW. IN REAR ARE THREE SHUTDOWN EMERGENCY PUMPS. INL NEGATIVE NO. 56-4176. Jack L. Anderson, Photographer, 12/21/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  8. Lead Coolant Test Facility Systems Design, Thermal Hydraulic Analysis and Cost Estimate

    SciTech Connect

    Soli Khericha; Edwin Harvego; John Svoboda; Ryan Dalling

    2012-01-01

    The Idaho National Laboratory prepared a preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed: (1) Develop and Demonstrate Feasibility of Submerged Heat Exchanger; (2) Develop and Demonstrate Open-lattice Flow in Electrically Heated Core; (3) Develop and Demonstrate Chemistry Control; (4) Demonstrate Safe Operation; and (5) Provision for Future Testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimate. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

  9. Effect of glycol-based coolants on the suppression and recovery of platinum fuel cell electrocatalysts

    NASA Astrophysics Data System (ADS)

    Garsany, Yannick; Dutta, Sreya; Swider-Lyons, Karen E.

    2012-10-01

    We use cyclic and rotating disk electrode voltammetry to study glycol-based coolant formulations to show that individual constituents have either negligible or significant poisoning effects on the nanoscale Pt/carbon catalysts used in proton exchange membrane fuel cells. The base fluid in all these coolants is glycol (1, 3 propanediol), commercially available in a BioGlycol coolant formulation with an ethoxylated nonylphenol surfactant, and azole- and polyol-based non-ionic corrosion inhibitors. Exposure of a Pt/Vulcan carbon electrode to glycol-water or glycol-water-surfactant mixtures causes the loss of Pt electrochemical surface area (ECSA), but the Pt ECSA is fully recovered in clean electrolyte. Only mixtures with the azole corrosion inhibitor cause irreversible losses to the Pt ECSA and oxygen reduction reaction (ORR) activity. The Pt ECSA and ORR activity can only be recovered to within 70% of its initial values after aggressive voltammetric cycling to 1.50 V after azole poisoning. When poisoned with a glycol mixture containing the polyol corrosion inhibitor instead, the Pt ECSA and ORR activity is completely recovered by exposure to a clean electrolyte. The results suggest that prior to incorporation in a fuel cell, voltammetric evaluation of the constituents of coolant formulations is worthwhile.

  10. ETR HEAT EXCHANGER BUILDING, TRA644. A PRIMARY COOLANT PUMP AND ...

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

    ETR HEAT EXCHANGER BUILDING, TRA-644. A PRIMARY COOLANT PUMP AND 24-INCH CHECK VALVE ARE MOUNTED IN A SHIELDED CUBICLE. NOTE CONNECTION AT RIGHT THROUGH SHIELD WALL TO PUMP MOTOR ON OTHER SIDE. INL NEGATIVE NO. 56-4177. Jack L. Anderson, Photographer, 12/21/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  11. An Improved Design for Air Removal from Aerospace Fluid Loop Coolant Systems

    NASA Technical Reports Server (NTRS)

    Ritchie, Stephen M. C.; Holladay, Jon B.; Holt, J. Mike; Clark, Dallas W.

    2003-01-01

    Aerospace applications with requirements for large capacity heat removal (launch vehicles, platforms, payloads, etc.) typically utilize a liquid coolant fluid as a transport media to increase efficiency and flexibility in the vehicle design. An issue with these systems however, is susceptibility to the presence of noncondensable gas (NCG) or air. The presence of air in a coolant loop can have numerous negative consequences, including loss of centrifugal pump prime, interference with sensor readings, inhibition of heat transfer, and coolant blockage to remote systems. Hardware ground processing to remove this air is also cumbersome and time consuming which continuously drives recurring costs. Current systems for maintaining the system free of air are tailored and have demonstrated only moderate success. An obvious solution to these problems is the development and advancement of a passive gas removal device, or gas trap, that would be installed in the flight cooling system simplifying the initial coolant fill procedure and also maintaining the system during operations. The proposed device would utilize commercially available membranes thus increasing reliability and reducing cost while also addressing both current and anticipated applications. In addition, it maintains current pressure drop, water loss, and size restrictions while increasing tolerance for pressure increases due to gas build-up in the trap.

  12. Desensitization of over tip leakage in an axial turbine rotor by tip surface coolant injection

    NASA Astrophysics Data System (ADS)

    Rao, Nikhil Molahally

    Mechanical energy extraction in axial flow turbine rotors occurs through a change in angular momentum of the working fluid. The gap between the turbine rotor and the stationary casing is referred to as the tip gap. High pressure turbine blades are typically un-shrouded and pressure driven flow through the tip gap is termed as over tip leakage. Over tip leakage reduces efficiency of the turbine stage and also causes thermal distress to blade tip surfaces. The gap height typically increases over the operational life of a turbine, leading to increased efficiency drop. The thermal load on the tip surface also increases with increasing gap height and is exacerbated by the radial transport of high temperature fluid found in the core of the combustor exit flow. Thus over tip leakage not only decreases stage efficiency, but also constrains it by limiting the maximum cycle temperature. Reducing the sensitivity of turbine performance to the effects of the tip gap is termed Tip Desensitization. An experimental investigation of tip desensitization through coolant injection from a tip surface trench was conducted in a large scale, low speed, rotating research turbine facility. Five out of twenty nine rotor blades, referred to as cooled blades, are provided with coolant injection at four locations, at 61%, 71%, 81%, and 91% blade tip axial chord length. At each of the first three locations the coolant jets are directed towards the blade pressure-side, while coolant is exhausted radially at the last location. The sensitivity of total pressure defect, due to over tip leakage, to tip gap height is reduced by both coolant injection and roughening of the casing surface. The total pressure defect due to the large gap height of 1.40% blade height is reduced to levels comparable to the defect due to a gap height of 0.72% blade height. The strong total pressure gradient that characterizes the leakage vortex due to the gap height of 1.40% blade height is considerably diminished by both

  13. Barriers to the Application of High-Temperature Coolants in Hybrid Electric Vehicles

    SciTech Connect

    Hsu, J.S.; Staunton, M.R.; Starke, M.R.

    2006-09-30

    This study was performed by the Oak Ridge National Laboratory (ORNL) to identify practical approaches, technical barriers, and cost impacts to achieving high-temperature coolant operation for certain traction drive subassemblies and components of hybrid electric vehicles (HEV). HEVs are unique in their need for the cooling of certain dedicated-traction drive subassemblies/components that include the electric motor(s), generators(s), inverter, dc converter (where applicable), and dc-link capacitors. The new coolant system under study would abandon the dedicated 65 C coolant loop, such as used in the Prius, and instead rely on the 105 C engine cooling loop. This assessment is important because automotive manufacturers are interested in utilizing the existing water/glycol engine cooling loop to cool the HEV subassemblies in order to eliminate an additional coolant loop with its associated reliability, space, and cost requirements. In addition, the cooling of power electronic devices, traction motors, and generators is critical in meeting the U.S. Department of Energy (DOE) FreedomCAR and Vehicle Technology (FCVT) goals for power rating, volume, weight, efficiency, reliability, and cost. All of these have been addressed in this study. Because there is high interest by the original equipment manufacturers (OEMs) in reducing manufacturing cost to enhance their competitive standing, the approach taken in this analysis was designed to be a positive 'can-do' approach that would be most successful in demonstrating the potential or opportunity of relying entirely on a high-temperature coolant system. Nevertheless, it proved to be clearly evident that a few formidable technical and cost barriers exist and no effective approach for mitigating the barriers was evident in the near term. Based on comprehensive thermal tests of the Prius reported by ORNL in 2005 [1], the continuous ratings at base speed (1200 rpm) with different coolant temperatures were projected from test data at

  14. Barriers to the Application of High-Temperature Coolants in Hybrid Electric Vehicles

    SciTech Connect

    Staunton, Robert H; Hsu, John S; Starke, Michael R

    2006-09-01

    This study was performed by the Oak Ridge National Laboratory (ORNL) to identify practical approaches, technical barriers, and cost impacts to achieving high-temperature coolant operation for certain traction drive subassemblies and components of hybrid electric vehicles (HEV). HEVs are unique in their need for the cooling of certain dedicated-traction drive subassemblies/components that include the electric motor(s), generators(s), inverter, dc converter (where applicable), and dc-link capacitors. The new coolant system under study would abandon the dedicated 65 C coolant loop, such as used in the Prius, and instead rely on the 105 C engine cooling loop. This assessment is important because automotive manufacturers are interested in utilizing the existing water/glycol engine cooling loop to cool the HEV subassemblies in order to eliminate an additional coolant loop with its associated reliability, space, and cost requirements. In addition, the cooling of power electronic devices, traction motors, and generators is critical in meeting the U.S. Department of Energy (DOE) FreedomCAR and Vehicle Technology (FCVT) goals for power rating, volume, weight, efficiency, reliability, and cost. All of these have been addressed in this study. Because there is high interest by the original equipment manufacturers (OEMs) in reducing manufacturing cost to enhance their competitive standing, the approach taken in this analysis was designed to be a positive 'can-do' approach that would be most successful in demonstrating the potential or opportunity of relying entirely on a high-temperature coolant system. Nevertheless, it proved to be clearly evident that a few formidable technical and cost barriers exist and no effective approach for mitigating the barriers was evident in the near term. Based on comprehensive thermal tests of the Prius reported by ORNL in 2005 [1], the continuous ratings at base speed (1200 rpm) with different coolant temperatures were projected from test data at

  15. Rapid electrochemical screening of engine coolants. Correlation of electrochemical potentiometric measurements with ASTM D 1384 glassware corrosion test

    SciTech Connect

    Doucet, G.P.; Jackson, J.M.; Kriegel, O.A.; Passwater, D.K.; Prieto, N.E.

    1999-08-01

    Engine coolants are typically subjected to comprehensive performance evaluations that involve multiple laboratory and field tests. These tests can take several weeks to conduct and can be expensive. The tests can involve everything from preliminary chemical screening to long term fleet tests. An important test conducted at the beginning of coolant formula development to screen the corrosion performance of engine coolants is described in ASTM D 1384. If the coolant formula passes the test, it is then subjected to more rigorous testing. Conducting the test described in ASTM D 1384 takes two weeks, and determining the coolant corrosion performance under several test parameters can takes resources and time that users seldom have. Therefore, it is very desirable to have tests that can be used for rapid screening and quality assurance of coolants. The purpose of this study was to conduct electrochemical tests that can ultimately be used for quick initial screening of engine coolants. The specific intent of the electrochemical tests is to use ASTM D 1384 as a model and to attempt to duplicate its results. Implementation of the electrochemical tests could accelerate the process of selecting promising coolant formulas and reduce coolant evaluation time and cost. Various electrochemical tests were conducted to determine the corrosion performance of several engine coolant formulas. The test results were compared to those obtained from the ASTM D 1384 test. These tests were conducted on the same metal specimens and under similar conditions as those used in the ASTM D 1384 test. The electrochemical tests included the determination of open circuit potential (OCP) for the various metal specimens, anodic and cathodic polarization curves for the various metal specimens, corrosion rate for metal specimens involved in a galvanic triad, and critical pitting potential (CPP) for aluminum (pitting of aluminum engine components and cooling systems is a cause for concern). The details for

  16. Investigating Liquid CO2 as a Coolant for a MTSA Heat Exchanger Design

    NASA Technical Reports Server (NTRS)

    Paul, Heather L.; Padilla, Sebastian; Powers, Aaron; Iacomini, Christie

    2009-01-01

    Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO 2) control for a future Portable Life Support System (PLSS), as well as water recycling. CO 2 removal and rejection is accomplished by driving a sorbent through a temperature swing of approximately 210 K to 280 K . The sorbent is cooled to these sub-freezing temperatures by a Sublimating Heat Exchanger (SHX) with liquid coolant expanded to sublimation temperatures. Water is the baseline coolant available on the moon, and if used, provides a competitive solution to the current baseline PLSS schematic. Liquid CO2 (LCO2) is another non-cryogenic coolant readily available from Martian resources which can be produced and stored using relatively low power and minimal infrastructure. LCO 2 expands from high pressure liquid (5800 kPa) to Mars ambient (0.8 kPa) to produce a gas / solid mixture at temperatures as low as 156 K. Analysis and experimental work are presented to investigate factors that drive the design of a heat exchanger to effectively use this sink. Emphasis is given to enabling efficient use of the CO 2 cooling potential and mitigation of heat exchanger clogging due to solid formation. Minimizing mass and size as well as coolant delivery are also considered. The analysis and experimental work is specifically performed in an MTSA-like application to enable higher fidelity modeling for future optimization of a SHX design. In doing so, the work also demonstrates principles and concepts so that the design can be further optimized later in integrated applications (including Lunar application where water might be a choice of coolant).

  17. Polonium Issue in Fast Reactor Lead Coolants and One of the Ways of Its Solution

    SciTech Connect

    Khorasanov, G.L.; Ivanov, A.P.; Blokhin, A.I.

    2002-07-01

    One of the main issues in using materials for nuclear facilities is to minimize the production of the most hazardous radionuclides. In the ideal case, all nuclear reactor materials, except a fuel, should be low-activation. The term 'low-activation material' means that this one loses its induced activity in a short time after removal from irradiation. Proposals for building a fusion reactor using low-activation materials are given in Ref.1, 2. For this purpose, low-activation structural materials based on V-Ti-Cr alloys are in the stage of R and D in several countries [3,4]. Another technique to avoid the hazardous activity is in using isotopically enriched materials [5-7]. Although isotopic tailoring option requires tremendous technical efforts and it is too expensive, its application can be first of all assumed for those structural and functional materials which generate very hazardous radionuclides under irradiation. In modern projects of next generation NPPs the preference is given to fast reactors (FRs) with a lead coolant [8]. As it known, the coolant circulating through a FR core is activated, and in the future we should have problems with handling a completed coolant after FR decommissioning or at realization of repair or emergency activities. There, it is desirable to have a low-activation coolant with the low contents of hazardous radionuclides. In papers [9,10] presented at the previous ICONE conferences it was proposed to use lead isotope, Pb-206, as a coolant instead of lead natural, Pb-nat. This paper is devoted to more detailed calculations of accumulating stable bismuth, Bi-209, and polonium radioisotopes, Po-209 (T{sub 1/2}=102 y) and Po-210 (T{sub 1/2}=138 d), in 1 kg of Pb-nat or Pb-206 placed in the core of the BOR-60 type FR. (authors)

  18. Cooling Characteristics of the V-1650-7 Engine. II - Effect of Coolant Conditions on Cylinder Temperatures and Heat Rejection at Several Engine Powers

    NASA Technical Reports Server (NTRS)

    Povolny, John H.; Bogdan, Louis J.; Chelko, Louis J.

    1947-01-01

    An investigation has been conducted on a V-1650-7 engine to determine the cylinder temperatures and the coolant and oil heat rejections over a range of coolant flows (50 to 200 gal/min) and oil inlet temperatures (160 to 2150 F) for two values of coolant outlet temperature (250 deg and 275 F) at each of four power conditions ranging from approximately 1100 to 2000 brake horsepower. Data were obtained for several values of block-outlet pressure at each of the two coolant outlet temperatures. A mixture of 30 percent by volume of ethylene glycol and 70-percent water was used as the coolant. The effect of varying coolant flow, coolant outlet temperature, and coolant outlet pressure over the ranges investigated on cylinder-head temperatures was small (0 deg to 25 F) whereas the effect of increasing the engine power condition from ll00 to 2000 brake horsepower was large (maximum head-temperature increase, 110 F).

  19. TACT1- TRANSIENT THERMAL ANALYSIS OF A COOLED TURBINE BLADE OR VANE EQUIPPED WITH A COOLANT INSERT

    NASA Technical Reports Server (NTRS)

    Gaugler, R. E.

    1994-01-01

    As turbine-engine core operating conditions become more severe, designers must develop more effective means of cooling blades and vanes. In order to design reliable, cooled turbine blades, advanced transient thermal calculation techniques are required. The TACT1 computer program was developed to perform transient and steady-state heat-transfer and coolant-flow analyses for cooled blades, given the outside hot-gas boundary condition, the coolant inlet conditions, the geometry of the blade shell, and the cooling configuration. TACT1 can analyze turbine blades, or vanes, equipped with a central coolant-plenum insert from which coolant-air impinges on the inner surface of the blade shell. Coolant-side heat-transfer coefficients are calculated with the heat transfer mode at each station being user specified as either impingement with crossflow, forced convection channel flow, or forced convection over pin fins. A limited capability to handle film cooling is also available in the program. The TACT1 program solves for the blade temperature distribution using a transient energy equation for each node. The nodal energy balances are linearized, one-dimensional, heat-conduction equations which are applied at the wall-outer-surface node, at the junction of the cladding and the metal node, and at the wall-inner-surface node. At the mid-metal node a linear, three-dimensional, heat-conduction equation is used. Similarly, the coolant pressure distribution is determined by solving the set of transfer momentum equations for the one-dimensional flow between adjacent fluid nodes. In the coolant channel, energy and momentum equations for one-dimensional compressible flow, including friction and heat transfer, are used for the elemental channel length between two coolant nodes. The TACT1 program first obtains a steady-state solution using iterative calculations to obtain convergence of stable temperatures, pressures, coolant-flow split, and overall coolant mass balance. Transient

  20. Study of coolant activation and dose rates with flow rate and power perturbations in pool-type research reactors

    SciTech Connect

    Mirza, N.M.; Mirza, S.M.; Ahmad, N. )

    1991-12-01

    This paper reports on a computer code using the multigroup diffusion theory based LEOPARD and ODMUG programs that has been developed to calculate the activity in the coolant leaving the core of a pool-type research reactor. Using this code, the dose rates at various locations along the coolant path with varying coolant flow rate and reactor power perturbations are determined. A flow rate decrease from 1000 to 145 m{sup 3}/h is considered. The results indicate that a flow rate decrease leads to an increase in the coolant outlet temperature, which affects the neutron group constants and hence the group fluxes. The activity in the coolant leaving the core increases with flow rate decrease. However, at the inlet of the holdup tank, the total dose rate first increases, then passes through a maximum at {approximately} 500 m{sup 3}/h, and finally decreases with flow rate decrease. The activity at the outlet of the holdup tank is mainly due to {sup 24}Na and {sup 56}Mn, and it increases by {approximately} 2% when the flow rate decreases from 1000 to 145 m{sup 3}/h. In an accidental power rise at constant flow rate, the activity in the coolant increases, and the dose rates at all the points along the coolant path show a slight nonlinear rise as the reactor power density increases.

  1. Neutronic and Physical Characteristics of an Accelerator Driven System with a Lead-208 Coolant

    SciTech Connect

    Khorasanov, Georgy L.; Ivonov, Anatoly P.; Blokhin, Anatoly I.

    2006-07-01

    In the paper a possibility of using a lead isotope, pure Pb-208, as a coolant for a subcritical core of 80 MW thermal capacity of the PDS-XADS type facility is considered. Calculations of neutronic characteristics were performed using Monte Carlo technique. The following initial data were chosen: an annular core with a target, as a neutron source, at its centre; the core coolant -- Pb-208 (100%); a fuel -- a mix of mono nitrides of depleted uranium and power plutonium with a small share of neptunium and americium; the target coolant -- a modified lead and bismuth eutectic, Pb-208(80%)-Bi(20%); proton beam energy -- 600 MeV; effective multiplication factor of the core under operation -- K{sub eff} = 0.97; thermal capacity of the core -- N = 80 MW. From calculations performed it follows that in using Pb-208 as the core coolant the necessary intensity of the external source of neutrons to deliver 80 MW thermal capacity is equal to S = 2.29-10{sup 17} n/s that corresponds to proton beam current I{sub p} = 2.8 mA and beam capacity P{sub p} 1.68 MW. In using natural lead instead of Pb-208 as the core coolant, effective multiplication factor of the core in normal operating regime falls down to the value equal to K{sub eff} = 0.95. In these conditions multiplication of external neutrons in the core and thermal capacity of the subcritical core are below nominal by 1.55 times. For achievement the rated core power N=80 MW it is required on {approx}20-30% to increase the fuel loading and volume of the core, or by 1.55 times to increase intensity of the external source of neutrons. In the last case, the required parameters of the neutron source and of the corresponding proton beam are following: intensity of the neutron source S = 3.55 10{sup 17} n/s., beam current I{sub p} = 4.32 mA, beam capacity P{sub p} 2.59 MW. To exploit the accelerator with the reduced proton beam current it will be required about 56 tons of Pb-208, as a minimum, for the core coolant. Charges for its

  2. Flow boiling with enhancement devices for cold plate coolant channel design

    NASA Technical Reports Server (NTRS)

    Boyd, Ronald D.; Turknett, Jerry C.; Smith, Alvin

    1989-01-01

    The effects of enhancement devices on flow boiling heat transfer in circular coolant channels, which are heated over a fraction of their perimeters, are studied. The variations were examined in both the mean and local (axial, and circumferential) heat transfer coefficients for a circular coolant channel with either smooth walls or with both a twisted tape and spiral finned walls. Improvements were initiated in the present data reduction analysis. These efforts should lead to the development of heat transfer correlations which include effects of single side heat flux and enhancement device configuration. It is hoped that a stage will be set for the study of heat transfer and pressure drop in single sided heated systems under zero gravity conditions.

  3. K-Reactor emergency core coolant system response during a double-ended guillotine break LOCA

    SciTech Connect

    Rodriguez, S.B. )

    1990-01-01

    This paper describes the modeling and benchmarking of the Savannah River Site K-Reactor emergency core coolant system (ECCS), using the Transient Reactor Analysis Code (TRAC). The ECCS model was benchmarked against plant data obtained from various ECCS configurations. Next, the benchmarked model was used to simulate various loss-of-coolant accidents (LOCAs). The adequacy of the model's behavior during the LOCAs was then analyzed. The K-Reactor ECCS model can adequately simulate a wide variety of system configurations. The TRAC output compared favorably with the plant data for the different ECCS configurations. The results of the plenum-inlet double-ended guillotine break LOCA simulation showed the ECCS protected the core.

  4. Algebraic grid generation for coolant passages of turbine blades with serpentine channels and pin fins

    NASA Technical Reports Server (NTRS)

    Shih, T. I.-P.; Roelke, R. J.; Steinthorsson, E.

    1991-01-01

    In order to study numerically details of the flow and heat transfer within coolant passages of turbine blades, a method must first be developed to generate grid systems within the very complicated geometries involved. In this study, a grid generation package was developed that is capable of generating the required grid systems. The package developed is based on an algebraic grid generation technique that permits the user considerable control over how grid points are to be distributed in a very explicit way. These controls include orthogonality of grid lines next to boundary surfaces and ability to cluster about arbitrary points, lines, and surfaces. This paper describes that grid generation package and shows how it can be used to generate grid systems within complicated-shaped coolant passages via an example.

  5. PIV measurements of coolant flow field in a diesel engine cylinder head

    NASA Astrophysics Data System (ADS)

    Ma, Hongwei; Zhang, Zhenyang; Xue, Cheng; Huang, Yunlong

    2015-04-01

    This paper presents experimental measurements of coolant flow field in the water jacket of a diesel engine cylinder head. The test was conducted at three different flow rates using a 2-D PIV system. Appropriate tracing particles were selected and delivery device was designed and manufactured before the test. The flow parameters, such as velocity, vorticity and turbulence, were used to analyze the flow field. The effects of vortex which was located between the intake valve and the exhaust valve were discussed. The experimental results showed an asymmetric distribution of velocity in the water jacket. This led to an asymmetric thermal distribution, which would shorten the service life of the cylinder head. The structure optimization to the water jacket of cylinder head was proposed in this paper. The experimental system, especially the 2-D PIV system, is a great help to study the coolant flow structure and analyze cooling mechanism in the diesel engine cylinder head.

  6. Analysis of coolant flow and heat transfer in the SSME HPOTP Number 4 bearing assembly

    NASA Technical Reports Server (NTRS)

    Owens, S. F.; Costes, N. C.

    1990-01-01

    The PHOENICS code has been applied to simulate the flow of liquid oxygen through the number 4 ball bearing assembly of the Space Shuttle Main Engine High-Pressure Oxidizer Turbopump. The Body-Fitted Coordinate system capability of PHOENICS was utilized to create a geometrically accurate model. Effects accounted for by the model include the rotation rate of the calculation domain, viscous heating in the liquid oxygen and two-phase effects due to LOX boiling. A separation program was used to account for conduction within a ball bearing element which was coupled with the PHOENICS flow and heat transfer analysis. Solutions have been obtained for the velocity and temperature fields within the LOX coolant and the temperature within the ball element. The predicted ball temperatures indicate that the coolant is boiling in the region near the ball surface.

  7. Large break loss-of-coolant accident analyses for the high flux isotope reactor

    SciTech Connect

    Taleyarkhan, R.P. )

    1989-01-01

    The US Department of Energy's High Flux Isotope Reactor (HFIR) was analyzed to evaluate it's response to a spectrum of loss-of-coolant accidents (LOCAs) with potential for leading to core damage. The MELCOR severe accident analysis code (version 1.7.1) was used to evaluate the overall dynamic response of HFIR. Before conducting LOCA analyses, the steady-state thermal-hydraulic parameters evaluated by MELCOR for various loop sections were verified against steady-state operating data. Thereafter, HFIR depressurization tests were simulated to evaluate the system pressure change for a given depletion in coolant inventory. Interesting and important safety-related phenomena were observed. The current analyses (which should be considered preliminary) that occur over a period from 1 to 3 seconds do not lead to core wide fuel melting. Core fluid flashing during the initial rapid depressurization does cause fuel temperature excursions due to adiabatic-like heatup. 3 refs., 4 figs.

  8. Evaluation of secondary coolant control design alternatives and their effects on heat removal performance

    SciTech Connect

    Khayat, M.I.; Anderson, J.; Battle, R.; March-Leuba, J.

    1994-03-01

    This report documents a series of calculations that evaluate the performance of the core-inlet temperature controller under different transient conditions and design options. The present analyses show that the core-inlet temperature can be controlled at {approx}45{degrees}C under all transient conditions analyzed using the controller design described in the conceptual design report, which includes variable-speed secondary coolant pumps and variable-speed cooling tower fans. This study also shows that a constant-speed secondary pump would be sufficient to maintain core-inlet temperature <45{degrees}C if this temperature is allowed to drop below the set point during some demanding transients, such as normal startup. The use of secondary loop hot coolant to warm the reactor building was also evaluated; however, optimization of the secondary hot-leg temperature can only be achieved by trading off control of the primary side core-inlet temperature.

  9. Flow boiling with enhancement devices for cold plate coolant channel design

    NASA Technical Reports Server (NTRS)

    Boyd, Ronald D.

    1991-01-01

    Future space exploration and commercialization will require more efficient heat rejection systems. For the required heat transfer rates, such systems must use advanced heat transfer techniques. Forced two phase flow boiling heat transfer with enhancements falls in this category. However, moderate to high quality two phase systems tend to require higher pressure losses. This report is divided into two major parts: (1) Multidimensional wall temperature measurement and heat transfer enhancement for top heated horizontal channels with flow boiling; and (2) Improved analytical heat transfer data reduction for a single side heated coolant channel. Part 1 summarizes over forty experiments which involve both single phase convection and flow boiling in a horizontal channel heated externally from the top side. Part 2 contains parametric dimensionless curves with parameters such as the coolant channel radius ratio, the Biot number, and the circumferential coordinate.

  10. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    SciTech Connect

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  11. Fluid-Structure Interaction for Coolant Flow in Research-type Nuclear Reactors

    SciTech Connect

    Curtis, Franklin G; Ekici, Kivanc; Freels, James D

    2011-01-01

    The High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), is scheduled to undergo a conversion of the fuel used and this proposed change requires an extensive analysis of the flow through the reactor core. The core consists of 540 very thin and long fuel plates through which the coolant (water) flows at a very high rate. Therefore, the design and the flow conditions make the plates prone to dynamic and static deflections, which may result in flow blockage and structural failure which in turn may cause core damage. To investigate the coolant flow between fuel plates and associated structural deflections, the Fluid-Structure Interaction (FSI) module in COMSOL will be used. Flow induced flutter and static deflections will be examined. To verify the FSI module, a test case of a cylinder in crossflow, with vortex induced vibrations was performed and validated.

  12. Direct Numerical Simulation of a Coolant Jet in a Periodic Crossflow

    NASA Technical Reports Server (NTRS)

    Sharma, Chirdeep; Acharya, Sumanta

    1998-01-01

    A Direct Numerical Simulation of a coolant jet injected normally into a periodic crossflow is presented. The physical situation simulated represents a periodic module in a coolant hole array with a heated crossflow. A collocated finite difference scheme is used which is fifth-order accurate spatially and second-order accurate temporally. The scheme is based on a fractional step approach and requires the solution of a pressure-Poisson equation. The simulations are obtained for a blowing ratio of 0.25 and a channel Reynolds number of 5600. The simulations reveal the dynamics of several large scale structures including the Counter-rotating Vortex Pair (CVP), the horse-shoe vortex, the shear layer vortex, the wall vortex and the wake vortex. The origins and the interactions of these vortical structures are identified and explored. Also presented are the turbulence statistics and how they relate to the flow structures.

  13. Literature review of tilting pad and turbulent hydrostatic journal bearings for nuclear main coolant pumps

    SciTech Connect

    Flack, R.D.; Allaire, P.E.

    1984-07-01

    Literature on the dynamic and static operating characteristics of tilting-pad and turbulent hydrostatic journal bearings is reviewed. Experimental and theoretical work as well as applications in which these bearings were used in a dynamical analysis of the rotor/bearing system are included. Journal bearings receive considerable use in the nuclear power industry such bearings in a reactor main coolant pump are considered herein.

  14. Measured effects of coolant injection on the performance of a film cooled turbine

    NASA Technical Reports Server (NTRS)

    Mcdonel, J. D.; Eiswerth, J. E.

    1977-01-01

    Tests have been conducted on a 20-inch diameter single-stage air-cooled turbine designed to evaluate the effects of film cooling air on turbine aerodynamic performance. The present paper reports the results of five test configurations, including two different cooling designs and three combinations of cooled and solid airfoils. A comparison is made of the experimental results with a previously published analytical method of evaluating coolant injection effects on turbine performance.

  15. Lead Coolant Test Facility Technical and Functional Requirements, Conceptual Design, Cost and Construction Schedule

    SciTech Connect

    Soli T. Khericha

    2006-09-01

    This report presents preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements of basis are identified: Develop and Demonstrate Prototype Lead/Lead-Bismuth Liquid Metal Flow Loop Develop and Demonstrate Feasibility of Submerged Heat Exchanger Develop and Demonstrate Open-lattice Flow in Electrically Heated Core Develop and Demonstrate Chemistry Control Demonstrate Safe Operation and Provision for Future Testing. These five broad areas are divided into twenty-one (21) specific requirements ranging from coolant temperature to design lifetime. An overview of project engineering requirements, design requirements, QA and environmental requirements are also presented. The purpose of this T&FRs is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420oC. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M. It is also estimated that the facility will require two years to be constructed and ready for operation.

  16. Additional requirements for leak-before-break application to primary coolant piping in Belgium

    SciTech Connect

    Roussel, G.

    1997-04-01

    Leak-Before-Break (LBB) technology has not been applied in the first design of the seven Pressurized Water Reactors the Belgian utility is currently operating. The design basis of these plants required to consider the dynamic effects associated with the ruptures to be postulated in the high energy piping. The application of the LBB technology to the existing plants has been recently approved by the Belgian Safety Authorities but with a limitation to the primary coolant loop. LBB analysis has been initiated for the Doel 3 and Tihange 2 plants to allow the withdrawal of some of the reactor coolant pump snubbers at both plants and not reinstall some of the restraints after steam generator replacement at Doel 3. LBB analysis was also found beneficial to demonstrate the acceptability of the primary components and piping to the new conditions resulting from power uprating and stretch-out operation. LBB analysis has been subsequently performed on the primary coolant loop of the Tihange I plant and is currently being performed for the Doel 4 plant. Application of the LBB to the primary coolant loop is based in Belgium on the U.S. Nuclear Regulatory Commission requirements. However the Belgian Safety Authorities required some additional analyses and put some restrictions on the benefits of the LBB analysis to maintain the global safety of the plant at a sufficient level. This paper develops the main steps of the safety evaluation performed by the Belgian Safety Authorities for accepting the application of the LBB technology to existing plants and summarizes the requirements asked for in addition to the U.S. Nuclear Regulatory Commission rules.

  17. A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

    SciTech Connect

    Siamidis, John; Mason, Lee

    2006-01-20

    This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H2O for the HRS pumped loop coolant working fluid. A detailed excel analytical model, HRS{sub O}pt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids.

  18. Influence of coolant pH on corrosion of 6061 aluminum under reactor heat transfer conditions

    SciTech Connect

    Pawel, S.J.; Felde, D.K.; Pawel, R.E.

    1995-10-01

    To support the design of the Advanced Neutron Source (ANS), an experimental program was conducted wherein aluminum alloy specimens were exposed at high heat fluxes to high-velocity aqueous coolants in a corrosion test loop. The aluminum alloys selected for exposure were candidate fuel cladding materials, and the loop system was constructed to emulate the primary coolant system for the proposed ANS reactor. One major result of this program has been the generation of an experimental database defining oxide film growth on 6061 aluminum alloy cladding. Additionally, a data correlation was developed from the database to permit the prediction of film growth for any reasonable thermal-hydraulic excursion. This capability was utilized effectively during the conceptual design stages of the reactor. During the course of this research, it became clear that the kinetics of film growth on the aluminum alloy specimens were sensitively dependent on the chemistry of the aqueous coolant and that relatively small deviations from the intended pH 5 operational level resulted in unexpectedly large changes in the corrosion behavior. Examination of the kinetic influences and the details of the film morphology suggested that a mechanism involving mass transport from other parts of the test loop was involved. Such a mechanism would also be expected to be active in the operating reactor. This report emphasizes the results of experiments that best illustrate the influence of the nonthermal-hydraulic parameters on film growth and presents data to show that comparatively small variations in pH near 5.0 invoke a sensitive response. Simply, for operation in the temperature and heat flux range appropriate for the ANS studies, coolant pH levels from 4.5 to 4.9 produced significantly less film growth than those from pH 5.1 to 6. A mechanism for this behavior based on the concept of treating the entire loop as an active corrosion system is presented.

  19. A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

    NASA Technical Reports Server (NTRS)

    Siamidis, John; Mason, Lee S.

    2006-01-01

    This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H2O for the HRS pumped loop coolant working fluid. A detailed Microsoft Excel (Microsoft Corporation, Redmond, WA) analytical model, HRS_Opt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids.

  20. A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

    NASA Technical Reports Server (NTRS)

    Mason, Lee S.; Siamidis, John

    2006-01-01

    This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H20 for the HRS pumped loop coolant working fluid. A detailed Microsoft Excel (Microsoft Corporation, Redmond, WA) analytical model, HRS_Opt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids.

  1. Main-coolant-pump shaft-seal guidelines. Volume 3. Specification guidelines. Final report. [PWR; BWR

    SciTech Connect

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and criteria to aid in the generation of procurement specifications for Main Coolant Pump Shaft Seals. The noted guidelines are developed from EPRI sponsored nuclear power plant seal operating experience studies, a review of pump and shaft seal literature and discussions with pump and seal designers. This report is preliminary in nature and could be expanded and finalized subsequent to completion of further design, test and evaluation efforts.

  2. Main-coolant-pump shaft-seal guidelines. Volume 2. Operational guidelines. Final report. [PWR; BWR

    SciTech Connect

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and criteria for improving main coolant pump shaft seal operational reliability. The noted guidelines are developed from EPRI sponsored nuclear power plant seal operating experience studies. Usage procedures/practices and operational environment influence on seal life and reliability from the most recent such survey are summarized. The shaft seal and its auxiliary supporting systems are discussed both from technical and operational related viewpoints.

  3. Method of and apparatus for removing silicon from a high temperature sodium coolant

    DOEpatents

    Yunker, W.H.; Christiansen, D.W.

    1983-11-25

    This patent discloses a method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

  4. Method for controlling coolant flow in airfoil, flow control structure and airfoil incorporating the same

    DOEpatents

    Itzel, Gary Michael; Devine, II, Robert Henry; Chopra, Sanjay; Toornman, Thomas Nelson

    2003-07-08

    A coolant flow control structure is provided to channel cooling media flow to the fillet region defined at the transition between the wall of a nozzle vane and a wall of a nozzle segment, for cooling the fillet region. In an exemplary embodiment, the flow control structure defines a gap with the fillet region to achieve the required heat transfer coefficients in this region to meet part life requirements.

  5. Probability of failure in BWR reactor coolant piping: Guillotine break indirectly induced by earthquakes

    SciTech Connect

    Hardy, G.S.; Campbell, R.D.; Ravindra, M.K.

    1986-12-01

    The requirements to design nuclear power plants for the effects of an instantaneous double-ended guillotine break (DEGB) of the reactor coolant piping have led to excessive design costs, interference with normal plant operation and maintenance, and unnecessary radiation exposure of plant maintenance personnel. This report describes an aspect of the NRC/Lawrence Livermore National laboratory-sponsored research program aimed at investigating whether the probability of DEGB in Reactor Coolant Loop Piping of nuclear power plants is acceptably small such that the requirements to design for the DEGB effects (e.g., provision of pipe whip restraints) may be removed. This study estimates the probability of indirect DEGB in Reactor Coolant piping as a consequence of seismic-induced structural failures within the containment of the GE supplied boiling water reactor at the Brunswick nuclear power plant. The median probability of indirect DEGB was estimated to be 2 x 10/sup -8/ per year. Using conservation assumptions, the 90% subjective probability value (confidence) of P/sub DEGB/ was found to be less than 5 x 10/sup -7/ per year.

  6. Probabilistic assessment of decoupling loss-of-coolant accident and earthquake in nuclear power plant design

    SciTech Connect

    Lu, S.C.; Harris, D.O.

    1981-01-01

    This paper describes a research project conducted at Lawrence Livermore National Laboratory to establish a technical basis for reassessing the requirement of combining large loss-of-coolant-accident (LOCA) and earthquake loads in nuclear power plant design. A large LOCA is defined herein as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressureized water reactor (PWR). A systematic probability approach has been employed to estimate the probability of a large LOCA directly and indirectly induced by earthquakes. The probability of a LOCA directly induced by earthquakes was assessed by a numerical simulation of pipe rupture of a reactor coolant system. The simulation employed a deterministic fracture mechanics model which dictates the fatigue growth of pre-existing cracks in the pipe. The simulation accounts for the stochastic nature of input elements such as the initial crack size distribution, the crack occurrence rate, crack and leak detection probabilities as functions of crack size, plant transient occurrence rates, the seismic hazard, stress histories, and crack growth model parameters. Effects on final results due to variation an uncertainty of input elements were assessed by a limited sensitivity study. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the orer of 10/sup -12/). The probability of a leak was found to be several orders of magnitudes greater than that of a complete break.

  7. Assessment of fiber optic sensors for aging monitoring of industrial liquid coolants

    NASA Astrophysics Data System (ADS)

    Riziotis, Christos; El Sachat, Alexandros; Markos, Christos; Velanas, Pantelis; Meristoudi, Anastasia; Papadopoulos, Aggelos

    2015-03-01

    Lately the demand for in situ and real time monitoring of industrial assets and processes has been dramatically increased. Although numerous sensing techniques have been proposed, only a small fraction can operate efficiently under harsh industrial environments. In this work the operational properties of a proposed photonic based chemical sensing scheme, capable to monitor the ageing process and the quality characteristics of coolants and lubricants in industrial heavy machinery for metal finishing processes is presented. The full spectroscopic characterization of different coolant liquids revealed that the ageing process is connected closely to the acidity/ pH value of coolants, despite the fact that the ageing process is quite complicated, affected by a number of environmental parameters such as the temperature, humidity and development of hazardous biological content as for example fungi. Efficient and low cost optical fiber sensors based on pH sensitive thin overlayers, are proposed and employed for the ageing monitoring. Active sol-gel based materials produced with various pH indicators like cresol red, bromophenol blue and chorophenol red in tetraethylorthosilicate (TEOS), were used for the production of those thin film sensitive layers deposited on polymer's and silica's large core and highly multimoded optical fibers. The optical characteristics, sensing performance and environmental robustness of those optical sensors are presented, extracting useful conclusions towards their use in industrial applications.

  8. Flow boiling with enhancement devices for cold plate coolant channel design

    NASA Technical Reports Server (NTRS)

    Boyd, Ronald D., Sr.; Smith, Alvin

    1990-01-01

    The use of flow boiling for thermal energy transport is intended to provide an alternative for accommodating higher heat fluxes in commercial space systems. The objectives are to: (1) examine the variations in both the mean and local (axial and circumferential) heat transfer coefficients for a circular coolant channel with either smooth walls, spiral fins, or both spiral fins and a twisted tape; (2) examine the effects of channel diameter and subcooling; and (3) develop an improved reduction analysis and/or suggest possible heat transfer correlation of the present data. Freon-11 is the working fluid. Two-dimensional (circumferential and axial) wall temperature distributions were measured for coolant channels with the above noted internal geometries. The flow regimes which are being studied are: (1) single phase; (2) subcooled flow boiling; and (3) stratified flow boiling. The inside diameter of all test sections is near 1.0 cm. Cicumferentially averaged heat transfer coefficients at several axial locations were obtained for selected coolant channels for a mass velocity of 210 kg/sq m s, an exit pressure of 0.19 MPa (absolute), and an inlet subcooling of 20.8 C. Overall (averaged over the entire channel) heat transfer coefficients were compared for the above channel geometries. This comparison showed that the channel with large pitch spiral fins had higher heat transfer coefficients at all power levels.

  9. Chimera grids in the simulation of three-dimensional flowfields in turbine-blade-coolant passages

    NASA Technical Reports Server (NTRS)

    Stephens, M. A.; Rimlinger, M. J.; Shih, T. I.-P.; Civinskas, K. C.

    1993-01-01

    When computing flows inside geometrically complex turbine-blade coolant passages, the structure of the grid system used can affect significantly the overall time and cost required to obtain solutions. This paper addresses this issue while evaluating and developing computational tools for the design and analysis of coolant-passages, and is divided into two parts. In the first part, the various types of structured and unstructured grids are compared in relation to their ability to provide solutions in a timely and cost-effective manner. This comparison shows that the overlapping structured grids, known as Chimera grids, can rival and in some instances exceed the cost-effectiveness of unstructured grids in terms of both the man hours needed to generate grids and the amount of computer memory and CPU time needed to obtain solutions. In the second part, a computational tool utilizing Chimera grids was used to compute the flow and heat transfer in two different turbine-blade coolant passages that contain baffles and numerous pin fins. These computations showed the versatility and flexibility offered by Chimera grids.

  10. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    SciTech Connect

    Chopra, O.K.; Shack, W.J.

    1998-03-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented.

  11. Parametric study on maximum transportable distance and cost for thermal energy transportation using various coolants

    SciTech Connect

    Su-Jong Yoon; Piyush Sabharwall

    2014-07-01

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as district heating, desalination, hydrogen production and other process heat applications, etc. The process heat industry/facilities will be located outside the nuclear island due to safety measures. This thermal energy from the reactor has to be transported a fair distance. In this study, analytical analysis was conducted to identify the maximum distance that thermal energy could be transported using various coolants such as molten-salts, helium and water by varying the pipe diameter and mass flow rate. The cost required to transport each coolant was also analyzed. The coolants analyzed are molten salts (such as: KClMgCl2, LiF-NaF-KF (FLiNaK) and KF-ZrF4), helium and water. Fluoride salts are superior because of better heat transport characteristics but chloride salts are most economical for higher temperature transportation purposes. For lower temperature water is a possible alternative when compared with He, because low pressure He requires higher pumping power which makes the process very inefficient and economically not viable for both low and high temperature application.

  12. Preparation, conduct, and experimental results of the AVR loss-of-coolant accident simulation test

    SciTech Connect

    Kruger, K.; Bergerfurth, A.; Burger, S.; Pohl, P.; Wimmers, M. ); Cleveland, J.C. )

    1991-02-01

    A loss-of-coolant accident (LOCA) is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into small high-temperature gas-cooled reactor (HTGR) design, LOCA simulation tests have been conducted at the Arbeitsgemeinschaft Versuchsreaktor (AVR), the German pebble-bed-high-temperature reactor plant. The AVR is the only nuclear power plant ever to have been intentionally subjected to LOCA conditions without emergency cooling. This paper presents the planning and licensing activities including pretest predictions performed for the LOCA test are described, and the conduct of the test and experimental results. The LOCA test was planned to create conditions that would exist if a rapid LOCA occurred with the reactor operating at full power. The test demonstrated this reactor's safe response to an accident in which the coolant escapes from the reactor core and no emergency system is available to provide coolant flow to the core. The test is of special interest because it demonstrates the inherent safety features incorporated into optimized modular HTGR designs. The main LOCA test lasted for 5 days. After the test began, core temperatures increased for {approx}13 h and then gradually and continually decreased as the rate of heat dissipation from the core exceeded the simulated decay power. Throughout the test, temperatures remained below limiting values for the core and other reactor components.

  13. Assessment of Thermal and Hydrodynamic Fragmentation in Molten Fuel Coolant Interaction With Simulant System

    SciTech Connect

    Narayanan, K.S.; Das, S.K.; Jasmin Sudha, A.; Rao, E.H.V.M.; Lydia, G.; Murthy, S.S.; Kumareshan, M.; Harvey, J.; Kasinathan, N.; Rajan, M.

    2006-07-01

    In the Safety analysis of Fast Breeder Reactor, assessment of Molten Fuel Coolant Interaction (MFCI) assumes importance for two aspects, namely the characterization of the debris and severity of pressure pulses generation. An attempt has been made to investigate the debris generation characteristics with molten Woods Metal (Alloy of Bi 50% Pb 25% Sn 12.5% and Cd 12.5% and melting point of 346 K) - Water simulant system. Liquid Woods metal and liquid Uranium dioxide physical properties (Density, Surface tension and Kinematic viscosity) are similar. Experimental studies were conducted for various melt temperatures covering non - boiling, convective boiling and film boiling regimes of water, to assess the debris generation resulting from both hydrodynamic and thermal interaction. Woods metal was heated to the desired temperature and poured through a hot funnel having a nozzle of 8 mm release diameter into a water column of height up to 140 cm. Experiments were repeated for different coolant temperature and melt inventory up to 5 kg. The melt entry velocity was determined from video recordings. The debris is analyzed on the basis of interface temperature, Rayleigh-Taylor and Kelvin - Helmholtz instabilities. It is observed that Kelvin-Helmholtz instability is the dominant fragmentation phenomena. Contribution due to coolant boiling resulted in more debris generation in the size less than 4 mm. (authors)

  14. Guidelines to achieve seals with minimal leak rates for HWR-NPR coolant system components

    SciTech Connect

    Finn, P.A.

    1991-03-01

    Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs.

  15. A {open_quotes}zero waste{close_quotes} coolant management strategy

    SciTech Connect

    Kennicott, M.A.

    1994-04-01

    In June of 1992 the Waste Minimization Program at Rocky Flats Plant (RFP) began a study to determine the best methods of managing water-based industrial metalworking fluids in the plant`s Tool Manufacturing Shop. The shop was faced with the challenge of managing fluids that could no longer be disposed of in the traditional manner, through the plant`s liquid process waste drains, due to a problem they, were having causing in the Liquid Waste Operations Evaporator. The study`s goal was to reduce the waste coolants being generated and to reduce worker exposure to a serious health risk. Results of this study and those of a subsequent study to determine relative compatibilities of various coolants and metals, led to the application of a {open_quotes}zero waste{close_quotes} machine coolant management program. This program is currently saving the generation of 10,000 gallons of liquid waste annually, has eliminated worker exposure to harmful bacteria and biocides, and should result in extended machine tool life, increased product quality, fewer rejected parts, and decreases labor costs.

  16. Nuclear-radiation-actuated valve. [Patent application; for increasing coolant flow to blanket

    DOEpatents

    Christiansen, D.W.; Schively, D.P.

    1982-01-19

    The present invention relates to a breeder reactor blanket fuel assembly coolant system valve which increases coolant flow to the blanket fuel assembly to minimize long-term temperature increases caused by fission of fissile fuel created from fertile fuel through operation of the breeder reactor. The valve has a valve first part (such as a valve rod with piston) and a valve second part (such as a valve tube surrounding the valve rod, with the valve tube having side slots surrounding the piston). Both valve parts have known nuclear radiation swelling characteristics. The valve's first part is positioned to receive nuclear radiation from the nuclear reactor's fuel region. The valve's second part is positioned so that its nuclear radiation induced swelling is different from that of the valve's first part. The valve's second part also is positioned so that the valve's first and second parts create a valve orifice which changes in size due to the different nuclear radiation caused swelling of the valve's first part compared to the valve's second part. The valve may be used in a nuclear reactor's core coolant system.

  17. The high-temperature sodium coolant technology in nuclear power installations for hydrogen power engineering

    NASA Astrophysics Data System (ADS)

    Kozlov, F. A.; Sorokin, A. P.; Alekseev, V. V.; Konovalov, M. A.

    2014-05-01

    In the case of using high-temperature sodium-cooled nuclear power installations for obtaining hydrogen and for other innovative applications (gasification and fluidization of coal, deep petroleum refining, conversion of biomass into liquid fuel, in the chemical industry, metallurgy, food industry, etc.), the sources of hydrogen that enters from the reactor plant tertiary coolant circuit into its secondary coolant circuit have intensity two or three orders of magnitude higher than that of hydrogen sources at a nuclear power plant (NPP) equipped with a BN-600 reactor. Fundamentally new process solutions are proposed for such conditions. The main prerequisite for implementing them is that the hydrogen concentration in sodium coolant is a factor of 100-1000 higher than it is in modern NPPs taken in combination with removal of hydrogen from sodium by subjecting it to vacuum through membranes made of vanadium or niobium. Numerical investigations carried out using a diffusion model showed that, by varying such parameters as fuel rod cladding material, its thickness, and time of operation in developing the fuel rods for high-temperature nuclear power installations (HT NPIs) it is possible to exclude ingress of cesium into sodium through the sealed fuel rod cladding. However, if the fuel rod cladding loses its tightness, operation of the HT NPI with cesium in the sodium will be unavoidable. Under such conditions, measures must be taken for deeply purifying sodium from cesium in order to minimize the diffusion of cesium into the structural materials.

  18. The impact of radiolytic yield on the calculated ECP in PWR primary coolant circuits

    NASA Astrophysics Data System (ADS)

    Urquidi-Macdonald, Mirna; Pitt, Jonathan; Macdonald, Digby D.

    2007-05-01

    A code, PWR-ECP, comprising chemistry, radiolysis, and mixed potential models has been developed to calculate radiolytic species concentrations and the corrosion potential of structural components at closely spaced points around the primary coolant circuits of pressurized water reactors (PWRs). The pH( T) of the coolant is calculated at each point of the primary-loop using a chemistry model for the B(OH) 3 + LiOH system. Although the chemistry/radiolysis/mixed potential code has the ability to calculate the transient reactor response, only the reactor steady state condition (normal operation) is discussed in this paper. The radiolysis model is a modified version of the code previously developed by Macdonald and coworkers to model the radiochemistry and corrosion properties of boiling water reactor primary coolant circuits. In the present work, the PWR-ECP code is used to explore the sensitivity of the calculated electrochemical corrosion potential (ECP) to the set of radiolytic yield data adopted; in this case, one set had been developed from ambient temperature experiments and another set reported elevated temperatures data. The calculations show that the calculated ECP is sensitive to the adopted values for the radiolytic yields.

  19. Radiolysis of the coolant in the VK-50 boiling water reactor

    SciTech Connect

    Zabelin, A.I.; Shmelev, V.E.

    1986-10-01

    Radiolysis of the coolant proceeds at a higher rate in a boiling water reactor as compared to a water-moderated, water-cooled reactor. The radiolytic gases (hydrogen and oxygen) exiting the reactor together with steam can form a potentially explosive mixture. Special interest attaches to the results obtained under the codnitions of prolonged operation of the VK-50 reactor. Tests of various water-chemistry conditions which were performed in the experimental reactor showed their critical influence on the rate of progress of radiolytic processes. The entire period of operation of the reactor may be arbitrarily divided into three stages, each of which is characterized by its own peculiar conditions of water chemistry and range of thermal power. From stage to stage, there is a noticeable improvement in the coolant quality which to a limited extent is reflected in the exit of radiolytic gases with the steam. The concentration of radiolytic gases increases with decreased power and with an increased content of corrosion products and other contaminants in the coolant.

  20. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    NASA Astrophysics Data System (ADS)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  1. Evaluation of the coolant reactivity coefficient influence on the dynamic response of a small LFR system

    SciTech Connect

    Lorenzi, S.; Bortot, S.; Cammi, A.; Ponciroli, R.

    2012-07-01

    An assessment of the coolant reactivity feedback influence on a small Lead-cooled Fast Reactor (LFR) dynamics has been made aimed at providing both qualitative and quantitative insights into the system transient behavior depending on the sign of the above mentioned coefficient. The need of such an investigation has been recognized since fast reactors cooled by heavy liquid metals show to be characterized by a strong coupling between primary and secondary systems. In particular, the coolant density and radial expansion coefficients have been attested to play a major role in determining the core response to any perturbed condition on the Steam Generator (SG) side. The European Lead-cooled System (ELSY)-based demonstrator (DEMO) has been assumed as the reference LFR case study. As a first step, a zero-dimensional dynamics model has been developed and implemented in MATLAB/SIMULINK{sup R} environment; then typical transient scenarios have been simulated by incorporating the actual negative lead density reactivity coefficient and its opposite. In all the examined cases results have shown that the reactor behaves in a completely different way when considering a positive coolant feedback instead of the reference one, the system free dynamics resulting moreover considerably slower due to the core and SG mutually conflicting reactions. The outcomes of the present analysis may represent a useful feedback for both the core and the control system designers. (authors)

  2. Composition of incipient passivating layers on heat-rejecting aluminum in carboxylate- and silicate-inhibited coolants: Correlation with ASTM D 4340 weight losses

    SciTech Connect

    Wagner, F.T.; Moylan, T.E.; Simko, S.J.; Militello, M.C.

    1999-08-01

    X-ray photoelectron spectroscopy identified compositional differences between passivating layers initially formed in carboxylated coolants, in silicated coolants, and in a mixture thereof on well-controlled 319 aluminum surfaces under heat-rejecting conditions. The layer formed in silicated coolant was primarily silica, while that in carboxylated coolant was primarily hydrated alumina. Competition between inhibitor packages when carboxylated coolant was contaminated from the start with low levels of silicated coolant produced films which were not simply patchwise mixtures of the pure-coolant film types. The surface analytical results aid the interpretation of subtle differences in weight losses under the ASTM Standard Test Method for Corrosion of Cast Aluminum Alloys in Engine Coolants Under Heat-Rejecting Conditions (D 4340) in carboxylated versus silicated coolants that became more pronounced when testing was carried out at a vehicle-like 50% coolant concentration. Results from time-resolved D4340 measurements and from a two-step cleaning procedure further contribute towards proper evaluation of D4340 weight losses in the different coolant types.

  3. Fleet test evaluation of fully formulated heavy-duty coolant technology maintained with a delayed-release filter compared with coolant inhibited with a nitrited organic acid technology: An interim report

    SciTech Connect

    Aroyan, S.S.; Eaton, E.R.

    1999-08-01

    This paper is a controlled extended service interval (ESI) study of the comparative behaviors of a nitrite/borate/low-silicate, low total dissolved solids (TDS) coolant maintained with delayed-release filters, and an organic acid inhibited coolant technology in heavy-duty engines. It reports both laboratory and fleet test data from 66 trucks, powered with different makes of heavy-duty diesel engines. The engines were cooled with three different types of inhibitors and two different glycol base (ethylene glycol and propylene glycol) coolants for an initial period exceeding two years and 500,000 km (300,000 miles). The data reported include chemical depletion rates, periodic coolant chemical analyses, and engine/cooling system reliability experience. The ongoing test will continue for approximately five years and a 1.6 million km (1 million miles) duration. Thirteen trucks were retained as controls, operating with ASTM D 4985 specification (GM-6038 type) coolant maintained with a standard ASTM D 57542 supplemental coolant additive (SCA). Engines produced by Caterpillar, Detroit Diesel Corp., Cummins Engine Co., and Mack Trucks are included in the test mix.

  4. Numerical analysis of the hot-gas-side and coolant-side heat transfer in liquid rocket engine combustors

    NASA Technical Reports Server (NTRS)

    Wang, Ten-See; Van, Luong

    1992-01-01

    The objective of this paper are to develop a multidisciplinary computational methodology to predict the hot-gas-side and coolant-side heat transfer and to use it in parametric studies to recommend optimized design of the coolant channels for a regeneratively cooled liquid rocket engine combustor. An integrated numerical model which incorporates CFD for the hot-gas thermal environment, and thermal analysis for the liner and coolant channels, was developed. This integrated CFD/thermal model was validated by comparing predicted heat fluxes with those of hot-firing test and industrial design methods for a 40 k calorimeter thrust chamber and the Space Shuttle Main Engine Main Combustion Chamber. Parametric studies were performed for the Advanced Main Combustion Chamber to find a strategy for a proposed combustion chamber coolant channel design.

  5. The effect of fuel thermal conductivity on the behavior of LWR cores during loss-of-coolant accidents

    SciTech Connect

    Terrani, Kurt A.; Wang, Dean; Ott, Larry J.; Montgomery, Robert O.

    2014-05-01

    The effect of variation in thermal conductivity of light water reactor fuel elements on core response during loss-of-coolant accident scenarios is examined. Initially, a simplified numerical analysis is utilized to determine the time scales associated with dissipation of stored energy from the fuel into the coolant once the fission reaction is stopped. The analysis is then followed by full reactor system thermal-hydraulics analysis of a typical boiling and pressurized water reactor subjected to a large break loss-of-coolant accident scenario using the TRACE code. Accordingly, sensitivity analyses to examine the effect of an increase in fuel thermal conductivity, up to 500%, on fuel temperature evolution during these transients are performed. Given the major differences in thermal-hydraulics design aspects of boiling and pressurized water reactors, different fuel and temperature responses during the simulated loss-of-coolant transients are observed.

  6. Local heat transfer in internally cooled turbine airfoil leading edge regions. I - Impingement cooling without film coolant extraction. II - Impingement cooling with film coolant extraction

    NASA Astrophysics Data System (ADS)

    Bunker, R. S.; Metzger, D. E.

    The highly localized internal heat transfer characteristics of large-scale models of impingement-cooled turbine blade leading edge regions presently studied derives its cooling from a single line of equally-spaced multiple jets aimed at the leading-edge apex, and exiting the leading-edge region in the opposite or chordwise direction. Detailed two-dimensional local surface Nusselt number distributions have been obtained with temperature-indicating coatings. Results indicate generally increasing heat transfer with the 0.6 power of jet Reynolds number. In the second part of this study, in which the same cooling process is used in conjunction with the extraction of the coolant fluid, the results obtained indicate that heat transfer is primarily dependent on jet Reynolds number, with smaller influences from the flow-extraction rate.

  7. A model for predicting coolant activity behaviour for fuel-failure monitoring analysis

    NASA Astrophysics Data System (ADS)

    El-Jaby, Ali

    A CANDU fuel element becomes defective when the Zircaloy-4 sheath is breached, allowing high pressure heavy water (D2O) coolant to enter the fuel-to-sheath gap, thereby creating a direct path for fission products (mainly volatile species of iodine and noble gases) and fuel debris to escape into the primary heat transport system (PHTS). In addition, the entry of D 2O coolant into the fuel-to-sheath gap may cause the UO2 fuel to oxidize, which in turn can augment the rate of fission product release into the PHTS. The release of fission products and fuel debris into the PHTS will elevate circuit contamination levels, consequently increasing radiation exposure to station personnel during maintenance tasks. Moreover, the continued operation of a defective fuel element may diminish its thermal performance due to fuel oxidation effects. It is therefore desirable to discharge defective fuel as soon as possible. Hence, a better understanding of defective fuel behaviour is required in order to develop an improved methodology for fuel-failure monitoring and PHTS coolant activity prediction. A mathematical model has been developed to predict the release of volatile fission products from operating defective nuclear fuel elements. The fission product activity in both the fuel-to-sheath gap and PHTS coolant as a function of time can be predicted during all reactor operations including steady-state operation as well as reactor shutdown, startup, and bundle-shifting manoeuvres. In addition, an improved ability to predict the PHTS coolant activity of the 135Xe isotope in commercial reactors is discussed. Moreover, a method to approximate both the burnup and the amount of the tramp uranium deposits in-core, as well as the tramp uranium fission rate is proposed. The model has been implemented as the STAR (Steady-state and Transient Activity Release) stand-alone code written in the C++ programming language using a custom developed finite-difference variable-mesh (FDVM) numerical

  8. Numerical Simulation of Non-Rotating and Rotating Coolant Channel Flow Fields. Part 1

    NASA Technical Reports Server (NTRS)

    Rigby, David L.

    2000-01-01

    Future generations of ultra high bypass-ratio jet engines will require far higher pressure ratios and operating temperatures than those of current engines. For the foreseeable future, engine materials will not be able to withstand the high temperatures without some form of cooling. In particular the turbine blades, which are under high thermal as well as mechanical loads, must be cooled. Cooling of turbine blades is achieved by bleeding air from the compressor stage of the engine through complicated internal passages in the turbine blades (internal cooling, including jet-impingement cooling) and by bleeding small amounts of air into the boundary layer of the external flow through small discrete holes on the surface of the blade (film cooling and transpiration cooling). The cooling must be done using a minimum amount of air or any increases in efficiency gained through higher operating temperature will be lost due to added load on the compressor stage. Turbine cooling schemes have traditionally been based on extensive empirical data bases, quasi-one-dimensional computational fluid dynamics (CFD) analysis, and trial and error. With improved capabilities of CFD, these traditional methods can be augmented by full three-dimensional simulations of the coolant flow to predict in detail the heat transfer and metal temperatures. Several aspects of turbine coolant flows make such application of CFD difficult, thus a highly effective CFD methodology must be used. First, high resolution of the flow field is required to attain the needed accuracy for heat transfer predictions, making highly efficient flow solvers essential for such computations. Second, the geometries of the flow passages are complicated but must be modeled accurately in order to capture all important details of the flow. This makes grid generation and grid quality important issues. Finally, since coolant flows are turbulent and separated the effects of turbulence must be modeled with a low Reynolds number

  9. Control of reactor coolant flow path during reactor decay heat removal

    DOEpatents

    Hunsbedt, Anstein N.

    1988-01-01

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

  10. A passively-safe fusion reactor blanket with helium coolant and steel structure

    SciTech Connect

    Crosswait, K.M.

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

  11. Estimating Loss-of-Coolant Accident Frequencies for the Standardized Plant Analysis Risk Models

    SciTech Connect

    S. A. Eide; D. M. Rasmuson; C. L. Atwood

    2008-09-01

    The U.S. Nuclear Regulatory Commission maintains a set of risk models covering the U.S. commercial nuclear power plants. These standardized plant analysis risk (SPAR) models include several loss-of-coolant accident (LOCA) initiating events such as small (SLOCA), medium (MLOCA), and large (LLOCA). All of these events involve a loss of coolant inventory from the reactor coolant system. In order to maintain a level of consistency across these models, initiating event frequencies generally are based on plant-type average performance, where the plant types are boiling water reactors and pressurized water reactors. For certain risk analyses, these plant-type initiating event frequencies may be replaced by plant-specific estimates. Frequencies for SPAR LOCA initiating events previously were based on results presented in NUREG/CR-5750, but the newest models use results documented in NUREG/CR-6928. The estimates in NUREG/CR-6928 are based on historical data from the initiating events database for pressurized water reactor SLOCA or an interpretation of results presented in the draft version of NUREG-1829. The information in NUREG-1829 can be used several ways, resulting in different estimates for the various LOCA frequencies. Various ways NUREG-1829 information can be used to estimate LOCA frequencies were investigated and this paper presents two methods for the SPAR model standard inputs, which differ from the method used in NUREG/CR-6928. In addition, results obtained from NUREG-1829 are compared with actual operating experience as contained in the initiating events database.

  12. Severe degradation of BWR fuel failures: Coolant activity analysis. Interim report

    SciTech Connect

    Klepfer, H.H.; Kjaer-Pedersen, N.; Rank, P.; Ozer, O.

    1993-11-01

    The Electric Power Research Institute (EPRI) has been investigating an apparent recent increase in the frequency of severely degraded failed fuel rods in BWRs. These incidents involve abnormal levels of coolant radioactivity from the failed rod, and are followed by a steep increase in the activity from tramp uranium, most of which is deposited on fuel rod surfaces and therefore tends to carry over to subsequent operating cycles. Analyses of coolant activity from several BWRs experiencing degradation of failed fuel rods were carried out with extensive use of the CHIRON code, developed for EPRI by S. Levy Incorporated. CHIRON is based on steady-state conditions, but its trending plot capability was extremely useful for studying the time variation of radiochemical parameters. It is observed that the failure degradation process usually starts three to eight months after a primary failure occurrence. It is possible from measurements of Np{sup 239} coolant activity to reach quite accurate estimates of the amount of fuel released from the failed rod(s) over time. The neptunium calculations can also be used to estimate the size/length of the cladding rupture. An equation has been developed to predict offgas at the start of the next cycle from offgas at end of cycle. It is concluded that a failed BWR fuel rod can deteriorate severely within a few months, and therefore all failure occurrences should be acted upon with the serious consequences of severe damage in mind. A set of recommended guidelines is established for monitoring and interpreting activity signals, and for acting upon failure observations.

  13. High pressure coolant effect on PVD coated inserts during end milling of Ti-6AL-4V

    NASA Astrophysics Data System (ADS)

    Sridharan, Arvind

    Titanium alloys are being employed extensively in engineering and aerospace applications for their high strength to weight ratio, mechanical strength and ability to withstand high temperatures. Out of the different alloys of titanium available, the most commonly used alloy is Ti-6Al-4V. It is also called `Grade-5 titanium alloy' or 'α+β titanium alloy'. High speed machining of titanium alloys generates high temperatures in the cutting zone, promoting accelerated tool wear and reducing the efficiency in metal cutting. Consequently, the ability of the coolant to remove heat from the cutting zone plays an increasingly important role in the economics of the process as well as on the life of tool inserts. With the introduction of thru-tool coolant delivery, the coolant can now be delivered directly at the point of machining without having to flood the area of machining. This research tries to address the effects that high pressure and thru-tool coolant has on insert wear while end milling Ti-6Al-4V. The parameters used in this study are speed, feed, axial depth of cut, radial depth of cut and coolant pressure. A structured design of experiments along with a central composite design approach is used to determine the main effects of coolant pressure and its interactions with the remaining parameters. The results show that, within the parameters of this experiment, coolant pressure was not a significant main effect. However, pressure seems to react positively with feed rate. Contributions from this research can be used to recommend settings of the cutting factors in order to obtain the minimal tool wear.

  14. Measurement of core coolant flow velocities in PWRs using temperature: neutron noise cross correlation

    SciTech Connect

    Sweeney, F.J.; Upadhyaya, B.R.

    1982-01-01

    To study the relationship between the time delay inferred from this phase angle and core coolant flow velocities, noise measurements were performed at the Loss of Fluid Test Facility (LOFT) reactor and at a commercial PWR. In-core, self-powered neutron detector (SPND) noise at LOFT and ex-core ionization chamber noise at the commercial PWR were cross correlated with core exit temperature noise. Time delays were inferred from the slope of the phase angle versus frequency plots over the frequency range from 0.05 to 2.0 Hz.

  15. Development of sputtered techniques for thrust chambers. [coolant passage closing by triode sputtering

    NASA Technical Reports Server (NTRS)

    Mullaly, J. R.; Hecht, R. J.; Broch, J. W.; Allard, P. A.

    1976-01-01

    Procedures for closing out coolant passages in regeneratively cooled thrust chambers by triode sputtering, using post and hollow Cu-0.15 percent Zr cathodes are described. The effects of aluminum composite filler materials, substrate preparation, sputter cleaning, substrate bias current density and system geometry on closeout layer bond strength and structure are evaluated. High strength closeout layers were sputtered over aluminum fillers. The tensile strength and microstructure of continuously sputtered Cu-0.15 percent Zr deposits were determined. These continuous sputtered deposits were as thick as 0.75 cm. Tensile strengths were consistently twice as great as the strength of the material in wrought form.

  16. The development of Sn-Li coolant/breeding material for APEX/ALPS applications.

    SciTech Connect

    Sze, D.-K.

    1999-07-08

    A Sn-Li alloy has been identified to be a coolant/breeding material for D-T fusion applications. The key feature of this material is its very low vapor pressure, which will be very useful for free surface concepts employed in APEX, ALPS and inertial confinement fission. The vapor is dominated by lithium, which has very low Z. Initial assessment of the material indicates acceptable tritium breeding capability, high thermal conductivity, expected low tritium volubility, and expected low chemical reactivities with water and air. Some key concerns are the high activation and material compatibility issues. The initial assessment of this material, for fission applications, is presented in this paper.

  17. Cooled-turbine aerodynamic performance prediction from reduced primary to coolant total-temperature-ratio results

    NASA Technical Reports Server (NTRS)

    Goldman, L. J.

    1976-01-01

    The prediction of the cooled aerodynamic performance, for both stators and turbines, at actual primary to coolant inlet total temperature ratios from the results obtained at a reduced total temperature ratio is described. Theoretical and available experimental results were compared for convection film and transpiration cooled stator vanes and for a film cooled, single stage core turbine. For these tests the total temperature ratio varied from near 1.0 to about 2.7. The agreement between the theoretical and the experimental results was, in general, reasonable.

  18. Magnet safety and stability related coolant states: critical fluid dynamics at peak flux

    NASA Astrophysics Data System (ADS)

    Ravikumar, K. V.; Carandang, R. M.; Frederking, T. H. K.

    The stability of superconducting magnets is endangered under certain distinct conditions of the fluid serving as magnet coolant. A severe compromising of safety takes place at the peak heat flux of nucleate boiling. Progress in analysing first order phase transitions for cryoliquids and room temperature liquids, in the presence of heat flow, has led to better understanding of the parameters related to vapour bubble phenomena. The present work addresses the consequences arising from bubble frequency results, including model calculations for the effective masses of the saturated fluids involved in the two-phase transport at the peak flux.

  19. Failures of the thermal barriers of 900 MWe reactor coolant pumps

    SciTech Connect

    Peyrouty, P.

    1996-12-01

    This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible.

  20. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    SciTech Connect

    Al-Falahi, A.; Haennine, M.; Porkholm, K.

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  1. The heat transfer characteristic of the reactor coolant pump canned motor

    NASA Astrophysics Data System (ADS)

    Gu, X. Y.; Xu, R.; Tao, G.; Yang, Y. L.; Wang, D. Z.

    2016-05-01

    This paper deals with the heat transfer characteristic of the reactor coolant pump canned motor. The cooling of the canned motor is an important issue for the design of the pump. In order to analyze the heat transfer characteristic of the canned motor, firstly the electromagnetic field of the canned motor is calculated with finite element method, and the magnetic resistance loss is gotten, then the heat distribution of the canned motor is obtained based on the electromagnetic field, finally the flow field and temperature field of the canned motor is calculated with CFD methods. The calculation indicates that the highest temperature and highest temperature rising are both occurred at the end winding.

  2. Lamp system with conditioned water coolant and diffuse reflector of polytetrafluorethylene(PTFE)

    SciTech Connect

    Zapata, Luis E.; Hackel, Lloyd

    1999-01-01

    A lamp system with a very soft high-intensity output is provided over a large area by water cooling a long-arc lamp inside a diffuse reflector of polytetrafluorethylene (PTFE) and titanium dioxide (TiO.sub.2) white pigment. The water is kept clean and pure by a one micron particulate filter and an activated charcoal/ultraviolet irradiation system that circulates and de-ionizes and biologically sterilizes the coolant water at all times, even when the long-arc lamp is off.

  3. New Hydrophilic, Composite Membranes for Air Removal from Water Coolant Systems

    NASA Technical Reports Server (NTRS)

    Ritchie, Stephen M. C.; Luo, Qiang; Curtis, Salina S.; Holladay, Jon B.; Clark, Dallas W.

    2004-01-01

    Liquid coolants are commonly used as thermal transport media to increase efficiency and flexibility in aerospace vehicle design. The introduction of gas bubbles into the coolant can have negative consequences, including: loss of centrifugal pump prime, irregular sensor readings, and blockage of coolant flow to remote systems. One solution to mitigate these problems is the development of a passive gas removal device, or gas trap, installed in the flight cooling system. In this study, a new hydrophilic, composite membrane has been developed for passage of the coolant fluid and retention of gas bubbles. The trapped bubbles are subsequently vented from the system by a thin, hydrophobic, microporous membrane. The original design for this work employed a homogeneous membrane that was susceptible to fouling and pore plugging. Spare gas traps of this variety have degraded during storage, and recreation of the membranes has been complicated due to problems with polymer duplication and property variations in the final membranes. In this work, replacements have been developed based on deposition of a hydrophilic polymer on the bore-side of a porous polyethylene (PE) tube. The tube provides excellent chemical and mechanical stability, and the hydrophilic layer provides retention of gas bubbles. Preliminary results have shown that intimate contact is required between the deposited layer and the substrate to overcome material differences. This has been accomplished by presoaking the membrane tube in the solvent to raise its surface energy. Polymer solutions of various concentrations have been used to promote penetration of the polymer layer into the porous substrate and to control separation layer thickness. The resulting composite membranes have shown repeatable decrease in nitrogen permeability, which is indicative of a decrease in membrane pore size. Studies with water permeation have yielded similar results. We have observed some swelling of the added polymer layer, which

  4. Theoretical and experimental study of flow through turbine cascades with coolant flow injection

    NASA Technical Reports Server (NTRS)

    Tabakoff, W.; Hamed, A.

    1975-01-01

    An analytical study is presented which deals with the change in the outlet flow conditions from a turbine blade row due to coolant air injection through slots. The analysis applies to small secondary to primary mass flow ratios, taking into consideration the change in boundary layer development resulting from injection. The effects of injection location, injection flow angle and injected air momentum flux are investigated. The results of the present analysis are compared with the values obtained using the mixing theory method and experimental data from cold flow tests in a turbine cascade tunnel.

  5. Proceedings of the CSNI specialists meeting on fuel-coolant interactions

    SciTech Connect

    1994-03-01

    A specialists meeting on fuel-coolant interactions was held in Santa Barbara, CA from January 5-7, 1993. The meeting was sponsored by the United States Nuclear Regulatory Commission in collaboration with the Committee on the Safety of Nuclear Installation (CSNI) of the OECD Nuclear Energy Agency (NEA) and the University of California at Santa Barbara. The objectives of the meeting are to cross-fertilize on-going work, provide opportunities for mutual check points, seek to focus the technical issues on matters of practical significance and re-evaluate both the objectives as well as path of future research. Individual papers have been cataloged separately.

  6. Photoelectrochemical protection of stainless alloys from the stress-corrosion cracking in BWR primary coolant environment

    SciTech Connect

    Akashi, Masatsune; Iso-o, Hiroyuki; Kubota, Nobuhiko; Fukuda, Takanori; Ayabe, Muneo; Hirano, Kenji

    1995-12-31

    The feasibility of counteracting or preventing the stress-corrosion cracking in the BWR core internals by the photoelectrochemical method has been examined. For the purpose TiO{sub 2} semiconductor is noted for its capability of photo electrochemically inducing the water-oxidizing anodic reaction in low enough potential domain if supplied with a light of a wavelength shorter than 410 nm. This paper offers an empirical proof by showing that Type 304 stainless steel and Alloy 600 stainless alloy that have been plasma-spray coated with TiO{sub 2} film will do quite well in environments of BWR primary coolant.

  7. Extravehicular Mobility Unit (EMU) / International Space Station (ISS) Coolant Loop Failure and Recovery

    NASA Technical Reports Server (NTRS)

    Lewis, John F.; Cole, Harold; Cronin, Gary; Gazda, Daniel B.; Steele, John

    2006-01-01

    Following the Colombia accident, the Extravehicular Mobility Units (EMU) onboard ISS were unused for several months. Upon startup, the units experienced a failure in the coolant system. This failure resulted in the loss of Extravehicular Activity (EVA) capability from the US segment of ISS. With limited on-orbit evidence, a team of chemists, engineers, metallurgists, and microbiologists were able to identify the cause of the failure and develop recovery hardware and procedures. As a result of this work, the ISS crew regained the capability to perform EVAs from the US segment of the ISS.

  8. Self-actuated nuclear reactor shutdown system using induction pump to facilitate sensing of core coolant temperature

    DOEpatents

    Sievers, Robert K.; Cooper, Martin H.; Tupper, Robert B.

    1987-01-01

    A self-actuated shutdown system incorporated into a reactivity control assembly in a nuclear reactor includes pumping means for creating an auxiliary downward flow of a portion of the heated coolant exiting from the fuel assemblies disposed adjacent to the control assembly. The shutdown system includes a hollow tubular member which extends through the outlet of the control assembly top nozzle so as to define an outer annular flow channel through the top nozzle outlet separate from an inner flow channel for primary coolant flow through the control assembly. Also, a latching mechanism is disposed in an inner duct of the control assembly and is operable for holding absorber bundles in a raised position in the control assembly and for releasing them to drop them into the core of the reactor for shutdown purposes. The latching mechanism has an inner flow passage extending between and in flow communication with the absorber bundles and the inner flow channel of the top nozzle for accommodating primary coolant flow upwardly through the control assembly. Also, an outer flow passage separate from the inner flow passage extends through the latching mechanism between and in flow communication with the inner duct and the outer flow channel of the top nozzle for accommodating inflow of a portion of the heated coolant from the adjacent fuel assemblies. The latching mechanism contains a magnetic material sensitive to temperature and operable to cause mating or latching together of the components of the latching mechanism when the temperature sensed is below a known temperature and unmating or unlatching thereof when the temperature sensed is above a given temperature. The temperature sensitive magnetic material is positioned in communication with the heated coolant flow through the outer flow passage for directly sensing the temperature thereof. Finally, the pumping means includes a jet induction pump nozzle and diffuser disposed adjacent the bottom nozzle of the control assembly

  9. The calculational modeling of impurity mass transfer in NPP circuits with liquid metal coolant

    NASA Astrophysics Data System (ADS)

    Alexeev, V.; Kozlov, F.; Kumaev, V.; Orlova, E.; Klimanova, Yu; Torbenkova, I.

    2008-02-01

    The authors create three levels of models (one-dimensional, two-dimensional and three-dimensional) for estimation of impurity mass transfer in sodium circuits units as well as applicable computational programs. In the one-dimensional model the flow path elements are simulated by annular channels. The Lagrange coordinate system is used in the mathematical description of processes in channels. The two-dimensional model is based on the porous body approximation and enables one to simulate global spatial distributions of coolant flow velocity fields, coolant and fuel rods temperatures, and concentration distribution of transferred substances. The mathematical description of passive multicomponent impurity transfer is carried out using the system of the differential equations with sources and impurity diffusion, written for each component. The equations are solved by the finite-difference method. The developed version of the three-dimensional code is based on a general approach of the spatial three-dimensional description of thermal-hydraulic and mass-transfer processes in fuel rod bundles. The determining system of finite-difference equations of hydrodynamics and heat exchange is obtained using the control volume approach. As a result of the performed calculations, valuable data on corrosion products transfer in the primary circuit of the BN-600 reactor are obtained.

  10. Effect of reactor coolant radioactivity upon configuration feasibility for a nuclear electric propulsion vehicle

    NASA Technical Reports Server (NTRS)

    Soffer, L.; Wright, G. N.

    1973-01-01

    A preliminary shielding analysis was carried out for a conceptual nuclear electric propulsion vehicle designed to transport payloads from low earth orbit to synchronous orbit. The vehicle employed a thermionic nuclear reactor operating at 1575 kilowatts and generated 120 kilowatts of electricity for a round-trip mission time of 2000 hours. Propulsion was via axially directed ion engines employing 3300 pounds of mercury as a propellant. The vehicle configuration permitted a reactor shadow shield geometry using LiH and the mercury propellant for shielding. However, much of the radioactive NaK reactor coolant was unshielded and in close proximity to the power conditioning electronics. An estimate of the radioactivity of the NaK coolant was made and its unshielded dose rate to the power conditioning equipment calculated. It was found that the activated NaK contributed about three-fourths of the gamma dose constraint. The NaK dose was considered a sufficiently high fraction of the allowable gamma dose to necessitate modifications in configuration.

  11. Numerical Analysis of Coolant Flow and Heat Transfer in ITER Diagnostic First Wall

    DOE PAGESBeta

    Khodak, A.; Loesser, G.; Zhai, Y.; Udintsev, V.; Klabacha, J.; Wang, W.; Johnson, D.; Feder, R.

    2015-07-24

    We performed numerical simulations of the ITER Diagnostic First Wall (DFW) using ANSYS workbench. During operation DFW will include solid main body as well as liquid coolant. Thus thermal and hydraulic analysis of the DFW was performed using conjugated heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously fluid dynamics analysis was performed only in the liquid part. This approach includes interface between solid and liquid part of the systemAnalysis was performed using ANSYS CFX software. CFX software allows solution of heat transfer equations in solid and liquid part, and solution ofmore » the flow equations in the liquid part. Coolant flow in the DFW was assumed turbulent and was resolved using Reynolds averaged Navier-Stokes equations with Shear Stress Transport turbulence model. Meshing was performed using CFX method available within ANSYS. The data cloud for thermal loading consisting of volumetric heating and surface heating was imported into CFX Volumetric heating source was generated using Attila software. Surface heating was obtained using radiation heat transfer analysis. Our results allowed us to identify areas of excessive heating. Proposals for cooling channel relocation were made. Additional suggestions were made to improve hydraulic performance of the cooling system.« less

  12. Heat transfer and friction characteristics in rectangular channel using different coolants

    NASA Astrophysics Data System (ADS)

    Elwekeel, Fifi N. M.; Zheng, Qun; Abdala, Antar M. M.

    2013-12-01

    Several industrial applications such as electronic devices, heat exchangers, gas turbine blades, etc. need cooling processes. The internal cooling technique is proper for some applications. In the present work, computational simulations were made using ANSYS CFX to predict the improvements of internal heat transfer in the rectangular ribbed channel using different coolants. Several coolants such as air, steam, air/mist and steam/mist were investigated. The shear stress transport model (SST) is selected by comparing the standard k-ω and Omega Reynolds Stress (ωRS) turbulence models with experimental results.The results indicate that the heat transfer coefficients are enhanced in the ribbed channel while injecting small amounts of mist. The heat transfer coefficients of air/mist, steam and steam/mist increase by 12.5%, 49.5% and 107% over that of air, respectively. Furthermore, in comparison to air, the air/mist heat transfer coefficient enhances by about 1.05 to 1.14 times when the mist mass fraction increases from 2% to 8%, respectively. The steam/mist heat transfer coefficient increases by about 1.12 to 1.27 times higher than that of steam over the considered range of mist mass fraction.

  13. A review of the effects of coolant environments on the fatigue life of LWR structural materials.

    SciTech Connect

    Chopra, O. K.; Shack, W. J.

    2009-04-01

    The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code specifies design curves for the fatigue life of structural materials in nuclear power plants. However, the effects of light water reactor (LWR) coolant environments were not explicitly considered in the development of the design curves. The existing fatigue-strain-versus-life ({var_epsilon}-N) data indicate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. Under certain environmental and loading conditions, fatigue lives in water relative to those in air can be a factor of 15 lower for austenitic stainless steels and a factor of {approx}30 lower for carbon and low-alloy steels. This paper reviews the current technical basis for the understanding of the fatigue of piping and pressure vessel steels in LWR environments. The existing fatigue {var_epsilon}-N data have been evaluated to identify the various material, environmental, and loading parameters that influence fatigue crack initiation and to establish the effects of key parameters on the fatigue life of these steels. Statistical models are presented for estimating fatigue life as a function of material, loading, and environmental conditions. An environmental fatigue correction factor for incorporating the effects of LWR environments into ASME Code fatigue evaluations is described. This paper also presents a critical review of the ASME Code fatigue design margins of 2 on stress (or strain) and 20 on life and assesses the possible conservatism in the current choice of design margins.

  14. Improvised source of water coolant for ultrasonic scaler: an appropriate technology in underserved communities.

    PubMed

    Ibiyemi, Olushola; Taiwo, Juliana O; Oke, Gbemisola A

    2012-10-01

    Traditionally dental plaque, calculus and stains have been removed by scaling and polishing manually with hand instruments such as curettes, chisels, hoes and scalers. However, ultrasonic scaling is becoming the preferred method of initial periodontal treatment and maintenance, due to improved patient and operator comfort. Ultrasonic scaling can be performed effectively using pipe-borne water as coolant. However, such a water supply is unavailable in many rural dental clinics, especially in underserved communities in Nigeria. This article reports on an improvised source of water coolant, designed and fabricated to make modern, easy and effective plaque control available to people in communities where there is no pipe-borne water. The device will improve operator efficiency in tooth cleaning and patient compliance with treatment. Due to its simple design but effective function, the device is ideal for use in Nigeria's primary healthcare delivery program, offering enhanced preventive and curative services to remote, rural and semi-urban populations. In doing so oral health can be improved with a reduction in the incidence of oral diseases. PMID:23276120

  15. Two-phase performance of scale models of a primary coolant pump. Final report

    SciTech Connect

    Kamath, P.S.; Swift, W.L.

    1982-09-01

    Scale models of PWR primary coolant pumps were tested in steady and transient two-phase flows in order to generate a data base to aid in the development and assessment of pump performance models for use in computer codes for the analysis of postulated Loss-of-Coolant Accidents (LOCA). This report summarizes and unifies the single and two-phase air/water and steam/water performance data on the relatively high specific speed pumps (4200 rpm (US gpm) /sup 1/2//ft /sup 3/4/) tested in these programs. These data are compared with those acquired from tests on the lower specific speed Semiscale pump (926 rpm (US gpm)/sup 1/2//ft/sup 3/4/) to better understand the mechanism of performance degradation with increasing void fraction. The study revealed that scaling down the size of the pump while maintaining the same design specific speed produces very similar performance characteristics both in single and two-phase flows. Effects due to size and operating speed were not discernible within the range of test conditions and within experimental uncertainties. System pressure appears to affect the rate of degradation as a function of void fraction. The report includes a survey of the existing two-phase pump performance correlations. A correlation synthesized from the B and W, C-E and Creare two-phase data is also presented.

  16. Gas production and behavior in the coolant of the SP-100 Space Nuclear Power System

    SciTech Connect

    McGhee, J.M.

    1989-08-01

    The radiologic generation and subsequent behavior of helium gas in the lithium coolant of SP-100 class space nuclear power reactors was investigated analytically in a two part study. Part One of the study consisted of a calculation of coolant radiologic helium gas production rates in a SP-100 class reactor using the discrete ordinates code TWODANT. Cross sections were developed from ENDF/B-V data via the MATXS6s master cross section library. Cross sections were self shielded assuming one homogeneous core region, and doppler broadened to 1300 K using the cross section preparation code TRANSX. Calculations were performed using an S{sub 4}/P{sub 1} approximation and 80 neutron energy groups. Part Two of the study consisted of a theoretical investigation into the behavior of helium gas in the primary loop of lithium cooled space reactors. The SP-100 space power system was used as a representative of such a system. Topics investigated included: (1) heterogeneous and homogeneous nucleation; (2) bubble growth/collapse by diffusion, mechanical temperature/pressure effects, and coalescence; and, (3) the effects on bubble distribution of microgravity, magnetic fields, and inertially induced buoyancy. 104 refs., 78 figs., 28 tabs.

  17. Fluid flow analysis of the SSME high pressure fuel and oxidizer turbine coolant systems

    NASA Technical Reports Server (NTRS)

    Teal, G. A.

    1989-01-01

    The objective is to provide improved analysis capability for the Space Shuttle Main Engine (SSME) high pressure fuel and oxidizer turbine coolant systems. Each of the systems was analyzed to determine fluid flow rate and thermodynamic and transport properties at all key points in the systems. Existing computer codes were used as a baseline for these analyses. These codes were modified to provide improved analysis capability. The major areas of improvement are listed. A review of the drawings was performed, and pertinent geometry changes were included in the models. Improvements were made in the calculation of thermodynamic and transport properties for a mixture of hydrogen and steam. A one-dimensional turbine model for each system is included as a subroutine to each code. This provides a closed loop analysis with a minimum of required boundary conditions as input. An improved labyrinth seal model is included in the high pressure fuel turbine coolant model. The modifications and the analysis results are presented in detail.

  18. Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop

    SciTech Connect

    Donna Post Guillen

    2012-11-01

    This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

  19. Numerical Analysis of Coolant Flow and Heat Transfer in ITER Diagnostic First Wall

    SciTech Connect

    Khodak, A.; Loesser, G.; Zhai, Y.; Udintsev, V.; Klabacha, J.; Wang, W.; Johnson, D.; Feder, R.

    2015-07-24

    We performed numerical simulations of the ITER Diagnostic First Wall (DFW) using ANSYS workbench. During operation DFW will include solid main body as well as liquid coolant. Thus thermal and hydraulic analysis of the DFW was performed using conjugated heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously fluid dynamics analysis was performed only in the liquid part. This approach includes interface between solid and liquid part of the systemAnalysis was performed using ANSYS CFX software. CFX software allows solution of heat transfer equations in solid and liquid part, and solution of the flow equations in the liquid part. Coolant flow in the DFW was assumed turbulent and was resolved using Reynolds averaged Navier-Stokes equations with Shear Stress Transport turbulence model. Meshing was performed using CFX method available within ANSYS. The data cloud for thermal loading consisting of volumetric heating and surface heating was imported into CFX Volumetric heating source was generated using Attila software. Surface heating was obtained using radiation heat transfer analysis. Our results allowed us to identify areas of excessive heating. Proposals for cooling channel relocation were made. Additional suggestions were made to improve hydraulic performance of the cooling system.

  20. Methods for incorporating effects of LWR coolant environment into ASME code fatigue evaluations.

    SciTech Connect

    Chopra, O. K.

    1999-04-15

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Appendix I to Section HI of the Code specifies design fatigue curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Recent test data illustrate potentially significant effects of LWR environments on the fatigue resistance of carbon and low-alloy steels and austenitic stainless steels (SSs). Under certain loading and environmental conditions, fatigue lives of carbon and low-alloy steels can be a factor of {approx}70 lower in an LWR environment than in air. These results raise the issue of whether the design fatigue curves in Section III are appropriate for the intended purpose. This paper presents the two methods that have been proposed for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations. The mechanisms of fatigue crack initiation in carbon and low-alloy steels and austenitic SSs in LWR environments are discussed.

  1. Distributed Raman temperature measurement system for monitoring of nuclear power plant coolant loops

    NASA Astrophysics Data System (ADS)

    Jensen, Fredrik B. H.; Takada, Eiji; Nakazawa, Masaharu; Kakuta, Tsunemi; Yamamoto, Satoshi

    1996-09-01

    A distributed temperature sensor based on Raman scattering in optical fibers has been tested for use as coolant loop monitor in nuclear power plants. Different types of pure- silica-core, polyimide-coated fibers have been subjected to 60Co-gamma-ray and fission-reactor irradiation at varying temperatures. 60Co-gamma-ray irradiations at dose rates from 4.8 kR/h up to 1 MR/h were done. Simultaneous gamma-ray and high temperature experiments up to 300 degrees Celsius have also been performed. The induced loss of the tested fibers was found to saturate with increasing dose at the anti-Stokes and Stokes wavelengths. This feature was then made use of to develop a model for radiation induced loss which was used to make system lifetime predictions. It has also been demonstrated that the induced loss of the optical fibers is favorably affected by high-temperature use. A 10-fold decrease in the radiation- induced loss levels when the system was operated at 300 degrees Celsius was observed, as compared with room- temperature operation. The experiments have shown that with a pure-silica-core, polyimide-coated fiber the temperature sensing capabilities of the RDTS will not be degraded excessively if used at primary coolant loops with an expected upper radiation level of 200 R/hr.

  2. Corrosion of Ferritic Steels in High Temperature Molten Salt Coolants for Nuclear Applications

    SciTech Connect

    Farmer, J; El-Dasher, B; de Caro, M S; Ferreira, J

    2008-11-25

    Corrosion of ferritic steels in high temperature molten fluoride salts may limit the life of advanced reactors, including some hybrid systems that are now under consideration. In some cases, the steel may be protected through galvanic coupling with other less noble materials with special neutronic properties such a beryllium. This paper reports the development of a model for predicting corrosion rates for various ferritic steels, with and without oxide dispersion strengthening, in FLiBe (Li{sub 2}BeF{sub 4}) and FLiNaK (Li-Na-K-F) coolants at temperatures up to 800 C. Mixed potential theory is used to account for the protection of steel by beryllium, Tafel kinetics are used to predict rates of dissolution as a function of temperature and potential, and the thinning of the mass-transfer boundary layer with increasing Reynolds number is accounted for with dimensionless correlations. The model also accounts for the deceleration of corrosion as the coolants become saturated with dissolved chromium and iron. This paper also reports electrochemical impedance spectroscopy of steels at their corrosion potentials in high-temperature molten salt environments, with the complex impedance spectra interpreted in terms of the interfacial charge transfer resistance and capacitance, as well as the electrolyte conductivity. Such in situ measurement techniques provide valuable insight into the degradation of materials under realistic conditions.

  3. An investigation of core liquid level depression in small break loss-of-coolant accidents

    SciTech Connect

    Schultz, R.R.; Watkins, J.C. ); Motley, F.E.; Stumpf, H. ); Chen, Y.S. . Div. of Systems Research)

    1991-08-01

    Core liquid level depression can result in partial core dryout and heatup early in a small break loss-of-coolant accident (SBLOCA) transient. Such behavior occurs when steam, trapped in the upper regions of the reactor primary system (between the loop seal and the core inventory), moves coolant out of the core region and uncovers the rod upper elevations. The net result is core liquid level depression. Core liquid level depression and subsequent core heatups are investigated using subscale data from the ROSA-IV Program's 1/48-scale Large Scale Test Facility (LSTF) and the 1/1705-scale Semiscale facility. Both facilities are Westinghouse-type, four-loop, pressurized water reactor simulators. The depression phenomena and factors which influence the minimum core level are described and illustrated using examples from the data. Analyses of the subject experiments, conducted using the TRAC-PF1/MOD1 (Version 12.7) thermal-hydraulic code, are also described and summarized. Finally, the response of a typical Westinghouse four-loop plant (RESAR-3S) was calculated to qualitatively study coal liquid level depression in a full-scale system. 31 refs., 37 figs., 6 tabs.

  4. In-vessel ITER tubing failure rates for selected materials and coolants

    SciTech Connect

    Marshall, T.D.; Cadwallader, L.C.

    1994-03-01

    Several materials have been suggested for fabrication of ITER in-vessel coolant tubing: beryllium, copper, Inconel, niobium, stainless steel, titanium, and vanadium. This report generates failure rates for the materials to identify the best performer from an operational safety and availability perspective. Coolant types considered in this report are helium gas, liquid lithium, liquid sodium, and water. Failure rates for the materials are generated by including the influence of ITER`s operating environment and anticipated tubing failure mechanisms with industrial operating experience failure rates. The analyses define tubing failure mechanisms for ITER as: intergranular attack, flow erosion, helium induced swelling, hydrogen damage, neutron irradiation embrittlement, cyclic fatigue, and thermal cycling. K-factors, multipliers, are developed to model each failure mechanism and are applied to industrial operating experience failure rates to generate tubing failure rates for ITER. The generated failure rates identify the best performer by its expected reliability. With an average leakage failure rate of 3.1e-10(m-hr){sup {minus}1}and an average rupture failure rate of 3.1e-11(m-hr){sup {minus}1}, titanium proved to be the best performer of the tubing materials. The failure rates generated in this report are intended to serve as comparison references for design safety and optimization studies. Actual material testing and analyses are required to validate the failure rates.

  5. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    SciTech Connect

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177{degrees}C (350{degrees}F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program.

  6. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    SciTech Connect

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J; Wilson, C L

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  7. Gas production and behavior in the coolant of the SP-100 space nuclear power system

    NASA Astrophysics Data System (ADS)

    McGhee, John Morton

    1989-08-01

    The radiologic generation and subsequent behavior of helium gas in the lithium coolant of SP-100 class space nuclear power reactors was investigated analytically in a two part study. Part One of the study consisted of a calculation of coolant radiologic helium gas production rates in a SP-100 class reactor using the discrete ordinates code TWODANT. Cross sections were developed from ENDF/B-V data via the MATXS6s master cross section library. Cross sections were self shielded assuming one homogeneous core region, and Doppler broadened to 1300 K using the cross section preparation code TRANSX. Calculations were performed using an S sub 4/P sub 1 approximation and 80 neutron energy groups. Part Two of the study consisted of a theoretical investigation into the behavior of helium gas in the primary loop of lithium cooled space reactors. The SP-100 space power system was used as a representative of such a system. Topics investigated included: (1) heterogeneous and homogeneous nucleation; (2) bubble growth/collapse by diffusion, mechanical temperature/pressure effects, and coalescence; and, (3) the effects on bubble distribution of microgravity, magnetic fields, and inertially induced buoyancy.

  8. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)

    SciTech Connect

    Williams, D.F.

    2006-03-24

    The Advanced High-Temperature Reactor (AHTR) is a novel reactor design that utilizes the graphite-matrix high-temperature fuel of helium-cooled reactors, but provides cooling with a high-temperature fluoride salt. For applications at temperatures greater than 900 C the AHTR is also referred to as a Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). This report provides an assessment of candidate salts proposed as the primary coolant for the AHTR based upon a review of physical properties, nuclear properties, and chemical factors. The physical properties most relevant for coolant service were reviewed. Key chemical factors that influence material compatibility were also analyzed for the purpose of screening salt candidates. Some simple screening factors related to the nuclear properties of salts were also developed. The moderating ratio and neutron-absorption cross-section were compiled for each salt. The short-lived activation products, long-lived transmutation activity, and reactivity coefficients associated with various salt candidates were estimated using a computational model. Table A presents a summary of the properties of the candidate coolant salts. Certain factors in this table, such as melting point, vapor pressure, and nuclear properties, can be viewed as stand-alone parameters for screening candidates. Heat-transfer properties are considered as a group in Sect. 3 in order to evaluate the combined effects of various factors. In the course of this review, it became apparent that the state of the properties database was strong in some areas and weak in others. A qualitative map of the state of the database and predictive capabilities is given in Table B. It is apparent that the property of thermal conductivity has the greatest uncertainty and is the most difficult to measure. The database, with respect to heat capacity, can be improved with modern instruments and modest effort. In general, ''lighter'' (low-Z) salts tend to exhibit better heat

  9. Performance of the Extravehicular Mobility Unit (EMU) Airlock Coolant Loop Remediation (A/L CLR) Hardware - Final

    NASA Technical Reports Server (NTRS)

    Steele, John W.; Rector, Tony; Gazda, Daniel; Lewis, John

    2011-01-01

    An EMU water processing kit (Airlock Coolant Loop Recovery -- A/L CLR) was developed as a corrective action to Extravehicular Mobility Unit (EMU) coolant flow disruptions experienced on the International Space Station (ISS) in May of 2004 and thereafter. A conservative duty cycle and set of use parameters for A/L CLR use and component life were initially developed and implemented based on prior analysis results and analytical modeling. Several initiatives were undertaken to optimize the duty cycle and use parameters of the hardware. Examination of post-flight samples and EMU Coolant Loop hardware provided invaluable information on the performance of the A/L CLR and has allowed for an optimization of the process. The intent of this paper is to detail the evolution of the A/L CLR hardware, efforts to optimize the duty cycle and use parameters, and the final recommendations for implementation in the post-Shuttle retirement era.

  10. Copper-based micro-channel cooler reliably operated using solutions of distilled-water and ethanol as a coolant

    NASA Astrophysics Data System (ADS)

    Chin, A. K.; Nelson, A.; Chin, R. H.; Bertaska, R.; Jacob, J. H.

    2015-03-01

    Copper-based micro-channel coolers (Cu-MCC) are the lowest thermal-resistance heat-sinks for high-power laserdiode (LD) bars. Presently, the resistivity, pH and oxygen content of the de-ionized water coolant, must be actively controlled to minimize cooler failure by corrosion and electro-corrosion. Additionally, the water must be constantly exposed to ultraviolet radiation to limit the growth of micro-organisms that may clog the micro-channels. In this study, we report the reliable, care-free operation of LD-bars attached to Cu-MCCs, using a solution of distilledwater and ethanol as the coolant. This coolant meets the storage requirements of Mil-Std 810G, e.g. exposure to a storage temperature as low as -51°C and no growth of micro-organisms during passive storage.

  11. Modeling the activity of 129I and 137Cs in the primary coolant and CVCS resin of an operating PWR

    NASA Astrophysics Data System (ADS)

    Hwang, K. H.; Lee, K. J.

    2006-04-01

    Mathematical models have been developed to describe the activities of 129I and 137Cs in the primary coolant and resin of the chemical and volume control system (CVCS) during constant power operation in a pressurized water reactor (PWR). The models, which account for the source releases from defective fuel rod(s) and tramp uranium, rely on the contribution of CVCS resin and boron recovery system as a removal process, and differences in behavior for each nuclide. The current models were validated through measured coolant activities of 137Cs. The resultant scaling factors agree reasonably well with the results of the test resin of the coolant and the actual resins from the PWRs of other countries.

  12. Probability of pipe failure in the reactor coolant loops of Babcock and Wilcox PWR plants. Volume 1. Summary report

    SciTech Connect

    Holman, G.S.; Chou, C.K.

    1986-05-01

    As part of its reevaluation of the double-ended guillotine break (DEGB) of reactor coolant piping as a design basis event for nuclear power plants, the US Nuclear Regulatory Commission (NRC) contracted the Lawrence Livermore National Laboratory (LLNL) to estimate the probability of occurrence of a DEGB, and to assess the effect that earthquakes have on DEGB probability. This report describes an evaluation of reactor coolant loop piping in PWR plants having nuclear steam supply systems designed by Babcock and Wilcox. Two causes of pipe break were considered: pipe fracture due to the growth of cracks at welded joints (''direct'' DEGB), and pipe rupture indirectly caused by failure of heavy component supports due to an earthquake (''indirect'' DEGB). Unlike in earlier evaluations of Westinghouse and Combustion Engineering reactor coolant loop piping, in which the probability of direct DEGB had been explicitly estimated using a probabilistic fracture mechanics model, no detailed fracture mechanics calculations were performed. Instead, a comparison of relevant plant data, mainly reactor coolant loop stresses, for one representative B and W plant with equivalent information for Westinghouse and C-E systems inferred that the probability of direct DEGB should be similarly low (less than le-10 per reactor year). The probability of indirect DEGB, on the other hand, was explicitly estimated for two representative plants. The results of this study indicate that the probability of a DEGB form either cause is very low for reactor coolant loop piping in these specific plants and, because of similarity in design, infer that the probability of DEGB is generally very low in B and W reactor coolant loop piping. The NRC should therefore consider eliminating DEGB as a design basis event in favor of more realistic criteria. 13 refs., 9 tabs.

  13. Three-dimensional turbulent-mixing-length modeling for discrete-hole coolant injection into a crossflow

    NASA Technical Reports Server (NTRS)

    Wang, C. R.; Papell, S. S.

    1983-01-01

    Three dimensional mixing length models of a flow field immediately downstream of coolant injection through a discrete circular hole at a 30 deg angle into a crossflow were derived from the measurements of turbulence intensity. To verify their effectiveness, the models were used to estimate the anisotropic turbulent effects in a simplified theoretical and numerical analysis to compute the velocity and temperature fields. With small coolant injection mass flow rate and constant surface temperature, numerical results of the local crossflow streamwise velocity component and surface heat transfer rate are consistent with the velocity measurement and the surface film cooling effectiveness distributions reported in previous studies.

  14. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    NASA Technical Reports Server (NTRS)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  15. Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

    SciTech Connect

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

    The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

  16. Apparatus for suppressing formation of vortices in the coolant fluid of a nuclear reactor and associated method

    DOEpatents

    Ekeroth, Douglas E.; Garner, Daniel C.; Hopkins, Ronald J.; Land, John T.

    1993-01-01

    An apparatus and method are provided for suppressing the formation of vortices in circulating coolant fluid of a nuclear reactor. A vortex-suppressing plate having a plurality of openings therein is suspended within the lower plenum of a reactor vessel below and generally parallel to the main core support of the reactor. The plate is positioned so as to intersect vortices which may form in the circulating reactor coolant fluid. The intersection of the plate with such vortices disrupts the rotational flow pattern of the vortices, thereby disrupting the formation thereof.

  17. Apparatus for suppressing formation of vortices in the coolant fluid of a nuclear reactor and associated method

    DOEpatents

    Ekeroth, D.E.; Garner, D.C.; Hopkins, R.J.; Land, J.T.

    1993-11-30

    An apparatus and method are provided for suppressing the formation of vortices in circulating coolant fluid of a nuclear reactor. A vortex-suppressing plate having a plurality of openings therein is suspended within the lower plenum of a reactor vessel below and generally parallel to the main core support of the reactor. The plate is positioned so as to intersect vortices which may form in the circulating reactor coolant fluid. The intersection of the plate with such vortices disrupts the rotational flow pattern of the vortices, thereby disrupting the formation thereof. 3 figures.

  18. Modelling of heat transfer in composite bodies reinforced with tubes with incompressible liquid coolant moving in laminar regime

    NASA Astrophysics Data System (ADS)

    Yankovskii, A. P.

    2015-01-01

    The equations were obtained for description of heat transfer in composite bodies with reinforcement with a set of smooth tubes with pumping of incompressible liquid coolant. The appropriate boundary-value problem of heat transfer was formulated with the following qualitative analysis. The steady temperature fields in cylindrical shells with spiral-shaped tube reinforcement (with liquid coolant flow) were simulated. The influence on the temperature fields from the side of reinforcement geometry parameters, flow direction and velocity, tubes cross sections was studied. It was found that these characteristics allow significant variation of temperature field in its gradients within the composite item: this open ways for effective control of temperature fields.

  19. Study on the effect of the impeller and diffuser blade number on reactor coolant pump performances

    NASA Astrophysics Data System (ADS)

    Long, Y.; Yin, J. L.; Wang, D. Z.; Li, T. B.

    2016-05-01

    In this paper, CFD approach was employed to study how the blade number of impeller and diffuser influences reactor coolant pump performances. The three-dimensional pump internal flow channel was modelled by pro/E software, Reynolds-averaged Naiver-Stokes equations with the k-ε turbulence model were solved by the computational fluid dynamics software CFX. By post-processing on the numerical results, the performance curves of reactor coolant pump were obtained. The results are as follows, with the blade number of the impeller increasing, the head of the pump with different diffuser universally increases in the 8Q n∼1.2Q n conditions, and at different blade number of the diffuser, the head increases with the blade number of the impeller increasing. In 1.0Q n condition, when the blades number combination of impeller and diffuser chooses 4+16, 7+14 and 6+18, the head curves exist singular points. In 1.2Q n condition, the head curve still exists singular point in 6+18. With the blade number of the impeller increasing, the efficiency of the pump with different diffuser universally decreases in the 0.8Q n and 1.0Q n conditions, but in 1.2Q n condition, the efficiency of the pump with different diffuser universally increases. In 1.0Q n condition, the impellers of 4 and 5 blades are better. When the blade number combination of impeller and diffuser choose 4+11, 4+17, 4+18, 5+12, 5+17 and 5+18, the efficiencies relatively have higher values. With the blade number of the impeller increasing, the hydraulic shaft power of the pump with different diffuser universally increases in the 0.8Q n∼1.2Q n conditions, and with the blade number of the diffuser increasing, the power of different impeller overall has small fluctuation, but tends to be uniform. This means the increase of the diffuser blade number has less influence on shaft power.The influence on the head and flow by the matching relationship of the blades number between impeller and diffuser is very complicated, which

  20. Safety Analysis of the US Dual Coolant Liquid Lead-Lithium ITER Test Blanket Module

    SciTech Connect

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2006-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER Test Blanket Module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER International Team (IT) to address specific reactor safety concerns, such as VV pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system, and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  1. Cobalt-60 simulation of LOCA (loss of coolant accident) radiation effects

    SciTech Connect

    Buckalew, W.H.

    1989-07-01

    The consequences of simulating nuclear reactor loss of coolant accident (LOCA) radiation effects with Cobalt-60 gamma ray irradiators have been investigated. Based on radiation induced damage in polymer base materials, it was demonstrated that electron/photon induced radiation damage could be related on the basis of average absorbed radiation dose. This result was used to estimate the relative effectiveness of the mixed beta/gamma LOCA and Cobalt-60 radiation environments to damage both bare and jacketed polymer base electrical insulation materials. From the results obtained, it is concluded that present simulation techniques are a conservative method for simulating LOCA radiation effects and that the practices have probably substantially overstressed both bare and jacketed materials during qualification testing. 9 refs., 8 figs., 5 tabs.

  2. Radionuclide analyses taken during primary coolant decontamination at Three Mile Island indicate general circulation

    SciTech Connect

    Hofstetter, K.J.; Baston, V.F.; Hitz, C.G.; Malinauskas, A.P.

    1983-12-01

    Radionuclide concentration data taken during decontamination of the primary reactor coolant system at Three Mile Island by a feed-and-bleed process have provided information on future defueling operations. Analysis of the radiocesium concentrations in samples taken at the letdown point indicates general circulation within the primary system, including the reactor vessel and both steam generators. A standard dilution model with parameters consistent with engineering estimates (volume, flow rate, etc.) accurately predicts the radiocesium decontamination rates. Unlike cesium, the behavior of other principal soluble radionuclides (/sup 90/Sr and /sup 3/H) cannot be readily described by dilution theory. A significant appearance rate is observed for /sup 90/Sr suggesting a chemical solubility mechanism. The use of processed water containing high /sup 3/H for makeup causes uncertainty in the interpretation of the /sup 3/H analysis.

  3. TRAC large-break loss-of-coolant accident analysis for the AP600 design

    SciTech Connect

    Lime, J.F.; Boyack, B.E.

    1994-02-01

    This report discusses a TRAC model of the Westinghouse AP600 advanced reactor design which has been developed for analyzing large-break loss-of-coolant accident (LBLOCA) transients. A preliminary LBLOCA calculation of a 80% cold-leg break has been performed with TRAC-PF1/MOD2. The 80% break size was calculated by Westinghouse to be the most severe large-break size. The LBLOCA transient was calculated to 92 s. Peak clad temperatures (PCT) were well below the Appendix K limit of 1478 K (2200{degrees}F). Transient event times and PCT for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code.

  4. Loss-of-coolant accident mitigation for the Advanced Neutron Source Reactor

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

    1994-09-01

    A RELAP5 Advanced Neutron Source Reactor system model has been developed for the conceptual design safety analysis. Three major regions modeled are the core, the heat exchanger loops, and letdown/pressurizing system. The model has been used to examine design alternatives for mitigation of loss-of-coolant accident (LOCA) transients. The safety margins to the flow excursion limit and critical heat flux are presented. The results show that the core can survive an instantaneous double-ended guillotine of the core outlet piping break (610 mm-diameter) provided a cavitating venturi is employed. RELAP5 calculations were also used to determine the effects of using a non-instantaneous break opening times. Both break opening time and break formation characteristics were included in these parametric calculations. Accumulator optimization studies were also performed which suggest that an optimum accumulator bubble size exists which improves system performance under some break scenarios.

  5. Conceptual design loss-of-coolant accident analysis for the Advanced Neutron Source reactor

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. )

    1994-01-01

    A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided.

  6. Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs

    SciTech Connect

    Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S.

    2012-07-01

    An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO{sub 2} volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)

  7. Failure probability of PWR reactor coolant loop piping. [Double-ended guillotine break

    SciTech Connect

    Lo, T.; Woo, H.H.; Holman, G.S.; Chou, C.K.

    1984-02-01

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria.

  8. Whole-core neutron transport calculations without fuel-coolant homogenization

    SciTech Connect

    Smith, M. A.; Tsoulfanidis, N.; Lewis, E. E.; Palmiotti, G.; Taiwo, T. A.

    2000-02-10

    The variational nodal method implemented in the VARIANT code is generalized to perform full core transport calculations without spatial homogenization of cross sections at either the fuel-pin cell or fuel assembly level. The node size is chosen to correspond to one fuel-pin cell in the radial plane. Each node is divided into triangular finite subelements, with the interior spatial flux distribution represented by piecewise linear trial functions. The step change in the cross sections at the fuel-coolant interface can thus be represented explicitly in global calculations while retaining the fill spherical harmonics capability of VARIANT. The resulting method is applied to a two-dimensional seven-group representation of a LWR containing MOX fuel assemblies. Comparisons are made of the accuracy of various space-angle approximations and of the corresponding CPU times.

  9. An overview of fuel-coolant interactions (FCI) research at NRC

    SciTech Connect

    Basu, S.; Speis, T.P.

    1996-03-01

    An overview of the fuel-coolant interactions (FCI) research programs sponsored by the U.S. Nuclear Regulatory Commission (NRC) is presented in this paper. A historical perspective of the program is provided with particular reference to in-vessel steam explosion and its consequences on the reactor pressure vessel and the containment integrity. Emphasis is placed on research in the last decade involving fundamentals of FCI phenomenology, namely, premixing, triggering, propagation, and energetics. The status of the current understanding of in-vessel steam explosion-induced containment failure (alpha-mode) issue, and other FCI issues related to reactor vessel and containment integrity are reported, including the extensive review and discussion of these issues at the recently held second Steam Explosion Review Group Workshop (SERG-2). Ongoing NRC research programs are discussed in detail. Future research programs including those recommended at the SERG-2 workshop are outlined.

  10. Nonintrusive Flow Rate Determination Through Space Shuttle Water Coolant Loop Floodlight Coldplate

    NASA Technical Reports Server (NTRS)

    Werlink, Rudolph; Johnson, Harry; Margasahayam, Ravi

    1997-01-01

    Using a Nonintrusive Flow Measurement System (NFMS), the flow rates through the Space Shuttle water coolant coldplate were determined. The objective of this in situ flow measurement was to prove or disprove a potential block inside the affected coldplate had contributed to a reduced flow rate and the subsequent ice formation on the Space Shuttle Discovery. Flow through the coldplate was originally calculated to be 35 to 38 pounds per hour. This application of ultrasonic technology advanced the envelope of flow measurements through use of 1/4-inch-diameter tubing, which resulted in extremely low flow velocities (5 to 30 pounds per hour). In situ measurements on the orbiters Discovery and Atlantis indicated both vehicles, on the average, experienced similar flow rates through the coldplate (around 25 pounds per hour), but lower rates than the designed flow. Based on the noninvasive checks, further invasive troubleshooting was eliminated. Permanent monitoring using the NFMS was recommended.

  11. Validation of advanced NSSS simulator model for loss-of-coolant accidents

    SciTech Connect

    Kao, S.P.; Chang, S.K.; Huang, H.C.

    1995-09-01

    The replacement of the NSSS (Nuclear Steam Supply System) model on the Millstone 2 full-scope simulator has significantly increased its fidelity to simulate adverse conditions in the RCS. The new simulator NSSS model is a real-time derivative of the Nuclear Plant Analyzer by ABB. The thermal-hydraulic model is a five-equation, non-homogeneous model for water, steam, and non-condensible gases. The neutronic model is a three-dimensional nodal diffusion model. In order to certify the new NSSS model for operator training, an extensive validation effort has been performed by benchmarking the model performance against RELAP5/MOD2. This paper presents the validation results for the cases of small-and large-break loss-of-coolant accidents (LOCA). Detailed comparisons in the phenomena of reflux-condensation, phase separation, and two-phase natural circulation are discussed.

  12. Reactor coolant pump startup under degraded conditions in a scaled OTSG lowered loop PWR

    SciTech Connect

    Tafreshi, A.M.; Marzo, M. di

    1996-12-31

    After a SB-LOCA or improper maintenance activities, the potential exists for a non-uniform distribution of boric acid in a PWR coolant system. This in turn presents the possibility of a reactivity excursion if sufficient volumes of boron-dilute water are transported into the core region without having first undergone substantial mixing. A research program is being conducted at the University of Maryland College Park (UMCP) 2 x 4 thermal-hydraulic test facility to assess the generation, transport and mixing of boron-dilute volumes. Start up of a pump and flow of a boron free slug of water in the cold leg and subsequent transport to the core downcomer in the facility is investigated here.

  13. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

    SciTech Connect

    Heames, T.J. ); Williams, D.A.; Johns, N.A.; Chown, N.M. ); Bixler, N.E.; Grimley, A.J. ); Wheatley, C.J. )

    1990-10-01

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs.

  14. Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue

    SciTech Connect

    Atwood, Corwin Lee; Shah, Vikram Naginbhai; Galyean, William Jospeh

    1999-09-01

    We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.

  15. Prototypic Thermal-Hydraulic Experiment in NRU to Simulate Loss-of-Coolant Accidents

    SciTech Connect

    Mohr, C. L.; Hesson, G. M.; Russcher, G. E.; Marsh, R. K.; King, L. L.; Wildung, N. J.; Rausch, W. N.; Bennett, W. D.

    1981-04-01

    Quick-look test results are reported for the initial test series of the Loss-of-Coolant Accident (LOCA) Simulation in the National Research Universal {NRU) test program, conducted by Pacific Northwest Laboratory (PNL) for the U.S. Nuclear Regulatory Commission (NRC). This test was devoted to evaluating the thermal-hydraulic characteristics of a full-length light water reactor (LWR) fuel bundle during the heatup, reflood, and quench phases of a LOCA. Experimental results from 28 tests cover reflood rates of 0.74 in./sec to 11 in./sec and delay times to initiate reflood of 3 sec to 66 sec. The results indicate that current analysis methods can predict peak temperatures within 10% and measured quench times for the bundle were significantly less than predicted. For reflood rates of 1 in./sec where long quench times were predicted (>2000 sec}, measured quench times of 200 sec were found.

  16. Fuel rod mechanical deformation during the PBF/LOFT lead rod loss-of-coolant experiments

    SciTech Connect

    Varacalle, Jr., D. J.; MacDonald, P. E.; Shiozawa, S.; Driskell, W. E.

    1980-01-01

    Results of four PBF/LOFT Lead Rod (LLR) sequential blowdown tests conducted in the Power Burst Facility (PBF) are presented. Each test employed four separately shrouded fuel rods. The primary objective of the test series was to evaluate the extent of mechanical deformation that would be expected to occur to low pressure (0.1 MPa), light water reactor design fuel rods when subjected to a series of double ended cold leg break loss-of-coolant accident (LOCA) tests, and to determine whether subjecting these deformed fuel rods to subsequent testing would result in rod failure. The extent of mechanical deformation (buckling, collapse, or waisting of the cladding) was evaluated by comparison of cladding temperature and system pressure measurements with out-of-pile experimental data, and by posttest visual examinations and cladding diametral measurements.

  17. A general model for predicting coolant activity behaviour for fuel-failure monitoring analysis

    NASA Astrophysics Data System (ADS)

    El-Jaby, A.; Lewis, B. J.; Thompson, W. T.; Iglesias, F.; Ip, M.

    2010-04-01

    A mathematical treatment has been developed to predict the release of volatile fission products from operating defective nuclear fuel elements. The fission product activity in both the fuel-to-sheath gap and primary heat transport system as a function of time can be predicted during all reactor operating conditions, including: startup, steady-state, shutdown, and bundle-shifting manoeuvres. In addition, an improved ability to predict the coolant activity of the 135Xe isotope in commercial reactors is discussed. A method is also proposed to estimate both the burnup and the amount of tramp uranium deposits in-core. The model has been validated against in-reactor experiments conducted with defective fuel elements containing natural and artificial failures at the Chalk River Laboratories. Lastly, the model has been benchmarked against a defective fuel occurrence in a commercial reactor.

  18. Thermohydraulic responses of a water-cooled tokamak fusion DEMO to loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Nakamura, M.; Tobita, K.; Someya, Y.; Utoh, H.; Sakamoto, Y.; Gulden, W.

    2015-11-01

    Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. As for the in-VV LOCA, we analysed the multiple double-ended break of the first wall cooling pipes around the outboard toroidal circumference. As for the ex-VV LOCA, we analysed the double-ended break of the primary cooling pipe. The thermohydraulic analysis results suggest that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. Mitigations of the loads to the confinement barriers are also discussed.

  19. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    SciTech Connect

    Marshall, William BJ J

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  20. Piston slap induced pressure fluctuation in the water coolant passage of an internal combustion engine

    NASA Astrophysics Data System (ADS)

    Ohta, Kazuhide; Wang, Xiaoyu; Saeki, Atsushi

    2016-02-01

    Liner cavitation is caused by water pressure fluctuation in the water coolant passage (WCP). When the negative pressure falls below the saturated vapor pressure, the impulsive pressure following the implosion of cavitation bubbles causes cavitation erosion of the wet cylinder liner surface. The present work establishes a numerical model for structural-acoustic coupling between the crankcase and the acoustic field in the WCP considering their dynamic characteristics. The coupling effect is evaluated through mutual interaction terms that are calculated from the mode shapes of the acoustic field and of the crankcase vibration on the boundary. Water pressure fluctuations in the WCP under the action of piston slap forces are predicted and the contributions of the uncoupled mode shapes of the crankcase and the acoustic field to the pressure waveform are analyzed. The influence of sound speed variations on the water pressure response is discussed, as well as the pressure on the thrust sides of the four cylinders.

  1. Liquid Cooling of Tractive Lithium Ion Batteries Pack with Nanofluids Coolant.

    PubMed

    Li, Yang; Xie, Huaqing; Yu, Wei; Li, Jing

    2015-04-01

    The heat generated from tractive lithium ion batteries during discharge-charge process has great impacts on the performances of tractive lithium ion batteries pack. How to solve the thermal abuse in tractive lithium ion batteries pack becomes more and more urgent and important for future development of electrical vehicles. In this work, TiO2, ZnO and diamond nanofluids are prepared and utilized as coolants in indirect liquid cooling of tractive lithium ion batteries pack. The results show that nanofluids present superior cooling performance to that of pure fluids and the diamond nanofluid presents relatively excellent cooling abilities than that of TiO2 and ZnO nanofluids. During discharge process, the temperature distribution of batteries in batteries pack is uniform and stable, due to steady heat dissipation by indirect liquid cooling. It is expected that nanofluids could be considered as a potential alternative for indirect liquid cooling in electrical vehicles. PMID:26353564

  2. WWER Expert System for Fuel Failure Analysis Using Data on Primary Coolant Activity

    SciTech Connect

    Likhanskii, V.V.; Evdokimov, I.A.; Sorokin, A.A.; Khromov, A.G.; Kanukova, V.D.; Apollonova, O.V.; Ugryumov, A.V.

    2007-07-01

    The computer expert system for fuel failure analysis of WWER during operation is presented. The diagnostics is based on the measurement of specific activity of reference nuclides in reactor primary coolant and application of a computer code for the data interpretation. The data analysis includes an evaluation of tramp uranium mass in reactor core, detection of failures by iodine and caesium spikes, evaluation of burnup of defective fuel. Evaluation of defective fuel burnup was carried out by applying the relation of caesium nuclides activity in spikes and relations of activities of gaseous fission products for steady state operational conditions. The method of burnup evaluation of defective fuel by use of fission gas activity is presented in detail. The neural-network analysis is performed for determination of failed fuel rod number and defect size. Results of the expert system application are illustrated for several fuel campaigns on operating WWER NPPs. (authors)

  3. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

  4. Sensitivity studies of loss-of-coolant accidents in the Savannah River production reactors

    SciTech Connect

    Edwards, J.N.; Motley, F.E.; Morgan, M.M.; Knight, T.D.; Fischer, S.R. )

    1990-01-01

    Loss-of-coolant accident (LOCA) analyses were completed using the Transient Reactor Analysis Code (TRAC) to support the U.S. Department of Energy efforts to restart the production reactors located at the Savannah River Site. The break location and pump operation after the LOCA were the parameters varied for these sensitivity studies. Three location of double-ended guillotine break were studied: plenum inlet, pump suction, and pump discharge. Three pump operation scenarios were also studied: continued operation of both ac and dc pumps, tripping of the ac motor at 2 s after the LOCA, and tripping of the ac motor at 200 s after the LOCA. The production reactors use low pressure and temperature heavy water as the process fluid. The reactor has a moderator tank that contains the fuel channels. Above the moderator tank is an upper plenum that distributes the heavy water to each fuel assembly. The heavy water flows down through the fuel channels and into the moderator tank. From the tank, the water is pumped back to the upper plenum through six loops. Each loop contains a pump and two heat exchangers. Four of the loops have an emergency core coolant system (ECCS) connection. This TRAC model has been benchmarked extensively against data taken in the actual reactors or in prototypical models of the components of the reactors. The calculations were completed using a version of TRAC-PF1/MOD 2 that was updated to include heavy water properties and other changes that are specific to the production reactors.

  5. Surface Treatment to Improve Corrosion Resistance in Lead-Alloy Coolants

    SciTech Connect

    Todd R. Allen; Kumar Sridharan; McLean T. Machut; Lizhen Tan

    2007-08-29

    One of the six proposed advanced reactor designs of the Generation IV Initiative, the Leadcooled Fast Reactor (LFR) possesses many characteristics that make it a desirable candidate for future nuclear energy production and responsible actinide management. These characteristics include favorable heat transfer, fluid dynamics, and neutronic performance compared to other candidate coolants. However, the use of a heavy liquid metal coolant presents a challenge for reactor designers in regards to reliable structural and fuel cladding materials in both a highly corrosive high temperature liquid metal and an intense radiation fieldi. Flow corrosion studies at the University of Wisconsin have examined the corrosion performance of candidate materials for application in the LFR concept as well as the viability of various surface treatments to improve the materials’ compatibility. To date this research has included several focus areas, which include the formulation of an understanding of corrosion mechanisms and the examination of the effects of chemical and mechanical surface modifications on the materials’ performance in liquid lead-bismuth by experimental testing in Los Alamos National Laboratory’s DELTA Loop, as well as comparison of experimental findings to numerical and physical models for long term corrosion prediction. This report will first review the literature and introduce the experiments and data that will be used to benchmark theoretical calculations. The experimental results will be followed by a brief review of the underlying theory and methodology for the physical and theoretical models. Finally, the results of theoretical calculations as well as experimentally obtained benchmarks and comparisons to the literature are presented.

  6. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

    SciTech Connect

    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  7. A hydrogen-oxygen rocket engine coolant passage design program (RECOP) for fluid-cooled thrust chambers and nozzles

    NASA Technical Reports Server (NTRS)

    Tomsik, Thomas M.

    1994-01-01

    The design of coolant passages in regeneratively cooled thrust chambers is critical to the operation and safety of a rocket engine system. Designing a coolant passage is a complex thermal and hydraulic problem requiring an accurate understanding of the heat transfer between the combustion gas and the coolant. Every major rocket engine company has invested in the development of thrust chamber computer design and analysis tools; two examples are Rocketdyne's REGEN code and Aerojet's ELES program. In an effort to augment current design capabilities for government and industry, the NASA Lewis Research Center is developing a computer model to design coolant passages for advanced regeneratively cooled thrust chambers. The RECOP code incorporates state-of-the-art correlations, numerical techniques and design methods, certainly minimum requirements for generating optimum designs of future space chemical engines. A preliminary version of the RECOP model was recently completed and code validation work is in progress. This paper introduces major features of RECOP and compares the analysis to design points for the first test case engine; the Pratt & Whitney RL10A-3-3A thrust chamber.

  8. Stochastic simulation of fission product activity in primary coolant due to fuel rod failures in typical PWRs under power transients

    NASA Astrophysics Data System (ADS)

    Javed Iqbal, M.; Mirza, Nasir M.; Mirza, Sikander M.

    2008-01-01

    During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the primary coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in primary coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the 131I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent 135I, 135Xe and 89Kr concentrations in primary coolant measured during EDITHMOX-1 experiments.

  9. TACT 1: A computer program for the transient thermal analysis of a cooled turbine blade or vane equipped with a coolant insert. 2. Programmers manual

    NASA Technical Reports Server (NTRS)

    Gaugler, R. E.

    1979-01-01

    A computer program to calculate transient and steady state temperatures, pressures, and coolant flows in a cooled axial flow turbine blade or vane with an impingement insert is described. Coolant-side heat transfer coefficients are calculated internally in the program, with the user specifying either impingement or convection heat transfer at each internal flow station. Spent impingement air flows in a chordwise direction and is discharged through the trailing edge and through film cooling holes. The ability of the program to handle film cooling is limited by the internal flow model. Input to the program includes a description of the blade geometry, coolant-supply conditions, outside thermal boundary conditions, and wheel speed. The blade wall can have two layers of different materials, such as a ceramic thermal barrier coating over a metallic substrate. Program output includes the temperature at each node, the coolant pressures and flow rates, and the coolant-side heat transfer coefficients.

  10. Comparative Evaluation of Two Different Ultrasonic Liquid Coolants on Dental Aerosols

    PubMed Central

    Bhandari, Vishnudas; Ugale, Gauri; Taru, Snehal; Khaparde, Surbhi; Kulkarni, Arun; Ardale, Mukesh; Marde, Shraddha

    2016-01-01

    Introduction Dentists are more prone for developing infectious diseases especially related to respiratory system. The ultrasonic scaler which is a major source of dental aerosol production is most frequently used contrivance in a dental set up. Aim The aim of this study was to evaluate the effect of povidone iodine and chlorhexidine gluconate as an ultrasonic liquid coolant on aerosols in comparison with distilled water. The objectives of this study were to compare the potency of povidone iodine and chlorhexidine gluconate on reducing dental aerosols and quantitative assessment of microbial content of dental aerosols at right, left and behind the dental chair. Materials and Methods In this study 30 subjects were selected who fulfilled the inclusion criteria and were divided into three groups. Group 1 (Control group): Ultrasonic scaling with distilled water (10 subjects), Group 2 (Test group): Ultrasonic scaling with 2% povidone iodine (10 subjects), Group 3 (Test group): Ultrasonic scaling with 0.12% chlorhexidine (10 subjects). At the baseline one blood agar plate was kept for 10 minutes in the fumigated chamber before ultrasonic scaling, thereafter three blood agar plates were kept at a distance of 0.4 meters away on either side of the patient and 2 meters behind the patient’s mouth during ultrasonic scaling. Blood agar plates were kept for gravitometric settling of dental aerosols. Results At baseline, no significant numbers of Colony-Forming Units (CFU) were detected. It is found that Group 3 (chlorhexidine gluconate) showed effective CFU reduction (27.17 ±12.5 CFU) when compared to distilled water (124.5 ± 30.08 CFU) and povidone iodine (60.43 ± 33.33 CFU). More CFU were found on blood agar plates which were kept on right side in all the three groups. The results obtained were statistically significant (p< 0.001). Conclusion Chlorhexidine gluconate is more effective in reducing dental aerosols when compared to povidone iodine and distilled water. Povidone

  11. Correlation of cylinder-head temperatures and coolant heat rejections of a multicylinder, liquid-cooled engine of 1710-cubic-inch displacement

    NASA Technical Reports Server (NTRS)

    Lundin, Bruce T; Povolny, John H; Chelko, Louis J

    1949-01-01

    Data obtained from an extensive investigation of the cooling characteristics of four multicylinder, liquid-cooled engines have been analyzed and a correlation of both the cylinder-head temperatures and the coolant heat rejections with the primary engine and coolant variables was obtained. The method of correlation was previously developed by the NACA from an analysis of the cooling processes involved in a liquid-cooled-engine cylinder and is based on the theory of nonboiling, forced-convection heat transfer. The data correlated included engine power outputs from 275 to 1860 brake horsepower; coolant flows from 50 to 320 gallons per minute; coolants varying in composition from 100 percent water to 97 percent ethylene glycol and 3 percent water; and ranges of engine speed, manifold pressure, carburetor-air temperature, fuel-air ratio, exhaust-gas pressure, ignition timing, and coolant temperature. The effect on engine cooling of scale formation on the coolant passages of the engine and of boiling of the coolant under various operating conditions is also discussed.

  12. Characterization of suspended particles in Three Mile Island Unit 2 reactor coolant water

    SciTech Connect

    Akamine, K. . Hitachi Works); Hofstetter, K.J. and Co., Aiken, SC . Savannah River Lab.); Boston, V.F. )

    1989-02-01

    On commencing defueling operations in the Three Mile Island Unit 2 (TMI-2) reactor vessel damaged core region, the defueling water cleanup system (DWCS) encountered rapid plugging of its filter media. Characterization of the suspended material was an important task in resolving DWCS filtration difficulties. The characterization of the suspended material involved laboratory analyses of reactor vessel coolant samples collected from May through November 1986. The results of these characterizations indicated that the major elements present in the suspended particles were silver, aluminum, cadmium, iron, indium, silicon, uranium, and zirconium, all of which correspond to the five known source terms in the TMI-2 reactor vessel control rod alloy, zeolite, diatomaceous earth, steel, fuel, and Zircaloy cladding. The particle analysis data indicate that the majority of the particles were <5 {mu}m and many of these suspended particles existed as colloidal particles; hence, these particulates are believed to have been the principal basis for filter plugging. In addition, based on these characterization data and data from previous analyses of reactor components, it was postulated that some mass fraction of the liquefied control rod alloy formed aerosols from mechanical formation due to high-velocity gas interaction with the moving liquid alloy.

  13. Loads on steam generator tubes during simulated loss-of-coolant accident conditions. Final report. [PWR

    SciTech Connect

    Guerrero, H.N.; Hiestand, J.W.; Rossano, F.V.; Shah, P.K.; Thakkar, J.G.

    1982-11-01

    This report presents the work performed to verify the CEFLASH digital computer code modeling of the hydro-dynamic loads in a steam generator tube during a loss-of-coolant accident (LOCA). The test loop simulated the primary side thermal-hydraulic conditions in an operational nuclear steam generator. The loop consisted of 5 full size double 90/sup 0/ bend tubes and steam generator plena, a pressurizer, a reactor resistance simulator, a heater, a pump, and associated pipes and valves to complete the system. The tubes used were of typical length and the same outside diameter as those used in C-E steam generators. Prototypical supports were provided for the bundle of 5 tubes. Cold leg guillotine breaks were simulated using quick opening valve and rupture disks. Break opening times ranged from less than 1 msec to as much as 67 milliseconds. The loop instrumentation was designed to measure the transient pressure history at various locations and monitor the structural response of the tube to the LOCA hydrodynamic loading. A series of blowdown tests was performed for different operating and boundary conditions. Analytically predicted transient pressure histories and the differential pressure history across the tube span were compared with the experimental data.

  14. Analysis of an AP600 intermediate-size loss-of-coolant accident

    SciTech Connect

    Boyack, B.E.; Lime, J.F.

    1995-04-01

    A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations performed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

  15. Long-term aging and loss-of-coolant accident (LOCA) testing of electrical cables

    SciTech Connect

    Nelson, C.F.; Gauthier, G.; Carlin, F.

    1996-10-01

    Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France: (1) U.S. 2 conductor with ethylene propylene rubber (EPR) insulation and a Hypalon jacket. (2) U.S. 3 conductor with cross-linked polyethylene (XLPE) insulation and a Hypalon jacket. (3) French 3 conductor with EPR insulation and a Hypalon jacket. (4) French coaxial with polyethylene (PE) insulation and a PE jacket. The data represent up to 5 years of simultaneous aging where the cables were exposed to identical aging radiation doses at either 40{degrees}C or 70{degrees}C; however, the dose rate used for the aging irradiation was varied over a wide range (2-100 Gy/hr). Aging was followed by exposure to simulated French LOCA conditions. Several mechanical, electrical, and physical-chemical condition monitoring techniques were used to investigate the degradation behavior of the cables. All the cables, except for the French PE cable, performed acceptably during the aging and LOCA simulations. In general, cable degradation at a given dose was highest for the lowest dose rate, and the amount of degradation decreased as the dose rate was increased.

  16. Uniform corrosion of FeCrAl alloys in LWR coolant environments

    NASA Astrophysics Data System (ADS)

    Terrani, K. A.; Pint, B. A.; Kim, Y.-J.; Unocic, K. A.; Yang, Y.; Silva, C. M.; Meyer, H. M.; Rebak, R. B.

    2016-10-01

    The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. The maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ∼2 μm, which is inconsequential for a ∼300-500 μm thick cladding.

  17. Large break loss of coolant severe accident sequences at the HFIR (High Flux Isotope Reactor)

    SciTech Connect

    Simpson, D.B.; Greene, S.R.

    1990-01-01

    An assessment of many potential HFIR severe accident phenomena was conducted during the HFIR design effort, and many severe accident mitigating features were designed into the plant. These evaluation typically incorporated a bounding'' or highly conservative analysis approach and employed tools and techniques representative of the state of knowledge in the mid-1960s. Recently, programs to address severe accident issues were initiated at the Oak Ridge National Laboratory (ORNL) to support the HFIR probabilistic risk assessment (PRA) and equipment qualification and accident management studies. This paper presents the results of environment condition calculations conducted to evaluate a response of HFIR's heat exchanger cell environment to a double-ended rupture of a 0.25 m diameter coolant loop downstream of the circulating pump and check valve. The confinement calculations were performed using an atmospheric fission product source for the heat exchanger cell consistent with, but more conservative than that stipulated in Regulatory Guide 1.89. The results of the calculations indicate that the heat exchanger cell atmospheric temperature peaks at 377 K 225 seconds into the transient and then begins decreasing at approximately 1.7 K per minute. 8 refs., 5 figs.

  18. Aging and loss-of-coolant accident (LOCA) testing of electrical connections

    SciTech Connect

    Nelson, C.F.

    1998-01-01

    This report presents the results of an experimental program to determine the aging and loss-of-coolant accident (LOCA) behavior of electrical connections in order to obtain an initial scoping of their performance. Ten types of connections commonly used in nuclear power plants were tested. These included 3 types of conduit seals, 2 types of cable-to-device connectors, 3 types of cable-to-cable connectors, and 2 types of in-line splices. The connections were aged for 6 months under simultaneous thermal (99 C) and radiation (46 Gy/hr) conditions. A simulated LOCA consisting of sequential high dose-rate irradiation (3 kGy/hr) and high-temperature steam exposures followed the aging. Connection functionality was monitored using insulation resistance measurements during the aging and LOCA exposures. Because only 5 of the 10 connection types passed a post-LOCA, submerged dielectric withstand test, further detailed investigation of electrical connections and the effects of cable jacket integrity on the cable-connection system is warranted.

  19. Uniform corrosion of FeCrAl alloys in LWR coolant environments

    DOE PAGESBeta

    Terrani, K. A.; Pint, B. A.; Kim, Y. -J.; Unocic, K. A.; Yang, Y.; Silva, C. M.; Meyer, III, H. M.; Rebak, R. B.

    2016-06-29

    In this study, the corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted inmore » the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. Finally, the maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ~2 μm, which is inconsequential for a ~300–500 μm thick cladding.« less

  20. Analysis of fission product revaporization in a BWR Reactor Coolant System during a station blackout accident

    SciTech Connect

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

    1988-01-01

    This paper presents an analysis of fission product revaporization from the Reactor Coolant System (RCS) following the Reactor Pressure Vessel (RPV) failure. The station blackout accident in a BWR Mark I Power Plant was considered. The TRAPMELT3 models for vaporization, chemisorption, and the decay heating of RCS structures and gases were used and extended beyond the RPV failure in the analysis. The RCS flow models based on the density-difference or pressure-difference between the RCS and containment pedestal region were developed to estimate the RCS outflow which carries the revaporized fission product to the containment. A computer code called REVAP was developed for the analysis. The REVAP code was incorporated with the MARCH, TRAPMELT3 and NAUA codes from the Source Term Code Package (STCP) to estimate the impact of revaporization on environmental release. The results show that the thermal-hydraulic conditions between the RCS and the pedestal region are important factors in determining the magnitude of revaporization and subsequent release of the volatile fission product into the environment. 6 refs., 8 figs.

  1. The effect of a disinfectant/coolant irrigant on microbes isolated from dental unit water lines.

    PubMed

    Epstein, Joel B; Dawson, J R; Buivids, Ilze A; Wong, Bea; Le, Nhu D

    2002-01-01

    The purpose of this study was to assess water samples from a hospital dental clinic to determine whether a disinfectant/coolant irrigant containing chlorhexidine (Lines, Micrylium Laboratories) affects the presence of microbial organisms in dental unit waterlines. Water samples from three hospital dental operatories were collected at baseline and after overnight treatment with a disinfectant-containing irrigant followed by sterile water irrigation. Saliva of treated patients and sterile water rinse specimens were collected from the waterlines of these operatories for three consecutive days, then weekly for eight weeks after treatment. Specimens were cultured to identify total heterotrophic plate counts as well as presence of Pseudomonas aeruginosa and Candida species. Baseline organism counts varied from 10(3) to 10(5) colony-forming units per milliliter. After treatment, no organisms were detected in waterline discharge. Decontamination of dental unit waterlines is possible using a disinfectant/irrigant followed by sterile water irrigation. The potential for contamination of the lines from patients' saliva may have been reduced due to use of anti-retraction valves and the disinfectant/sterile water irrigation, as conducted in this study. PMID:12449456

  2. Aging, Loss-of-Coolant Accident (LOCA), and high potential testing of damaged cables

    SciTech Connect

    Vigil, R.A.; Jacobus, M.J.

    1994-04-01

    Experiments were conducted to assess the effects of high potential testing of cables and to assess the survivability of aged and damaged cables under Loss-of-Coolant Accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undamaged cables suggested that no damage was incurred on the selected virgin cables. During aging and LOCA testing, Okonite ethylene propylene rubber (EPR) cables with a bonded jacket experienced unexpected failures. The failures appear to be primarily related to the level of thermal aging and the presence of a bonded jacket that ages more rapidly than the insulation. For Brand Rex crosslinked polyolefin (XLPO) cables, the results suggest that 7 mils of insulation remaining should give the cables a high probability of surviving accident exposure following aging. The voltage necessary to detect when 7 mils of insulation remain on unaged Brand Rex cables is approximately 35 kVdc. This voltage level would almost certainly be unacceptable to a utility for use as a damage assessment tool. However, additional tests indicated that a 35 kvdc voltage application would not damage virgin Brand Rex cables when tested in water. Although two damaged Rockbestos silicone rubber cables also failed during the accident test, no correlation between failures and level of damage was apparent.

  3. Analysis of an AP600 intermediate-size loss-of-coolant accident

    SciTech Connect

    Boyack, B.E.; Lime, J.F.

    1995-09-01

    A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

  4. Thermodynamic considerations and prediction of the primary coolant activity of 99Tc

    NASA Astrophysics Data System (ADS)

    Lewis, B. J.; Thompson, W. T.; Akbari, F.; Morrison, C.; Husain, A.

    2005-04-01

    A physical model has been developed to describe the coolant activity behaviour of 99Tc, during constant and reactor shutdown operations. This analysis accounts for the fission production of technetium and molybdenum, in which their chemical form and volatility is determined by a thermodynamic treatment using Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix and vaporization from the fuel-grain surface. Based on several in-reactor tests with defective fuel elements, and as supported by the thermodynamic analysis, the model accounts for the washout of molybdenum from the defective fuel on reactor shutdown. The model also considers the recoil release of both 99Mo and 99Tc from uranium contamination, as well as a corrosion source due to activation of 98Mo. The model has provided an estimate of the activity ratio 99Tc/ 137Cs in the ion-exchange columns of the Darlington Nuclear Generating Station, i.e., 6 × 10 -6 (following ˜200 days of steady reactor operation) and 4 × 10 -6 (with reactor shutdown). These results are consistent with that measured by the Battelle Pacific Northwest Laboratories with a mixed-bed resin-sampling device installed in a number of Pressurized Water Reactor and Boiling Water Reactor plants.

  5. Review of Failure Probability Calculations for HFIR Primary Coolant System Piping

    SciTech Connect

    Simonen, Fredric A.

    2001-10-31

    During July 2001, Pacific Northwest National Laboratory was requested by the U.S. Department of Energy, Office of Nuclear Facilities Management, Office of Nuclear Energy, Science and Technology, Germantown, Maryland, to review calculations of piping failure probabilities for the High Flux Test Reactor (HFIR) located at and operated by the Oak Ridge National Laboratory (ORNL). The objective of the failure probability calculations was to estimate the probabilities of large leaks (>1500 gpm) that are of sufficient size to disable the primary coolant system of HFIR to the extent that there is a potential for core damage. PNNL reviewed the computational methods and the inputs to the calculations along with an evaluation of potential failure mechanisms not explicitly addressed by the ORNL calculations. The review concluded that the calculated failure probabilities even with consideration of uncertainties in the calculations and of other potential failure mechanisms provide a high level of confidence that failure frequencies are less than the stated goal of 10-6 piping failures per year.

  6. An analysis method for multistage transonic turbines with coolant mass flow addition

    SciTech Connect

    Mildner, F.; Gallus, H.E.

    1998-10-01

    The subject of this paper is a numerical method for the calculation of the transonic flow field of multistage turbines, taking high coolant flow into account. To reduce the processing time, a throughflow method based on the principles of Wu is used for the hub-to-tip calculation. The flow field is obtained by an iterative solution between a three-dimensional inviscid hyperbolic time-dependent algorithm with an implicit finite volume method for the blade-to-blade calculations using C-meshes and a single representative meridional S{sub 2m}-stream surface. Along the 2{sub 2m}-plane with respect to nonorthogonal curvilinear coordinates, the stream function equation governing fluid flow is established. The cooling air inflow inside the blade passage forbids the assumption of a constant mass flow along the main stream direction. To consider the change of the aerodynamic and thermodynamic behavior, a cooling air model was developed and implemented in the algorithm, which allows the mixing of radially arbitrarily distributed cooling air in the trailing edge section of each blade row. The viscous effects and the influence of cooling air mixing are considered by the use of selected loss correlations for profile, tip leakage, secondary flow and mixing losses in the S{sub 2m}-plane in terms of entropy. The method is applied to the four-stage high-temperature gas turbine Siemens KWU V84.3. The numerical results obtained are in good agreement with the experimental data.

  7. Interaction study between MOX fuel and eutectic lead-bismuth coolant

    NASA Astrophysics Data System (ADS)

    Vigier, Jean-François; Popa, Karin; Tyrpekl, Vaclav; Gardeur, Sébastien; Freis, Daniel; Somers, Joseph

    2015-12-01

    In the frame of the MYRRHA reactor project, the interaction between fuel pellets and the reactor coolant is essential for safety evaluations, e.g. in case of a pin breach. Therefore, interaction tests between uranium-plutonium mixed oxide (MOX) pellets and molten lead bismuth eutectic (LBE) have been performed and three parameters were studied, namely the interaction temperature (500 °C and 800 °C), the oxygen content in LBE and the stoichiometry of the MOX (U0.7Pu0.3O2-x and U0.7Pu0.3O2.00). After 50 h of interaction in closed containers, the pellet integrity was preserved in all cases. Whatever the conditions, neither interaction compounds (crystalline or amorphous) nor lead and bismuth diffusion into the surface regions of the MOX pellets has been detected. In most of the conditions, actinide releases into LBE were very limited (in the range of 0.01-0.15 mg), with a homogeneous release of the different actinides present in the MOX. Detected values were significantly higher in the 800 °C and low LBE oxygen content tests for both U0.7Pu0.3O2-x and U0.7Pu0.3O2.00, with 1-2 mg of actinide released in these conditions.

  8. Simulating experimental investigation on the safety of nuclear heating reactor in loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Xu, Zhanjie

    1996-12-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology (INET) of Tsinghua University of China in 1989. Its main loop is a thermal-hydraulic system with natural circulation. This paper studies the safety of NHR under the condition of loss-of-coolant accidents (LOCAs) by means of simulant experiments. First, the background and necessity of the experiments are presented, then the experimental system, including the thermal-hydraulic system and the data collection system, and similarity criteria are introduced. Up to now, the discharge experiments with the residual heating power (20% rated heating power) have been carried out on the experimental system. The system parameters including circulation flow rate, system pressure, system temperature, void fraction, discharge mass and so on have been recorded and analyzed. Based on the results of the experiments, the conclusions are shown as follos: on the whole, the reactor is safe under the condition of LOCAs, but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  9. Correlation of analysis with high level vibration test results for primary coolant piping

    SciTech Connect

    Park, Y.J.; Hofmayer, C.H. ); Costello, J.F. )

    1992-01-01

    Dynamic tests on a modified 1/2.5-scale model of pressurized water reactor (PWR) primary coolant piping were performed using a large shaking table at Tadotsu, Japan. The High Level Vibration Test (HLVT) program was part of a cooperative study between the United States (Nuclear Regulatory Commission/Brookhaven National Laboratory, NRC/BNL) and Japan (Ministry of International Trade and Industry/Nuclear Power Engineering Center). During the test program, the excitation level of each test run was gradually increased up to the limit of the shaking table and significant plastic strains, as well as cracking, were induced in the piping. To fully utilize the test results, NRC/BNL sponsored a project to develop corresponding analytical predictions for the nonlinear dynamic response of the piping for selected test runs. The analyses were performed using both simplified and detailed approaches. The simplified approaches utilize a linear solution and an approximate formulation for nonlinear dynamic effects such as the use of a deamplification factor. The detailed analyses were performed using available nonlinear finite element computer codes, including the MARC, ABAQUS, ADINA and WECAN codes. A comparison of various analysis techniques with the test results shows a higher prediction error in the detailed strain values in the overall response values. A summary of the correlation analyses was presented before the BNL. This paper presents a detailed description of the various analysis results and additional comparisons with test results.

  10. Correlation of analysis with high level vibration test results for primary coolant piping

    SciTech Connect

    Park, Y.J.; Hofmayer, C.H.; Costello, J.F.

    1992-05-01

    Dynamic tests on a modified 1/2.5-scale model of pressurized water reactor (PWR) primary coolant piping were performed using a large shaking table at Tadotsu, Japan. The High Level Vibration Test (HLVT) program was part of a cooperative study between the United States (Nuclear Regulatory Commission/Brookhaven National Laboratory, NRC/BNL) and Japan (Ministry of International Trade and Industry/Nuclear Power Engineering Center). During the test program, the excitation level of each test run was gradually increased up to the limit of the shaking table and significant plastic strains, as well as cracking, were induced in the piping. To fully utilize the test results, NRC/BNL sponsored a project to develop corresponding analytical predictions for the nonlinear dynamic response of the piping for selected test runs. The analyses were performed using both simplified and detailed approaches. The simplified approaches utilize a linear solution and an approximate formulation for nonlinear dynamic effects such as the use of a deamplification factor. The detailed analyses were performed using available nonlinear finite element computer codes, including the MARC, ABAQUS, ADINA and WECAN codes. A comparison of various analysis techniques with the test results shows a higher prediction error in the detailed strain values in the overall response values. A summary of the correlation analyses was presented before the BNL. This paper presents a detailed description of the various analysis results and additional comparisons with test results.

  11. Effects of LWR coolant environments on fatigue lives of austenitic stainless steels

    SciTech Connect

    Chopra, O.K.; Gavenda, D.J.

    1997-07-01

    The ASME Boiler and Pressure Vessel Code fatigue design curves for structural materials do not explicitly address the effects of reactor coolant environments on fatigue life. Recent test data indicate a significant decrease in fatigue life of pressure vessel and piping materials in light water reactor (LWR) environments. Fatigue tests have been conducted on Types 304 and 316NG stainless steel in air and LWR environments to evaluate the effects of various material and loading variables, e.g., steel type, strain rate, dissolved oxygen (DO) in water, and strain range, on fatigue lives of these steels. The results confirm the significant decrease in fatigue life in water. The environmentally assisted decrease in fatigue life depends both on strain rate and DO content in water. A decrease in strain rate from 0.4 to 0.004%/s decreases fatigue life by a factor of {approx} 8. However, unlike carbon and low-alloy steels, environmental effects are more pronounced in low-DO than in high-DO water. At {approx} 0.004%/s strain rate, reduction in fatigue life in water containing <10 ppb D is greater by a factor of {approx} 2 than in water containing {ge} 200 ppb DO. Experimental results have been compared with estimates of fatigue life based on the statistical model. The formation and growth of fatigue cracks in austenitic stainless steels in air and LWR environments are discussed.

  12. FEM Analysis and Experimental Verification of the Integral Forging Process for AP1000 Primary Coolant Pipe

    NASA Astrophysics Data System (ADS)

    Wang, Shenglong; Yu, Xiaoyi; Yang, Bin; Zhang, Mingxian; Wu, Huanchun

    2016-08-01

    AP1000 primary coolant pipes must be manufactured by integral forging technology according to the designer—Westinghouse Electric Co. The characteristics of these large, special-shaped pipes create nonuniform temperatures, effective stress, and effective strain during shaping of the pipes. This paper presents a three-dimensional finite element simulation (3D FEM) of the integral forging process, and qualitatively evaluates the likelihood of forging defects. By analyzing the evolution histories of the three field variables, we concluded that the initial forging temperature should be strictly controlled within the interval 1123 K to 1423 K (850 °C to 1150 °C) to avoid second-phase precipitation. In the hard deformation zones, small strains do not contribute to recrystallization resulting in coarse grains. Conversely, in the free deformation zone, the large strains can contribute to the dynamic recrystallization, favoring grain refinement and closure of voids. Cracks are likely to appear, however, on the workpiece surface when forging leads to large deformations. Based on the simulation results, an eligible workpiece with good mechanical properties, few macroscopic defects, and favorable grain size has been successfully forged by experiments at an industrial scale, which validates the FEM simulation.

  13. Analysis of Flow in Pilot Operated Safety and Relief Valve of Nuclear Reactor Coolant System

    SciTech Connect

    Kwon, Soon-Bum; Lee, Dong-Won; Kim, In-Goo; Ahn, Hyung-Joon; Kim, Hho-Jung

    2004-07-01

    When the POSRV equipped in a nuclear power plant opens in instant by a failure in coolant system of PWR, a moving shock wave generates, and propagates downstream of the valve, inducing a complicated unsteadiness. The moving shock wave may exert severe load to the structure. In this connection, a method of gradual opening of the valve is used to reduce the load acting on the wall at the downstream of the POSRV. In the present study, experiments and calculations are performed to investigate the detail unsteady flow at the various pipe units and the effect of valve opening time on the flow downstream of the valve. In calculation by using of air as working fluid, 2-dimensional, unsteady compressible Navier-Stokes equations are solved by finite volume method. It was found that when the incident shock wave passes through the pipe unit, it may experience diffraction, reflection and interaction with a vortex. Furthermore, the geometry of the pipe unit affects the reflection type of shock wave and changes the load acting on the wall of pipe unit. It was also turned out that the maximum force acting on the wall of the pipe unit becomes in order of T-junction, 108 deg. elbow and branch in magnitude, respectively. And, the results obtained that show that the rapid pressure rise due to the moving shock wave by instant POSRV valve opening is attenuated by employing the gradual opening. (authors)

  14. The solution-adaptive numerical simulation of the 3D viscous flow in the serpentine coolant passage of a radial inflow turbine blade

    NASA Astrophysics Data System (ADS)

    Dawes, W. N.

    1992-06-01

    This paper describes the application of a solution-adaptive, three-dimensional Navier-Stokes solver to the problem of the flow in turbine internal coolant passages. First the variation of Nusselt number in a cylindrical, multi-ribbed duct is predicted and found to be in acceptable agreement with experimental data. Then the flow is computed in the serpentine coolant passage of a radial inflow turbine including modeling the internal baffles and pin fins. The aerodynamics of the passage, particularly that associated with the pin fins, is found to be complex. The predicted heat transfer coefficients allow zones of poor coolant penetration and potential hot spots to be identified.

  15. The cryogenic on-orbit liquid analytical tool (COOLANT) - A computer program for evaluating the thermodynamic performance of orbital cryogen storage facilities

    NASA Technical Reports Server (NTRS)

    Taylor, W. J.; Honkonen, S. C.; Williams, G. E.; Liggett, M. W.; Tucker, S. P.

    1991-01-01

    The United States plans to establish a permanent manned presence at the Space Station Freedom in low earth orbit (LEO) and then carry out exploration of the solar system from this base. These plans may require orbital cryogenic propellant storage depots. The COOLANT program has been developed to analyze the thermodynamic performance of these depots to support design tradeoff studies. It was developed as part of the Long Term Cryogenic Storage Facility Systems Study for NASA/MSFC. This paper discusses the program structure and capabilities of the COOLANT program. In addition, the results of an analysis of a 200,000 lbm hydrogen/oxygen storage depot tankset using COOLANT are presented.

  16. Analysis of metal temperature and coolant flow with a thermal-barrier coating on a full-coverage-film-cooled turbine vane

    NASA Technical Reports Server (NTRS)

    Meitner, P. L.

    1978-01-01

    The potential benefits of combining full-coverage film cooling with a thermal-barrier coating were investigated analytically for sections on the suction and pressure sides a high-temperature, high-pressure turbine vane. Metal and ceramic coating temperatures were calculated as a function of coating thickness and coolant flow. With a thermal-barrier coating, the coolant flows required for the chosen sections were half those of an uncoated design, and the metal outer temperatures were simultaneously reduced by over 111 K (200 F). For comparison, transpiration cooling was also investigated. Full-coverage film cooling of a coated vane required more coolant flow than did transpiration cooling.

  17. Vortex-generating coolant-flow-passage design for increased film-cooling effectiveness and surface coverage

    NASA Technical Reports Server (NTRS)

    Papell, S. S.

    1984-01-01

    The thermal film-cooling footprints observed by infrared imagery for three coolant-passage configurations embedded in adiabatic-test plates are discussed. The configurations included a standard round-hole cross section and two orientations of a vortex-generating flow passage. Both orientations showed up to factors of four increases in both film-cooling effectiveness and surface coverage over that obtained with the round coolant passage. The crossflow data covered a range of tunnel velocities from 15.5 to 45 m/sec with blowing rates from 0.20 to 2.05. A photographic streakline flow visualization technique supported the concept of the counterrotating apability of the flow passage design and gave visual credence to its role in inhibiting flow separation.

  18. A new sensor for detection of coolant leakage in nuclear power plants using off-axis integrated cavity output spectroscopy

    NASA Astrophysics Data System (ADS)

    Lee, Lim; Park, Hyunmin; Kim, Taek-Soo; Ko, Kwang-Hoon; Jeong, Do-Young

    2012-06-01

    A new sensor based on laser absorption spectroscopy was developed for the detection of coolant leakage which may happen in pressurized heavy water reactor (PHWR). Off-axis integrated output spectroscopy (OA-ICOS) technique was adopted for developing a simple and robust sensor with sufficient sensitivity. Leak events could be monitored by detecting a small change in semi-heavy water (HDO) concentration induced by the exchange reaction of leaked heavy water (D2O) with light water (H2O). From the results of feasibility tests, we have shown that the measured area of absorption features was linearly correlated with HDO concentration, and the minimum detectable change of HDO concentration with the developed sensor was evaluated as 3.2 ppm. This new sensor is expected to be a reliable and promising device for the detection of coolant leakage since it has some advantages on real-time monitoring and early detection for nuclear safety.

  19. Performance Comparison of Axisymmetric and Three-dimensional Hydrogen Film Coolant Injection in a 110N Hydrogen/oxygen Rocket

    NASA Technical Reports Server (NTRS)

    Arrington, Lynn A.; Reed, Brian D.

    1992-01-01

    An experimental performance comparison of two geometrically different fuel film coolant injection sleeves was conducted on a 110 N gaseous hydrogen/oxygen rocket. One sleeve had slots milled axially down the walls and the other had a smooth surface to give axisymmetric flow. The comparison was made to investigate a conclusion in an earlier study that attributed a performance underprediction to a symplifying modeling assumption of axisymmetric fuel film flow. The smooth sleeve had higher overall performance at one film coolant percentage and approximately the same or slightly better at another. The study showed that the lack of modeling of three-dimensional effects was not the cause of the performance underprediction as speculated in earlier analytical studies.

  20. Evidence of zinc ferrite formation on carbon steel in primary-side coolant with added soluble zinc

    NASA Astrophysics Data System (ADS)

    Sawicki, Jerzy A.; Allsop, Heather A.

    1996-12-01

    Conversion electron and X-ray backscattering Mössbauer spectroscopy was used to identify iron compounds that were formed on carbon steel in simulated CANDU® reactor coolant containing 15 to 60 ppb (μg/L H 2O) soluble zinc. Analyses of the coupons exposed to coolant in the absence of zinc indicated the formation of magnetite, whereas, in the presence of zinc, the formation of a layer of zinc-bearing ferrite Zn xFe 3- xO 4 (with x > 0.8 was observed. The role of the high electrical resistivity of zinc ferrite in reducing corrosion rates, corrosion-product release rates, and 60Co activity buildup on carbon steel is briefly discussed.

  1. EXPERIMENT OPERATIONS PLAN FOR A LOSS-OF-COOLANT ACCIDENT SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    SciTech Connect

    Russcher, G. E.; Cannon, L. W.; Goodman, R. L.; Hesson, G. M.; King, L. L.; McDuffie, P. N.; Marshall, R. K.; Nealley, C.; Pilger, J. P.; Mohr, C. L.

    1981-04-01

    Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship between the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. This document contains both experiment proposal and assembly proposal information. The intent of this document is to supply information required by the Chalk River Nuclear Laboratories (CRNL), and to identify the planned procedures and data that will be used both to establish readiness to proceed from one test phase to the next and to operate the experiment. Operating control settings and limits are provided for both experimenter systems and CRNL systems. A hazards review summarizes safety issues that have been addressed during the development of the experiment plan.

  2. UO2 and PuO2 utilization in high temperature engineering test reactor with helium coolant

    NASA Astrophysics Data System (ADS)

    Waris, Abdul; Aji, Indarta K.; Novitrian, Pramuditya, Syeilendra; Su'ud, Zaki

    2016-03-01

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO2 fuel. In this study, we have evaluated the use of UO2 and PuO2 in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of 235U in loaded fuel is 18.0% or above.

  3. An Analysis of an Automatic Coolant Bypass in the International Space Station Node 2 Internal Active Thermal Control System

    NASA Technical Reports Server (NTRS)

    Clanton, Stephen E.; Holt, James M.; Turner, Larry D. (Technical Monitor)

    2001-01-01

    A challenging part of International Space Station (ISS) thermal control design is the ability to incorporate design changes into an integrated system without negatively impacting performance. The challenge presents itself in that the typical ISS Internal Active Thermal Control System (IATCS) consists of an integrated hardware/software system that provides active coolant resources to a variety of users. Software algorithms control the IATCS to specific temperatures, flow rates, and pressure differentials in order to meet the user-defined requirements. What may seem to be small design changes imposed on the system may in fact result in system instability or the temporary inability to meet user requirements. The purpose of this paper is to provide a brief description of the solution process and analyses used to implement one such design change that required the incorporation of an automatic coolant bypass in the ISS Node 2 element.

  4. Experimental results for film cooling in 2-D supersonic flow including coolant delivery pressure, geometry, and incident shock effects

    NASA Technical Reports Server (NTRS)

    Olsen, George C.; Nowak, Robert J.; Holden, Michael S.; Baker, N. R.

    1990-01-01

    An experimental program was conducted to establish some design parameters important to a supersonic film cooling system in a scramjet engine. A simple non-combusting two-dimensional flow configuration was used to isolate the film cooling phenomena. Parameters investigated include coolant delivery pressure, slot height and lip thickness, and incident shock location and strength. Design guidelines for use in engineering and trade studies are presented.

  5. Main-coolant-pump shaft-seal guidelines. Volume 1. Maintenance-manual guidelines. Final report. [PWR; BWR

    SciTech Connect

    Fair, C.E.; Greer, A.O.

    1983-03-01

    This report presents a set of guidelines and a listing of information and data which should be included in maintenance manuals and procedures for Main Coolant Pump Shaft Seals. The noted guidelines and data listing are developed from EPRI sponsored nuclear plant seal operating experience studies. The maintenance oriented results of the most recent such study is summarized. The shaft seal and its auxiliary supporting systems are discussed from both technical and maintenance related viewpoints.

  6. Coolant pressure and airflow distribution in a strut-supported transpiration-cooled vane for a gas turbine engine

    NASA Technical Reports Server (NTRS)

    Kaufman, A.; Poferl, D. J.; Richards, H. T.

    1972-01-01

    An analysis to predict pressure and flow distribution in a strut-supported wire-cloth vane was developed. Results were compared with experimental data obtained from room-temperature airflow tests conducted over a range of vane inlet airflow rates from 10.7 to 40.4 g/sec (0.0235 to 0.0890 lb/sec). The analytical method yielded reasonably accurate predictions of vane coolant flow rate and pressure distribution.

  7. In-core coolant velocity measurements in a pressurized water reactor using temperature-neutron noise cross correlation

    SciTech Connect

    Sweeney, F.J.

    1984-01-01

    Noise signals from an in-core, movable, neutron flux-mapping detector and a core-exit thermocouple at the Sequoyah-1 plant were cross correlated in order to investigate the feasibility of inferring in-core coolant flow velocities. If feasible, this technique might provide a cost effective way to detect local in-core flow blockages or to verify fuel assembly thermal-hydraulic performance. Previous ex-core neutron detector/core-exit thermocouple cross correlations at Sequoyah-1, indicated that, while coolant velocity measurements appeared to be feasible, the inferred velocities required correction for the relatively slow (approx.0.7-s) thermocouple time response. Furthermore, the effect on the inferred velocity due to the spatial averaging of the 1.8-m-long ex-core detector was not known. Noise measurements were therefore performed using a smaller neutron detector (0.48-cm diam by 5.33-cm long fission chamber), and a technique was developed to correct the inferred coolant velocities for the thermocouple response time.

  8. Probability of pipe failure in the reactor coolant loops of Combustion Engineering PWR plants. Volume 1. Summary report

    SciTech Connect

    Holman, G.S.; Lo, T.; Chou, C.K.

    1985-01-01

    As part of its reevaluation of the double-ended guillotine break (DEGB) as a design requirement for reactor coolant piping, the US Nuclear Regulatory Commission (NRC) contracted with the Lawrence Livermore National Laboratory (LLNL) to estimate the probability of occurrence of a DEGB, and to assess the effect that earthquakes have on DEGB probability. This report describes a probabilistic evaluation of reactor coolant loop piping in PWR plants having nuclear steam supply systems designed by Combustion Engineering. Two causes of pipe break were considered: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by failure of component supports due to an earthquake (indirect DEGB). The probability of direct DEGB was estimated using a probabilistic fracture mechanics model. The probability of indirect DEGB was estimated by estimating support fragility and then convolving fragility with seismic hazard. The results of this study indicate that the probability of a DEGB from either cause is very low for reactor coolant loop piping in these plants, and that NRC should therefore consider eliminating DEGB as a design basis in favor of more realistic criteria.

  9. Probability of pipe failure in the reactor coolant loops of Westinghouse PWR Plants. Volume 1. Summary report

    SciTech Connect

    Holman, G.S.; Chou, C.K.

    1985-07-01

    As part of its reevaluation of the double-ended guillotine break (DEGB) of reactor coolant loop piping as a design basis event for nuclear power plants, the US Nuclear Regulatory Commission (NRC) contracted with the Lawrence Livermore National Laboratory (LLNL) to estimate the probability of occurrence probability. This report describes a probabilistic evaluation of reactor coolant loop piping in PWR plants having nuclear steam supply systems designed by Westinghouse. Two causes of pipe break were considered: pipe fracture due to the growth of cracks at welded joints (''direct'' DEGB), and pipe rupture indirectly caused by failure of component supports due to an earthquake (''indirect'' DEGB). The probability of direct DEGB was estimated using a probabilistic fracture mechanics model. The probability of indirect DEGB was estimated by estimating support fragility and then convolving fragility and seismic hazard. The results of this study indicate that the probability of a DEGB from either cause is very low for reactor coolant loop piping in these plants, and that NRC should therefore consider eliminating DEGB as a design basis event in favor of more realistic criteria. 17 refs., 15 figs., 11 tabs.

  10. Development and Application of the Reactor Coolant On-Line Leakage Evaluation Model for Pressurized Water Reactors

    SciTech Connect

    Liang, Thomas K.S.; Hung, H.-J.; Chang, C.-J.

    2001-12-15

    With the consideration of mass unbalance, coolant shrinking, and compressibility, a model for reactor coolant leakage evaluation has been developed to quantify on-line the system leakage rate with conventional system measurements, regardless of where the leak occurs. This model has been derived from the system of total continuity, and it divides the reactor coolant system (RCS) into two regions, namely, the saturated and subcooled regions. The pressurizer is considered as a saturated region, and the remaining part of the RCS is regarded as a subcooled region. Taking the on-line measurements of the RCS including the RCS pressure, temperature, pressurizer water level, and charging and letdown flow rates, this model can directly evaluate on-line the RCS leakage rate. It is noted that this model is applicable only if the RCS remains subcooled. To verify the applicability of this model, data generated by RELAP5/MOD3 simulation and experimental measurements from the Institute of Nuclear Energy Research, Taiwan, Integral System Test Facility were adopted to assess this model. With further on-line verification against the Maanshan training simulator, this model was finally delivered to the Maanshan nuclear power plant (a three-looped Westinghouse pressurized water reactor) to assist the operator training and on-line evaluation of the RCS leakage rate. The smallest amount of leak flow that can be detected by the ROCK model is 3 gal/min.

  11. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Steady state and transient computations

    SciTech Connect

    Martin, A.; Alvarez, D.; Cases, F.

    1996-06-01

    After the Tchernobyl accident a working group was created to analyze the French PWR Safety with a respect to potential risk of reactivity accident. Potentially risky situations are those which can lead to heterogeneous boron concentration or temperature of the primary coolant fluid. This paper reports a Research and Development action based on numerical simulations and experiments on the primary coolant temperature or boron mixing capabilities in a PWR vessel. New numerical results obtained with the thermal hydraulic Finite Element (FE) Code N3S are presented. In these calculations the FE mesh takes into account the geometry of the lower plenum plates and columns. Two configurations have been investigated The first one is a steady state fluid flow mixing case. The reactor is cooled by free convection and the three loops, balanced in mass flow rate, are in operation. The second is a free boron plug transient case. It is related to the mixing of a clear plug injected in the vessel when a primary coolant pump starts-up. Two clear plug volumes have been investigated (3 and 8 m{sup 3}). Comparisons between these new numerical results and the data previously obtained (see Alvarez et al., 1992, Alvarez, Martin and Schneider, 1994) are presented in this paper.

  12. Development of a coolant channel helium and nitrogen gas ratio sensor for a high temperature gas reactor

    SciTech Connect

    Cadell, S. R.; Woods, B. G.

    2012-07-01

    To measure the changing gas composition of the coolant during a postulated High Temperature Gas Reactor (HTGR) accident, an instrument is needed. This instrument must be compact enough to measure the ratio of the coolant versus the break gas in an individual coolant channel. This instrument must minimally impact the fluid flow and provide for non-direct signal routing to allow minimal disturbance to adjacent channels. The instrument must have a flexible geometry to allow for the measurement of larger volumes such as in the upper or lower plenum of a HTGR. The instrument must be capable of accurately functioning through the full operating temperature and pressure of a HTGR. This instrument is not commercially available, but a literature survey has shown that building off of the present work on Capacitance Sensors and Cross-Capacitors will provide a basis for the development of the desired instrument. One difficulty in developing and instrument to operate at HTGR temperatures is acquiring an electrical conductor that will not melt at 1600 deg. C. This requirement limits the material selection to high temperature ceramics, graphite, and exotic metals. An additional concern for the instrument is properly accounting for the thermal expansion of both the sensing components and the gas being measured. This work covers the basic instrument overview with a thorough discussion of the associated uncertainty in making these measurements. (authors)

  13. Subcooled freon-11 flow boiling in top-heated finned coolant channels with and without a twisted tape

    NASA Technical Reports Server (NTRS)

    Smith, Alvin; Boyd, Ronald D., Sr.

    1989-01-01

    An experimental study was conducted in top-heated finned horizontal tubes to study the effect of enhancement devices on flow boiling heat transfer in coolant channels. The objectives are to examine the variations in both the mean and local (axial and circumferential) heat transfer coefficients for circular coolant channels with spiral finned walls and/or spiral fins with a twisted tape, and improve the data reduction technique of a previous investigator. The working fluid is freon-11 with an inlet temperature of 22.2 C (approximately 21 C subcooling). The coolant channel's exit pressure and mass velocity are 0.19 M Pa (absolute) and 0.21 Mg/sq. ms, respectively. Two tube configurations were examined; i.e., tubes had either 6.52 (small pitch) or 4.0 (large pitch) fins/cm of the circumferential length (26 and 16 fins, respectively). The large pitch fins were also examined with a twisted tape insert. The inside nominal diameter of the copper channels at the root of the fins was 1.0 cm. The results show that by adding enhancement devices, boiling occurs almost simultaneously at all axial locations. The case of spiral fins with large pitch resulted in larger mean (circumferentially averaged) heat transfer coefficients, h sub m, at all axial locations. Finally, when twisted tape is added to the tube with large-pitched fins, the power required for the onset of boiling is reduced at all axial and circumferential locations.

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

    SciTech Connect

    Scheele, Randall D.; Casella, Andrew M.

    2010-09-28

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

  15. An analytical study of the effect of coolant flow variables on the kinetic energy output of a cooled turbine blade flow

    NASA Technical Reports Server (NTRS)

    Prust, H. W., Jr.

    1971-01-01

    The results of an analytical study to determine the effect of changes in the amount, velocity, injection location, injection angle, and temperature of coolant flow on blade row performance are presented. The results show that the change in output of a cooled turbine blade row relative to the specific output of the uncooled blade row can be positive, negative, or zero. Comparisons between the analytical results and experimental results for four different cases of coolant discharge, all at a coolant temperature ratio of unity, show good agreement for three cases and rather poor agreement for the other. To further test the validity of the method, more experimental data is needed, particularly at different coolant temperature ratios.

  16. Models and numerical methods for the simulation of loss-of-coolant accidents in nuclear reactors

    NASA Astrophysics Data System (ADS)

    Seguin, Nicolas

    2014-05-01

    In view of the simulation of the water flows in pressurized water reactors (PWR), many models are available in the literature and their complexity deeply depends on the required accuracy, see for instance [1]. The loss-of-coolant accident (LOCA) may appear when a pipe is broken through. The coolant is composed by light water in its liquid form at very high temperature and pressure (around 300 °C and 155 bar), it then flashes and becomes instantaneously vapor in case of LOCA. A front of liquid/vapor phase transition appears in the pipes and may propagate towards the critical parts of the PWR. It is crucial to propose accurate models for the whole phenomenon, but also sufficiently robust to obtain relevant numerical results. Due to the application we have in mind, a complete description of the two-phase flow (with all the bubbles, droplets, interfaces…) is out of reach and irrelevant. We investigate averaged models, based on the use of void fractions for each phase, which represent the probability of presence of a phase at a given position and at a given time. The most accurate averaged model, based on the so-called Baer-Nunziato model, describes separately each phase by its own density, velocity and pressure. The two phases are coupled by non-conservative terms due to gradients of the void fractions and by source terms for mechanical relaxation, drag force and mass transfer. With appropriate closure laws, it has been proved [2] that this model complies with all the expected physical requirements: positivity of densities and temperatures, maximum principle for the void fraction, conservation of the mixture quantities, decrease of the global entropy… On the basis of this model, it is possible to derive simpler models, which can be used where the flow is still, see [3]. From the numerical point of view, we develop new Finite Volume schemes in [4], which also satisfy the requirements mentioned above. Since they are based on a partial linearization of the physical

  17. Probability of failure in BWR reactor coolant piping: Volume 1, Summary report

    SciTech Connect

    Holman, G.S.; Chou, C.K.

    1989-03-01

    This report summarizes a probabilistic reliability evaluation of BWR reactor coolant piping performed for the US Nuclear Regulatory Commission (NRC) by the Lawrence Livermore National Laboratory (LLNL). In this evaluation, LLNL estimated the probability of a double-ended guillotine break (DEGB) in the main steam, feedwater, and recirculation loop piping of a representative Mark I BWR plant. Two causes of pipe break were considered: crack growth at welded joints, and the earthquake-induced failure of supports for piping and components. A probabilistic fracture mechanics model, including intergranular stress corrosion cracking (IGSCC) in Types 304 and 316NG stainless steels, was used to estimate the probability of crack-induced pipe break. The probability of pipe break indirectly caused by support failure was estimated by applying reliability techniques to supports for ''heavy components'' such as the reactor pressure vessel, as well as to conventional pipe supports such as spring hangers and snubbers. Our probabilistic fracture mechanics evaluation found that the probability of crack-induced DEGB in main steam, feedwater and, if IGSCC is not a factor, recirculation piping is very low. In ISGCC-susceptible Type 304SS piping, stress corrosion dominates the probability of DEGB due mainly to cracks that initiate during the first few years of plant life; replacing Type 304 piping with IGSCC-resistant Type 316NG lowers DEGB probabilities by several orders of magnitude. We also found that the probability of pipe break caused by seismically-induced support failure is low regardless of whether ''heavy component'' supports or conventional pipe supports are being considered. 23 refs., 34 figs., 16 tabs.

  18. Test plan for high-burnup fuel cladding behavior under loss-of- coolant accident conditions

    SciTech Connect

    Chung, H.M.; Neimark, L.A.; Kassner, T.F.

    1996-10-01

    Excessive oxidation, hydriding, and extensive irradiation damage occur in high-burnup fuel cladding, and as result, mechanical properties of high-burnup fuels are degraded significantly. This may influence the current fuel cladding failure limits for loss-of- coolant-accident (LOCA) situations, which are based on fuel cladding behavior for zero burnup. To avoid cladding fragmentation and fuel dispersal during a LOCA, 10 CFR 50.46 requires that peak cladding temperature shall not exceed 1204 degrees C (2200 degrees F) and that total oxidation of the fuel cladding nowhere exceeds 0.17 times total cladding thickness before oxidation. Because of the concern, a new experimental program to investigate high-burnup fuel cladding behavior under LOCA situations has been initiated under the sponsorship of the U.S. Nuclear Regulatory Commission. A hot-cell test plan to investigate single-rod behavior under simulated LOCA conditions is described in this paper. In the meantime, industry fuel design and operating conditions are expected to undergo further changes as more advanced cladding materials are developed. Under these circumstances, mechanical properties of high-burnup fuel cladding require further investigation so that results from studies on LOCA, reactivity- initiated-accident (RIA), operational transient, and power-ramping situations, can be extrapolated to modified or advanced cladding materials and altered irradiation conditions without repeating major integral experiments in test reactors. To provide the applicable data base and mechanistic understanding, tests will be conducted to determine dynamic and static fracture toughness and tensile properties. Background and rationale for selecting the specific mechanical properties tests are also described.

  19. Influence of coolant on ductile mode processing of binderless nanocrystalline tungsten carbide through ultraprecision diamond turning

    NASA Astrophysics Data System (ADS)

    Doetz, Marius; Dambon, Olaf; Klocke, Fritz; Fähnle, Oliver

    2015-08-01

    Molds made of tungsten carbide are typically used for the replicative mass production of glass lenses by precision glass molding. Consequently an ultra-precision grinding process with a subsequent fresh-feed polishing operation is conventionally applied. These processes are time consuming and have a relatively low reproducibility. An alternative manufacturing technology, with a high predictability and efficiency, which additionally allows a higher geometrical flexibility, is the single point diamond turning technique (SPDT). However, the extreme hardness and the chemical properties of tungsten carbide lead to significant tool wear and therefore the impossibility of machining the work pieces in an economical way. One approach to enlarge the tool life is to affect the contact zone between tool and work piece by the use of special cutting fluids. This publication emphasizes on the most recent investigations and results in direct machining of nano-grained tungsten carbide with mono crystal diamonds under the influence of various kinds of cutting fluids. Therefore basic ruling experiments on binderless nano grained tungsten carbide were performed, where the tool performed a linear movement with a steadily increasing depth of cut. As the ductile cutting mechanism is a prerequisite for the optical manufacturing of tungsten carbide these experiments serve the purpose for establish the influence of different cutting fluid characteristics on the cutting performance of mono crystal diamonds. Eventually it is shown that by adjusting the coolant fluid it is possible to significantly shift the transition point from ductile to brittle removal to larger depths of cut eventually enabling a SPDT of binderless tungsten carbide molds.

  20. Comparison of Lead-Bismuth and Lead as Coolants for Accelerator Driven Systems

    SciTech Connect

    Bianchi, F.; Mattioda, F.; Meloni, P.

    2002-07-01

    In the framework of the Italian research program TRASCO (TRAsmutazione SCOrie, namely transmutation of radioactive wastes) and of the European research program PDS-XADS (Preliminary Design Study on an eXperimental Accelerator Driven System) the feasibility and operability of gas or liquid metal cooled accelerator driven system prototypes are currently under investigation. Initially the attention of the thermal-hydraulics group of ENEA research centre in Bologna has been focussed toward a lead-bismuth cooled subcritical system under natural or enhanced natural circulation according to the prototype design proposed. The interest in using lead as a coolant, which is characterized by a higher melting point, is explained by the need to increase the plant efficiency for the economic competitiveness, though the higher temperatures pose some technological problems. Moreover, the amount of activation products should result significantly lower. Of course the results obtained and the experience gained analysing the dynamical behaviour of the lead-bismuth cooled system cannot be directly transferred to lead cooled systems. This paper aims at presenting a preliminary comparison of lead-bismuth and lead in a simplified liquid metal cooled subcritical system, mainly from the thermal-hydraulics and system dynamics points of view. By means of the modified RELAP5 version, the dynamical behavior of a lead-bismuth or lead cooled system, which is intended to be a quite accurate representation of the Italian accelerator driven prototype XADS, has been studied. Although a more exhaustive comparison should take into account the necessarily different structural characteristics of lead-bismuth and lead cooled systems, the neutronic feedback on reactor power and also the slightly different neutronic properties of lead-bismuth and lead, the purely thermal-hydraulic analysis presented in this paper has shown that the dynamical behaviour of the XADS does not differ noticeable when lead is used

  1. Electrochemical behaviour of stainless steel in PWR primary coolant conditions: Effects of radiolysis

    NASA Astrophysics Data System (ADS)

    Muzeau, Benoist; Perrin, Stéphane; Corbel, Catherine; Simon, Dominique; Feron, Damien

    2011-12-01

    Few data are available in the literature on the role of the water radiolysis on the corrosion of stainless steel core components in PWR operating conditions (300 °C, 155 bar). The present approach uses a high energy proton beam to control the production of radiolytic species at the interface between a stainless steel sample and water in a high temperature and high pressure (HP-HT) electrochemical cell working in the range 25 °C/1 bar-300 °C/90 bar. The cell is designed to record the free corrosion potential of the AISI 316L/water interface mounted in line with a cyclotron delivering the proton beam. The evolution of the potential is compared before, during and after the proton irradiation. The first results are obtained with an aqueous solution containing boron, lithium and dissolved hydrogen, as in PWR primary coolant circuit. The stainless steel/water interfaces are irradiated between 25 °C and 300 °C with protons emerging at 22 MeV at the interface. The flux is varied by five orders of magnitude, from 6.6 × 10 11 to 6.6 × 10 15 H + m -2 s -1. The evolution of the free corrosion potential is highly dependent on the temperature and/or pressure. For a given temperature and pressure, it evolves with the flux and the ageing of the AISI 316L/water interfaces. An important role of the temperature of irradiation on the electrochemical response was observed. These results give a better understanding of the role of radiolysis on stainless steel corrosion in high temperature conditions.

  2. Water chemistry used in the secondary coolant circuit of unit 3 at the rovno nuclear power station involving correction treatment of working medium with lithium hydroxide and ethanolamine

    NASA Astrophysics Data System (ADS)

    Kozlov, V. Ya.; Vlasenko, N. I.; Kozlova, T. Yu.

    2011-03-01

    The all-volatile water chemistry used in the secondary coolant circuit involving correction treatment of the steam generator's boiler water with lithium hydroxide and the ethanolamine water chemistry are analyzed from the viewpoint of their effect on the erosion-corrosion wear of equipment used in the secondary coolant system and damageability of heat-transfer tubes used in PGV-1000M steam generators.

  3. Comparison of Calculated and Experimental Temperatures and Coolant Pressure Losses for a Cascade of Small Air-Cooled Turbine Rotor Blades

    NASA Technical Reports Server (NTRS)

    Stepka, Francis S

    1958-01-01

    Average spanwise blade temperatures and cooling-air pressure losses through a small (1.4-in, span, 0.7-in, chord) air-cooled turbine blade were calculated and are compared with experimental nonrotating cascade data. Two methods of calculating the blade spanwise metal temperature distributions are presented. The method which considered the effect of the length-to-diameter ratio of the coolant passage on the blade-to-coolant heat-transfer coefficient and assumed constant coolant properties based on the coolant bulk temperature gave the best agreement with experimental data. The agreement obtained was within 3 percent at the midspan and tip regions of the blade. At the root region of the blade, the agreement was within 3 percent for coolant flows within the turbulent flow regime and within 10 percent for coolant flows in the laminar regime. The calculated and measured cooling-air pressure losses through the blade agreed within 5 percent. Calculated spanwise blade temperatures for assumed turboprop engine operating conditions of 2000 F turbine-inlet gas temperature and flight conditions of 300 knots at a 30,000-foot altitude agreed well with those obtained by the extrapolation of correlated experimental data of a static cascade investigation of these blades.

  4. Modeling Cladding-Coolant Heat Transfer of High-Burnup Fuel During RIA

    SciTech Connect

    Wenfeng Liu; Kazimi, Mujid S.

    2006-07-01

    This paper describes a model for the cladding-coolant heat transfer of high burnup fuel during a Reactivity Initiated Accident (RIA) which is implemented in the fuel performance code FRAPTRAN 1.2. The minimum stable film boiling temperature, affected by the subcooling and the clad oxidation, is modeled by a modified Henry correlation. This accounts for the effects of thermal properties of the cladding surface on the transient temperature drop during liquid-solid contact. The transition boiling regime is described as the interpolation of the heat flux between two anchor points on the boiling curve: the Critical Heat Flux (CHF) and minimum stable film boiling. The CHF correlation is based on the Zuber hydrodynamic model multiplied by a subcooling factor. Frederking correlation is chosen to model the film boiling regime. The heat conduction through the oxide layer of the cladding surface of high burnup fuel is calculated by solving heat conduction equations with thermal properties of zirconia taken from MATPRO. This model is validated in the FRAPTRAN code for test cases of both high burnup and fresh test fuel rods including the burnup level (0--56 MW d/kg), peak fuel enthalpy deposit (70--190 cal/g), degree of subcooling (0--80 deg. C), and extent of oxidation (0--25 micron). The modified code demonstrates the capability of differentiating between the departure from nucleate boiling (DNB) and none-DNB cases. The predicted peak cladding temperature (PCT) and duration of DNB achieves generally good agreement with the experimental data. It is found that the cladding surface oxidation of high burnup fuel causes an early rewetting of cladding or suppresses DNB due to two factors: 1) Thick zirconia layer may delay the heat conducted to the surface while keeping the surface heat transfer in the most effective nucleate boiling regime. 2) The transient liquid-solid contact resulting from vapor breaking down would cause a lower interface temperature for an oxidized surface

  5. Flibe Coolant Cleanup and Processing in the HYLIFE-II Inertial Fusion Energy Power Plant

    SciTech Connect

    Moir, R W

    2001-03-23

    In the HYLIFE-II chamber design, a thick flowing blanket of molten-salt (Li{sub 2}BeF{sub 4}) called flibe is used to protect structures from radiation damage. Since it is directly exposed to the fusion target, the flibe will absorb the target debris. Removing the materials left over from target explosions at the rate of {approx}6/s and then recycling some of these materials poses a challenge for the inertial fusion energy power plant. The choice of target materials derives from multi-disciplinary criteria such as target performance, fabricability, safety and environment, corrosion, and cost of recycle. Indirect-drive targets require high-2 materials for the hohlraum. Gold and gadolinium are favorite target materials for laboratory experiments but cost considerations may preclude their use in power plants or at least requires cost effective recycle because a year's supply of gold and gadolinium is estimated at 520 M$ and 40 M$. Environmental and waste considerations alone require recycle of this material. Separation by volatility appears to be the most attractive (e.g., Hg and Xe); centrifugation (e.g., Pb) is acceptable with some problems (e.g., materials compatibility) and chemical separation is the least attractive (e.g. Gd and Hf). Mercury, hafnium and xenon might be substituted with equal target performance and have advantages in removal and recycle due to their high volatility, except for hafnium. Alternatively, lead, tungsten and xenon might be used due to the ability to use centrifugation and gaseous separation. Hafnium or tantalum form fluorides, which will complicate materials compatibility, corrosion and require sufficient volatility of the fluoride for separation. Further complicating the coolant cleanup and processing is the formation of free fluorine due to nuclear transformation of lithium and beryllium in the flibe, which requires chemical control of the fluoride level to minimize corrosion. The study of the choice of target materials and the

  6. The aerodynamic effects of wheelspace coolant injection into the mainstream flow of a high pressure gas turbine

    NASA Astrophysics Data System (ADS)

    McLean, Christopher Elliot

    Modern gas turbine engines operate with mainstream gas temperatures exceeding 1450°C in the high-pressure turbine stage. Unlike turbine blades, rotor disks and other internal components are not designed to withstand the extreme temperatures found in mainstream flow. In modern gas turbines, cooling air is pumped into the wheelspace cavities to prevent mainstream gas ingestion and then exits through a seal between the rotor and the nozzle guide vane (NGV) thereby mixing with the mainstream flow. The primary purpose for the wheelspace cooling air is the cooling of the turbine wheelspace. However, secondary effects arise from the mixing of the spent cooling air with the mainstream flow. The exiting cooling air is mixed with the hot mainstream flow effecting the aerodynamic and performance characteristics of the turbine stage. The physics underlying this mixing process and its effects on stage performance are not yet fully understood. The relative aerodynamic and performance effects associated with rotor - NGV gap coolant injections were investigated in the Axial Flow Turbine Research Facility (AFTRF) of the Center for Gas Turbines and Power of The Pennsylvania State University. This study quantifies the secondary effects of the coolant injection on the aerodynamic and performance character of the turbines main stream flow for root injection, radial cooling, and impingement cooling. Measurement and analysis of the cooling effects were performed in both stationary and rotational frames of reference. The AFTRF is unique in its ability to perform long duration cooling measurements in the stationary and rotating frames. The effects of wheelspace coolant mixing with the mainstream flow on total-to-total efficiency, energy transport, three dimensional velocity field, and loading coefficient were investigated. Overall, it was found that a small quantity (1%) of cooling air can have significant effects on the performance character and exit conditions of the high pressure stage

  7. Radiogenic lead with dominant content of {sup 208}Pb: New coolant and neutron moderator for innovative nuclear reactors

    SciTech Connect

    Shmelev, A. N.; Kulikov, G. G.; Kryuchkov, E. F.; Apse, V. A.; Kulikov, E. G.

    2012-07-01

    The advantages of radiogenic lead with dominant content of {sup 208}Pb as a reactor coolant with respect to natural lead are caused by unique nuclear properties of {sup 208}Pb which is a double-magic nucleus with closed proton and neutron shells. This results in significantly lower micro cross section and resonance integral of radiative neutron capture by {sup 208}Pb than those for numerous light neutron moderators. The extremely weak ability of {sup 208}Pb to absorb neutrons results in the following effects. Firstly, neutron moderating factor (ratio of scattering to capture cross sections) is larger than that for graphite and light water. Secondly, age and diffusion length of thermal neutrons are larger than those for graphite, light and heavy water. Thirdly, neutron lifetime in {sup 208}Pb is comparable with that for graphite, beryllium and heavy water what could be important for safe reactor operation. The paper presents some results obtained in neutronics and thermal-hydraulics evaluations of the benefits from the use of radiogenic lead with dominant content of {sup 208}Pb instead of natural lead as a coolant of fast breeder reactors. The paper demonstrates that substitution of radiogenic lead for natural lead can offer the following benefits for operation of fast breeder reactors. Firstly, improvement of the reactor safety thanks to the better values of coolant temperature reactivity coefficient and, secondly, improvement of some thermal-hydraulic reactor parameters. Radiogenic lead can be extracted from thorium sludge without isotope separation as {sup 208}Pb is a final isotope in the decay chain of {sup 232}Th. (authors)

  8. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    SciTech Connect

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.

  9. An L2F-measurement device with image rotator prism for flow velocity analysis in rotating coolant channels

    NASA Astrophysics Data System (ADS)

    Beversdorff, M.; Hein, O.; Schodl, R.

    1993-02-01

    For further improvement of the turbine blade cooling process, the knowledge concerning the heat transfer in radial coolant channels has to be deepened. Due to rotation, the velocity distribution, as well as the turbulence structure and therefore the heat transfer, will be influenced. To carry out experimental data of the flow field within a rotating duct a non-intrusive continuous measuring system (Laser-Two-Focus) with an image rotator prism is presented. The design of the system is explained in detail. Problems of application are discussed and results of the first successful measurements compared with numerical results are presented.

  10. 10 CFR 830 Major Modification Determination for Replacement of ATR Primary Coolant Pumps and Motors

    SciTech Connect

    Noel Duckwitz

    2011-05-01

    The continued safe and reliable operation of the ATR is critical to the Department of Energy (DOE) Office of Nuclear Energy (NE) mission. While ATR is safely fulfilling current mission requirements, a variety of aging and obsolescence issues challenge ATR engineering and maintenance personnel’s capability to sustain ATR over the long term. First documented in a series of independent assessments, beginning with an OA Environmental Safety and Health Assessment conducted in 2003, the issues were validated in a detailed Material Condition Assessment (MCA) conducted as a part of the ATR Life Extension Program in 2007.Accordingly, near term replacement of aging and obsolescent original ATR equipment has become important to ensure ATR capability in support of NE’s long term national missions. To that end, a mission needs statement has been prepared for a non-major system acquisition which is comprised of three interdependent subprojects. The first project will replace the existent diesel-electrical bus (E-3), switchgear, and the 50-year-old obsolescent marine diesels with commercial power that is backed with safety related emergency diesel generators, switchgear, and uninterruptible power supply (UPS). The second project, the subject of this major modification determination, will replace the four, obsolete, original primary coolant pumps (PCPs) and motors. Completion of this and the two other age-related projects (replacement of the ATR diesel bus [E-3] and switchgear and replacement of the existent emergency firewater injection system) will resolve major age-related operational issues plus make a significant contribution in sustaining the ATR safety and reliability profile. The major modification criteria evaluation of the project pre-conceptual design identified several issues that lead to the conclusion that the project is a major modification: 1. Evaluation Criteria #3 (Change of existing process). The proposed strategy for equipping the replacement PCPs with VFDs

  11. Cooling Characteristics of the V-1650-7 Engine. 1; Coolant-Flow Distribution, Cylinder Temperatures, and Heat Rejections at Typical Operating Conditions

    NASA Technical Reports Server (NTRS)

    Povolny, John H.; Bogdan, Louis J.

    1947-01-01

    An investigation was conducted to determine the coolant-flow distribu tion, the cylinder temperatures, and the heat rejections of the V-165 0-7 engine . The tests were run a t several power levels varying from minimum fuel consumption to war emergency power and at each power l evel the coolant flows corresponded to the extremes of those likely t o be encountered in typical airplane installations, A mixture of 30-p ercent ethylene glycol and 70-percent water was used as the coolant. The temperature of each cylinder was measured between the exhaust val ves, between the intake valves, in the center of the head, on the exh aust-valve guide, at the top of the barrel on the exhaust side, and o n each exhaust spark-plug gasket. For an increase in engine power fro m 628 to approximately 1700 brake horsepower the average temperature for the cylinder heads between the exhaust valves increased from 437 deg to 517 deg F, the engine coolant heat rejection increased from 12 ,600 to 22,700 Btu. per minute, the oil heat rejection increased from 1030 to 4600 Btu per minute, and the aftercooler-coolant heat reject ion increased from 450 to 3500 Btu -per minute.

  12. Investigation of aerodynamic effects of coolant ejection at the trailing edge of a turbine blade model by PIV and pressure measurements

    NASA Astrophysics Data System (ADS)

    Raffel, M.; Kost, F.

    In order to simulate the thick trailing edges of turbine blades a slotted plate profile together with a newly designed nozzle was installed into the high-speed wind tunnel of the DLR Göttingen. At different supersonic Mach numbers and at four coolant flow rates in the range of 0-2.5% pressure distribution measurements and probe measurements were performed. The flow field was visualized by schlieren photos and the instantaneous velocity field was quantitatively investigated by Particle Image Velocimetry (PIV). The measurements of the velocity field gave an insight into stationary effects, for example the change of shock strength with coolant flow rate, and instationary effects such as the existence of a vortex street in the wake. The PIV technique offers special advantages for the investigation of transonic flow fields, but also yields to special experimental difficulties, which are also described in this article. Measured losses display a maximum at the downstream Mach number 1. This is strongly related to the behaviour of the base pressure. A loss minimum is achieved at moderate coolant flow rates, showing that an optimum coolant flow rate exists. The loss was analysed and separated into the loss contributions from the profile upstream of the trailing edge and the mixing loss due to the coolant flow.

  13. Comparison of central axis and jet ring coolant supply for turbine disk cooling on a SSME-HPOTP model

    NASA Technical Reports Server (NTRS)

    Kim, Y. W.; Metzger, D. E.

    1992-01-01

    The test facility, test methods and results are presented for an experimental study modeling the cooling of turbine disks in the blade attachment regions with multiple impinging jets, in a configuration simulating the disk cooling method employed on the Space Shuttle Main Engine oxygen turbopump. The study's objective was to provide a comparison of detailed local convection heat transfer rates obtained for a single center-supply of disk coolant with those obtained with the present flight configuration where disk coolant is supplied through an array of 19 jets located near the disk outer radius. Specially constructed disk models were used in a program designed to evaluate possible benefits and identify any possible detrimental effects involved in employing an alternate disk cooling scheme. The study involved the design, construction and testing of two full scale rotating model disks, one plane and smooth for baseline testing and the second contoured to the present flight configuration, together with the corresponding plane and contoured stator disks. Local heat transfer rates are determined from the color display of encapsulated liquid crystals coated on the disk in conjunction with use of a computer vision system. The test program was composed of a wide variety of disk speeds, flowrates, and geometrical configurations, including testing for the effects of disk boltheads and gas ingestion from the gas path region radially outboard of the disk-cavity.

  14. Effects of turbulence model on convective heat transfer of coolant flow in a prismatic very high temperature reactor core

    SciTech Connect

    Lee, S. N.; Tak, N. I.; Kim, M. H.; Noh, J. M.

    2012-07-01

    The existing study of Spall et al. shows that only {nu}{sup 2}-f turbulence model well matches with the experimental data of Shehata and McEligot which were obtained under strongly heated gas flows. Significant over-predictions in those literatures were observed in the convective heat transfer with the other famous turbulence models such as the k-{epsilon} and k-{omega} models. In spite of such good evidence about the performance of the{nu}{sup 2}-f model, the application of the {nu}{sup 2}-f model to the thermo-fluid analysis of a prismatic core is very rare. In this paper, therefore, the convective heat transfer of the coolant flow in a prismatic core has been investigated using the {nu}{sup 2}-f model. Computational fluid dynamics (CFD) calculations have been carried out for the typical unit cell geometry of a prismatic fuel column with typical operating conditions of prismatic designs. The tested Reynolds numbers of the coolant flow are 10,000, 20,000, 30,000 and 50,000. The predicted Nusselt numbers with the {nu}{sup 2}-f model are compared with the results by the other turbulence models (k-{epsilon} and SST) as well as the empirical correlations. (authors)

  15. A Case Study Of Applying Infrared Thermography To Identify A Coolant Leak In A Municipal Ice Skating Rink

    NASA Astrophysics Data System (ADS)

    Wallace, Jay R.

    1989-03-01

    This paper deals with the application of infrared imaging radiometry as a diagnostic inspection tool for locating a concealed leak in the refrigeration system supplying glycol coolant to the arena floor of an ice skating rink in a municipal coliseum facility. Scanning approximately 10 miles of black iron tubing embedded in the arena floor resulted in locating a leak within the supply/return side of the system. A secondary disclosure was a restriction to normal coolant flow in some delivery loops caused by sludge build-up. Specific inspection procedures were established to enhance temperature differentials suitable for good thermal imaging. One procedure utilized the temperature and pressure of the city water supply; a second the availability of 130F hot water from the facility's boiler system; and a third the building's own internal ambient temperature. Destructive testing and other data collection equipment confirmed the thermographic findings revealing a section of corrosion damaged pipe. Repair and flushing of the system was quickly completed with a minimum of construction costs and inconvenience. No financial losses were incurred due to the interruption of scheduled revenue events. Probable cause for the shutdown condition was attributed to a flawed installation decision made 15 years earlier during the initial construction stage.

  16. Performance of the Extravehicular Mobility Unit (EMU): Airlock Coolant Loop Recovery (A/L CLR) Hardware - Phase II

    NASA Technical Reports Server (NTRS)

    Steele, John; Rector, tony; Gazda, Daniel; Lewis, John

    2009-01-01

    An EMU water processing kit (Airlock Coolant Loop Recovery A/L CLR) was developed as a corrective action to Extravehicular Mobility Unit (EMU) coolant flow disruptions experienced on the International Space Station (ISS) in May of 2004 and thereafter. Conservative schedules for A/L CLR use and component life were initially developed and implemented based on prior analysis results and analytical modeling. The examination of postflight samples and EMU hardware in November of 2006 indicated that the A/L CLR kits were functioning well and had excess capacity that would allow a relaxation of the initially conservative schedules of use and component life. A relaxed use schedule and list of component lives was implemented thereafter. Since the adoption of the relaxed A/L CLR schedules of use and component lives, several A/L CLR kit components, transport loop water samples and sensitive EMU transport loop components have been examined to gage the impact of the relaxed requirements. The intent of this paper is to summarize the findings of that evaluation, and to outline updated schedules for A/L CLR use and component life.

  17. International Space Station Active Thermal Control Sub-System On-Orbit Pump Performance and Reliability Using Liquid Ammonia as a Coolant

    NASA Technical Reports Server (NTRS)

    Morton, Richard D.; Jurick, Matthew; Roman, Ruben; Adamson, Gary; Bui, Chinh T.; Laliberte, Yvon J.

    2011-01-01

    The International Space Station (ISS) contains two Active Thermal Control Sub-systems (ATCS) that function by using a liquid ammonia cooling system collecting waste heat and rejecting it using radiators. These subsystems consist of a number of heat exchangers, cold plates, radiators, the Pump and Flow Control Subassembly (PFCS), and the Pump Module (PM), all of which are Orbital Replaceable Units (ORU's). The PFCS provides the motive force to circulate the ammonia coolant in the Photovoltaic Thermal Control Subsystem (PVTCS) and has been in operation since December, 2000. The Pump Module (PM) circulates liquid ammonia coolant within the External Active Thermal Control Subsystem (EATCS) cooling the ISS internal coolant (water) loops collecting waste heat and rejecting it through the ISS radiators. These PM loops have been in operation since December, 2006. This paper will discuss the original reliability analysis approach of the PFCS and Pump Module, comparing them against the current operational performance data for the ISS External Thermal Control Loops.

  18. Cold-air study of the effect on turbine stator blade aerodynamic performance of coolant ejection from various trailing-edge slot geometries. 1: Experimental results

    NASA Technical Reports Server (NTRS)

    Prust, H. W., Jr.; Bartlett, W. M.

    1974-01-01

    Trailing-edge slot configurations were investigated in a two-dimensional cascade of turbine stator blades. The trailing-edge slots were incorporated into blades with round trailing edges. The five blade configurations investigated included blades with two different trailing-edge thicknesses and four different slot widths. The results of the investigation showed that there was, in general, a significant increase in primary-air efficiency due to the coolant flow, the increase varying with slot configuration. For the five configurations tested, the average percent change in primary-air efficiency per percent coolant flow varied almost linearly from zero to about 1.4 percent over a range of coolant- to primary-air exit-velocity ratios between 0 and 1.2. However, for different configurations there was considerable deviation from the average values in the lower range of exit velocity ratios.

  19. A MODEL FOR PREDICTING FISSION PRODUCT ACTIVITIES IN REACTOR COOLANT: APPLICATION OF MODEL FOR ESTIMATING I-129 LEVELS IN RADIOACTIVE WASTE

    SciTech Connect

    Lewis, B.J.; Husain, A.

    2003-02-27

    A general model was developed to estimate the activities of fission products in reactor coolant and hence to predict a value for the I-129/Cs-137 scaling factor; the latter can be applied along with measured Cs-137 activities to estimate I-129 levels in reactor waste. The model accounts for fission product release from both defective fuel rods and uranium contamination present on in-core reactor surfaces. For simplicity, only the key release mechanisms were modeled. A mass balance, considering the two fuel source terms and a loss term due to coolant cleanup was solved to estimate fission product activity in the primary heat transport system coolant. Steady state assumptions were made to solve for the activity of shortlived fission products. Solutions for long-lived fission products are time-dependent. Data for short-lived radioiodines I-131, I-132, I-133, I-134 and I-135 were analyzed to estimate model parameters for I-129. The estimated parameter values were then used to determine I-1 29 coolant activities. Because of the chemical affinity between iodine and cesium, estimates of Cs-137 coolant concentrations were also based on parameter values similar to those for the radioiodines; this assumption was tested by comparing measured and predicted Cs-137 coolant concentrations. Application of the derived model to Douglas Point and Darlington Nuclear Generating Station plant data yielded estimates for I-129/I-131 and I-129/Cs-137 which are consistent with values reported for pressurized water reactors (PWRs) and boiling water reactors (BWRs). The estimated magnitude for the I-129/Cs-137 ratio was 10-8 - 10-7.

  20. Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Tests 1 and 2

    SciTech Connect

    Russcher, G. E.; Wilson, C. L.; Marshall, R, K.; King, L. L.; Parchen, L. J.; Pilger, J. P.; Hesson, G. M.; Mohr, C. L.

    1981-09-01

    A loss of Coolant Accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects of LOCA conditions on pressurized water reactor test fuel bundles. This experiment operation plan for the second and third experiments of the program will provide peak fuel cladding temperatures of up to 1172K (1650{degree}F) and 1061K (1450{degree}) respectively. for a long enough time to cause test fuel cladding deformation and rupture in both. Reflood coolant delay times and the reflooding rates for the experiments were selected from thermal-hydraulic data measured in the National Research Universal (NRU) reactor facilities and test train assembly during the first experiment.

  1. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 6: failure mode analysis. Final report

    SciTech Connect

    Streit, R.D.

    1981-09-01

    Material properties and failure criteria were evaluated to assess the requirements for double-ended guillotine break in the primary coolant loop of the Zion Unit 1 pressurized water reactor. The properties of the 316 stainless steel piping materials were obtained from the literature. Statistical distributions of both the tensile and fracture properties at room and operating temperatures were developed. Yield and ultimate strength tensile properties were combined to estimate the material flow strength. The flow strength and fracture properties were used in the various failure models analyzed. Linear-elastic, elastic-plastic, and fully plastic fracture models were compared, and the governing fracture criterion was determined. For the particular case studied, the fully plastic flow requirement was found to be the controlling fracture criterion leading to a double-ended guillotine pipe break.

  2. FORTRAN program for calculating coolant flow and metal temperatures of a full-coverage-film-cooled vane or blade

    NASA Technical Reports Server (NTRS)

    Meitner, P. L.

    1978-01-01

    A computer program that calculates the coolant flow and the metal temperatures of a full-coverage-film-cooled vane or blade was developed. The analysis was based on compressible, one-dimensional fluid flow and on one-dimensional heat transfer and treats the vane or blade shell as a porous wall. The calculated temperatures are average values for the shell outer-surface area associated with each film-cooling hole row. A thermal-barrier coating may be specified on the shell outer surface, and centrifugal effects can be included for blade calculations. The program is written in FORTRAN 4 and is operational on a UNIVAC 1100/42 computer. The method of analysis, the program input, the program output, and two sample problems are provided.

  3. Large-Break Loss-of-Coolant Accident Testing and Simulation for 200-MWe Simplified Boiling Water Reactor

    SciTech Connect

    Revankar, S.T.; Xu, Y.; Yoon, H.J.; Ishii, M.

    2002-07-01

    The performance of the safety systems of a new design of the 200-MWe simplified boiling water reactor during a large-break, loss-of-coolant accident transient was investigated through code modeling and integral system testing. The accident considered was a break in the main steam line which is the major design basis accident. RELAP5/MOD3 best estimate reactor thermalhydraulic code was used and its applicability to the reactor safety system evaluation was examined. The integral tests were performed to assess the safety systems and the response of the emergency core cooling systems to accident conditions in a scaled facility called PUMA. The details of the safety system behavior are presented. The integral test simulations examined code applicability at the scaled facility level as well as prototype key safety system performance. (authors)

  4. Internal cooling of a lithium-ion battery using electrolyte as coolant through microchannels embedded inside the electrodes

    NASA Astrophysics Data System (ADS)

    Mohammadian, Shahabeddin K.; He, Ya-Ling; Zhang, Yuwen

    2015-10-01

    Two and three dimensional transient thermal analysis of a prismatic Li-ion cell has been carried out to compare internal and external cooling methods for thermal management of Lithium Ion (Li-ion) battery packs. Water and liquid electrolyte have been utilized as coolants for external and internal cooling, respectively. The effects of the methods on decreasing the temperature inside the battery and also temperature uniformity were investigated. The results showed that at the same pumping power, using internal cooling not only decreases the bulk temperature inside the battery more than external cooling, but also decreases the standard deviation of the temperature field inside the battery significantly. Finally, using internal cooling decreases the intersection angle between the velocity vector and the temperature gradient which according to field synergy principle (FSP) causes to increase the convection heat transfer.

  5. Hydrothermal corrosion of SiC in LWR coolant environments in the absence of irradiation

    NASA Astrophysics Data System (ADS)

    Terrani, K. A.; Yang, Y.; Kim, Y.-J.; Rebak, R.; Meyer, H. M.; Gerczak, T. J.

    2015-10-01

    Assessment of the thermodynamics of SiC corrosion under light water reactor coolant environments suggests that silica formation is always expected in the range of applicable pH and potential. Autoclave testing of SiC-based materials in the absence of ionizing radiation was performed. The kinetics data from these tests, when compared with kinetics of silica dissolution in water and post-exposure characterization of SiC samples, suggest that oxidation of SiC to form silica is the rate-limiting step for recession of SiC in high temperature water. Oxygen activity in water was determined to play an important role in SiC recession kinetics. A simplified model of a power loop shows the effect of silica dissolution from the hot region (resembling fuel) and deposition in the cold regions.

  6. Local heat transfer in turbine disk-cavities. I - Rotor and stator cooling with hub injection of coolant

    NASA Astrophysics Data System (ADS)

    Bunker, R. S.; Metzger, D. E.; Wittig, S.

    1990-06-01

    Detailed radial heat-transfer coefficient distributions applicable to the cooling of disk-cavity regions of gas turbines are obtained experimentally from local heat-transfer data on both the rotating and stationary surfaces of a parallel-geometry disk-cavity system. Attention is focused on the hub injection of a coolant over a wide range of parameters including disk rotational Reynolds numbers of 200,000 to 50,000, rotor/stator spacing-to-disk ratios of 0.025 to 0.15, and jet mass flow rates between 0.10 and 0.40 times the turbulent pumped flow rate of a free disk. It is shown that rotor heat transfer exhibits regions of impingement and rotational domination with a transition region between, while stator heat transfer displays flow reattachment and convection regions with an inner recirculation zone.

  7. Local heat transfer in turbine disk-cavities. II - Rotor cooling with radial location injection of coolant

    NASA Astrophysics Data System (ADS)

    Bunker, R. S.; Metzger, D. E.; Wittig, S.

    1990-06-01

    The detailed radial distributions of rotor heat-transfer coefficients for three basic disk-cavity geometries applicable to gas turbines are presented. The coefficients are obtained over a range of parameters including disk rotational Reynolds numbers of 200,000 to 50,000, rotor/stator spacing-to-disk ratios of 0.025 to 0.15, and jet mass flow rates between 0.10 and 0.40 times the turbulent pumped flow rate of a free disk. The effects of a parallel rotor are analyzed, and strong variations in local Nusselt numbers for all but the rotational speed are pointed out and compared with the associated hub-injection data from a previous study. It is demonstrated that the overall rotor heat transfer is optimized by either the hub injection or radial location injection of a coolant, dependent on the configuration.

  8. Survey of tracking systems and rotary joints for coolant piping. Final report, August 15, 1978-August 14, 1978. [Includes patents

    SciTech Connect

    Furaus, J P; Gruchalla, M E; Sower, G D

    1980-01-01

    Problems were surveyed and evaluated with respect to solar tracking mechanisms and rotary joints for coolant piping. An analytical development of celestial mechanics, one- and two-axis tracking configurations and the effect of tracking accuracy versus collector efficiency are reported. Daily operational requirements and tracking modes were defined and evaluated. A literature and patent search on solar tracking technology was performed. Tracking system and control system performance specifications were determined. Alternative conceptual tracking approaches were defined and a cost and performance evaluation of a mechanical tracking concept was performed. Fluid coupling service specifications were determined. The cost and performance of several types of actuators and error detectors were evaluated with respect to solar tracking mechanisms.

  9. Impact of mechanical- and maintenance-induced failures of main reactor coolant pump seals on plant safety

    SciTech Connect

    Azarm, M A; Boccio, J L; Mitra, S

    1985-12-01

    This document presents an investigation of the safety impact resulting from mechanical- and maintenance-induced reactor coolant pump (RCP) seal failures in nuclear power plants. A data survey of the pump seal failures for existing nuclear power plants in the US from several available sources was performed. The annual frequency of pump seal failures in a nuclear power plant was estimated based on the concept of hazard rate and dependency evaluation. The conditional probability of various sizes of leak rates given seal failures was then evaluated. The safety impact of RCP seal failures, in terms of contribution to plant core-melt frequency, was also evaluated for three nuclear power plants. For leak rates below the normal makeup capacity and the impact of plant safety were discussed qualitatively, whereas for leak rates beyond the normal make up capacity, formal PRA methodologies were applied. 22 refs., 17 figs., 19 tabs.

  10. JAEA Studies on High Burnup Fuel Behaviors during Reactivity-Initiated Accident and Loss-of-Coolant Accident

    SciTech Connect

    Fuketa, Toyoshi; Sugiyama, Tomoyuki; Nagase, Fumihisa; Suzuki, Motoe

    2007-07-01

    The objectives of fuel safety research program at Japan Atomic Energy Agency (JAEA) are; to evaluate adequacy of present safety criteria and safety margins; to provide a database for future regulation on higher burnup UO{sub 2} and MOX fuels, new cladding and pellets; and to provide reasonably mechanistic computer codes for regulatory application. The JAEA program is comprised of reactivity-initiated accident (RIA) studies including pulse-irradiation experiments in the NSRR and cladding mechanical tests, loss-of-coolant accident (LOCA) tests including integral thermal shock test and oxidation rate measurement, development and verification of computer codes FEMAXI-6 and RANNS, and so on. In addition to an overview of the fuel safety research at JAEA, most recent progresses in the RIA and LOCA tests programs and the codes development are described and discussed in the paper. (authors)

  11. Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

    SciTech Connect

    Roth, P.A.; Schultz, R.R. ); Choi, C.J. )

    1992-07-01

    Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests.

  12. Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

    SciTech Connect

    Roth, P.A.; Schultz, R.R.; Choi, C.J.

    1992-07-01

    Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d`Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d`Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests.

  13. CFD Analysis of Coolant Flow in VVER-440 Fuel Assemblies with the Code ANSYS CFX 10.0

    SciTech Connect

    Toth, Sandor; Legradi, Gabor; Aszodi, Attila

    2006-07-01

    From the aspect of planning the power upgrading of nuclear reactors - including the VVER-440 type reactor - it is essential to get to know the flow field in the fuel assembly. For this purpose we have developed models of the fuel assembly of the VVER-440 reactor using the ANSYS CFX 10.0 CFD code. At first a 240 mm long part of a 60 degrees segment of the fuel pin bundle was modelled. Implementing this model a sensitivity study on the appropriate meshing was performed. Based on the development of the above described model, further models were developed: a 960 mm long part of a 60-degree-segment and a full length part (2420 mm) of the fuel pin bundle segment. The calculations were run using constant coolant properties and several turbulence models. The impacts of choosing different turbulence models were investigated. The results of the above-mentioned investigations are presented in this paper. (authors)

  14. High Pressure Coolant Injection (HPCI) System Risk-Based Inspection Guide for Browns Ferry Nuclear Power Station

    SciTech Connect

    Wong, S.; DiBiasio, A.; Gunther, W.

    1993-09-01

    The High Pressure Coolant Injection (HPCI) system has been examined from a risk perspective. A System Risk-Based Inspection Guide (S-RIG) has been developed as an aid to HPCI system inspections at the Browns Ferry Nuclear Power Plant, Units 1, 2 and 3. The role of. the HPCI system in mitigating accidents is discussed in this S-RIG, along with insights on identified risk-based failure modes which could prevent proper operation of the system. The S-RIG provides a review of industry-wide operating experience, including plant-specific illustrative examples to augment the PRA and operational considerations in identifying a catalogue of basic PRA failure modes for the HPCI system. It is designed to be used as a reference for routine inspections, self-initiated safety system functional inspections (SSFIs), and the evaluation of risk significance of component failures at the nuclear power plant.

  15. Development of a reactor-coolant pump monitoring and diagnostic system. Project management report, October 1980-November 1981

    SciTech Connect

    Sommerfield, G. A.; Morris, D. J.

    1982-02-01

    The objective of the project is to develop a reactor coolant pump monitoring and diagnostic system and collect sufficient data to permit analysts to determine why high outleakage and failures occur at Davis-Besse Nuclear Power Station, Unit 1. This report summarizes the work completed during the first period of performance, October 1, 1980, through November 30, 1981. During this period the members of the Project Team were selected and the detailed work management plans developed to take this project from conception, through detailed engineering, and finally to construction in the early part of 1982. The scope of work described includes system design, equipment selection for the computer-based data collection and diagnostic system, and computer software development to permit data collection and analysis.

  16. Technical report on material selection and processing guidelines for BWR coolant pressure boundary piping. Draft report. Revision 2

    SciTech Connect

    Hazelton, W.S.

    1986-06-01

    This report updates and supersedes the technical positions NRC established in NUREG-0313, ''Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,'' published in July 1977, and its subsequent revision published in July 1980. This report sets forth the NRC staff's revised acceptable methods to control the intergranular stress corrosion cracking susceptibility of BWR ASME Code Class 1, 2, and 3 pressure boundary piping and safe ends. For piping that does not fully comply with the material selection, testing, and processing guideline combinations of this document, varying degrees of augmented inservice inspection will be required, pursuant to 10 CFR 50.55a(g)(6)(ii). This revision also includes guidance regarding crack evaluation and weld overlay repair methods for long term operation or for continuing interim operation of plants until a more permanent solution is implemented.

  17. Simulations and field tests of a reactor coolant pump emergency start-up by means of remote gas units

    SciTech Connect

    Omahen, P.; Gubina, F. )

    1992-12-01

    The problem of the reactor coolant pump start-up in case of emergency by means of remote gas power plant units was analyzed. In this paper a simulation model is developed which enabled a detailed simulation of the transient process occurring at the start-up. The start-up of the RCP motor set was simulated in case of available one and two gas units. The field tests were performed and the measured variable values complied well with the simulation results. Two gas units have been determined as a safe start-up scheme of the RCP motor set considering for safety reasons accepted busbars and motor protection settings. A derived model for deep rotor bars was experimentally confirmed as effective means for the RCP motor set start-up transient simulation. Start-up procedures have been designed and adopted to the safety procedures of the Nuclear Power Plant Krsko.

  18. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    SciTech Connect

    Heams, T J; Williams, D A; Johns, N A; Mason, A; Bixler, N E; Grimley, A J; Wheatley, C J; Dickson, L W; Osborn-Lee, I; Domagala, P; Zawadzki, S; Rest, J; Alexander, C A; Lee, R Y

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

  19. ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) simulation of a loss of coolant accident in a space reactor

    SciTech Connect

    Roth, P.A.; Shumway, R.W.

    1988-01-01

    The Advanced Thermal Hydraulic Energy Network Analyzer (ATHENA) code was used to simulate a loss-of-coolant accident (LOCA) in a conceptual space reactor design. ATHENA provides the capability of simulating the thermal-hydraulic behavior of the wide variety of systems which are being considered for use in space reactors. Flow loops containing any one of several available working fluids may interact through thermal connections with other loops containing the same or a different working fluid. The code can be used to model special systems such as: heat pipes, point reactor kinetics, plant control systems, turbines, valves, and pumps. This work demonstrates the application of the thermal radiation model which has been recently incorporated into ATHENA and verifies the need for supplemental reactor cooling to prevent reactor fuel damage in the event of a LOCA.

  20. Dose to man from a hypothetical loss-of-coolant accident at the Rancho Seco Nuclear Power Plant

    SciTech Connect

    Peterson, K.R.; Greenly, G.D.

    1981-02-01

    At the request of the Sacramento Municipal Utilities District, we used our computer codes, MATHEW and ADPIC, to assess the environmental impact of a loss-of-coolant accident at the Rancho Seco Nuclear Power Plant, about 40 kilometres southeast of Sacramento, California. Meteorological input was selected so that the effluent released by the accident would be transported over the Sacramento metropolitan area. With the release rates provided by the Sacramento Municipal Utilities District, we calculated the largest total dose for a 24-hour release as 70 rem about one kilometre northwest of the reactor. The largest total dose in the Sacramento metropolitan area is 780 millirem. Both doses are from iodine-131, via the forage-cow-milk pathway to an infant's thyroid. The largest dose near the nuclear plant can be minimized by replacing contaminated milk and by giving the cows dry feed. To our knowledge, there are no milk cows within the Sacramento metropolitan area.

  1. Experimental investigation on the chemical precipitation generation under the loss of coolant accident of nuclear power plants

    SciTech Connect

    Kim, C. H.; Sung, J. J.; Chung, Y. W.

    2012-07-01

    The PWR containment buildings are designed to facilitate core cooling in the event of a Loss of Coolant Accident (LOCA). The cooling process requires water discharged from the break and containment spray to be collected in a sump for recirculation. The containment sump contains screens to protect the components of the Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) from debris. Since the containment materials may dissolve or corrode when exposed to the reactor coolant and spray solutions, various chemical precipitations can be generated in a post-LOCA environment. These chemical precipitations may become another source of debris loading to be considered in sump screen performance and downstream effects. In this study, new experimental methodology to predict the type and quantity of chemical precipitations has been developed. To generate the plant-specific chemical precipitation in a post-LOCA environment, the plant specific chemical condition of the recirculation sump during post-LOCA is simulated with the experimental reactor for the chemical effect. The plant-specific containment materials are used in the present experiment such as glass fibers, concrete blocks, aluminum specimens, and chemical reagent - boric acid, spray additives or buffering chemicals (sodium hydroxide, Tri-Sodium Phosphate (TSP), or others). The inside temperature of the reactor is controlled to simulate the plant-specific temperature profile of the recirculation sump. The total amount of aluminum released from aluminum specimens is evaluated by ICP-AES analysis to determine the amount of AlOOH and NaAlSi{sub 3}O{sub 8} which induce very adverse effect on the head loss across the sump screens. The amount of these precipitations generated in the present experimental study is compared with the results of WCAP-16530-NP-A. (authors)

  2. An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant

    SciTech Connect

    Marshall, T.D.; Watson, R.D.; McDonald, J.M.

    1995-09-01

    In an off-normal event, water-cooled copper divertor plates in the International Thermonuclear Experimental Reactor (ITER) may either experience heat loads beyond their design basis, or the normal heat loads may be accompanied by low coolant pressure and velocity. The purpose of this experiment was to illustrate that during one-sided heating, as in ITER, a copper divertor plate with the proper side wall thickness, at low system pressure and velocity can absorb without failing an incident heat flux, q{sub i}, that significantly exceed the value, q{sub i}{sup CHF}, which is associated with local CHF at the wall of the coolant channel. The experiment was performed using a 30 kW electron beam test system for heating of a square cross-section divertor heat sink with a smooth circular channel of 7.63 mm diameter. The heated width, length, and wall thickness were 16, 40, and 3 mm, respectively. Stable surface temperatures were observed at incident heat fluxes greater than the local CHF point, presumably due to circumferential conduction around the thick tube walls when q{sub i}{sup CHF} was exceeded. The Post-CHF enhancement factor, {eta}, is defined as the ratio of the incident burnout heat flux, q{sub i}{sup BO}, to q{sub i}{sup CHF}. For this experiment with water at inlet conditions of 70{degrees}C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1100 W/cm{sup 2}, respectively, which gave an {eta} of 1.8.

  3. The thermal process diagram and equipment of the secondary coolant circuit of a nuclear power station unit based on the BREST-OD-300 reactor installation for subcritical steam conditions

    NASA Astrophysics Data System (ADS)

    Nesterov, Yu. V.; Lisyanskii, A. S.; Makarova, E. I.; Bal'Va, L. Ya.; Prikhod'Ko, P. Yu.

    2011-06-01

    The 300-MWe power unit based on an experimental-demonstration two-circuit 700-MWt reactor installation with lead coolant is briefly described. The thermal process diagram of the secondary coolant circuit for the subcritical steam conditions 17 MPa and 505°C at the outlet from steam generators is presented.

  4. TACT1, a computer program for the transient thermal analysis of a cooled turbine blade or vane equipped with a coolant insert. 1. Users manual

    NASA Technical Reports Server (NTRS)

    Gaugler, R. E.

    1978-01-01

    A computer program to calculate transient and steady state temperatures, pressures, and coolant flows in a cooled, axial flow turbine blade or vane with an impingement insert is described. Coolant side heat transfer coefficients are calculated internally in the program, with the user specifying either impingement or convection heat transfer at each internal flow station. Spent impingement air flows in a chordwise direction and is discharged through the trailing edge and through film cooling holes. The ability of the program to handle film cooling is limited by the internal flow model. Sample problems, with tables of input and output, are included in the report. Input to the program includes a description of the blade geometry, coolant supply conditions, outside thermal boundary conditions, and wheel speed. The blade wall can have two layers of different materials, such as a ceramic thermal barrier coating over a metallic substrate. Program output includes the temperature at each node, the coolant pressures and flow rates, and the inside heat-transfer coefficients.

  5. Modeling of Coolant Flow in the Fuel Assembly of the Reactor of a Floating Nuclear Power Plant Using the Logos CFD Program

    NASA Astrophysics Data System (ADS)

    Dmitriev, S. M.; Dobrov, A. A.; Legchanov, M. A.; Khrobostov, A. E.

    2015-09-01

    Results of computer modeling of coolant flow in the fuel assembly of the reactor of a floating nuclear power plant using the LOGOS CFD programs have been given. The possibility of using the obtained results to improve models built into the engineering programs of thermohydraulic calculation of nuclear-reactor cores has been considered.

  6. Analysis of an Air Conditioning Coolant Solution for Metal Contamination Using Atomic Absorption Spectroscopy: An Undergraduate Instrumental Analysis Exercise Simulating an Industrial Assignment

    ERIC Educational Resources Information Center

    Baird, Michael J.

    2004-01-01

    A real-life analytical assignment is presented to students, who had to examine an air conditioning coolant solution for metal contamination using an atomic absorption spectroscopy (AAS). This hands-on access to a real problem exposed the undergraduate students to the mechanism of AAS, and promoted participation in a simulated industrial activity.

  7. Assessment and Accommodation of Thermal Expansion of the Internal Active Thermal Control System Coolant During Launch to On-Orbit Activation of International Space Station Elements

    NASA Technical Reports Server (NTRS)

    Edwards, J. Darryl; Ungar, Eugene K.; Holt, James M.; Turner, Larry D. (Technical Monitor)

    2001-01-01

    The International Space Station (ISS) employs an Internal Active Thermal Control System (IATCS) comprised of several single-phase water coolant loops. These coolant loops are distributed throughout the ISS pressurized elements. The primary element coolant loops (i.e., US Laboratory module) contain a fluid accumulator to accommodate thermal expansion of the system. Other element coolant loops are parasitic (i.e., Airlock), have no accumulator, and require an alternative approach to insure that the system Maximum Design Pressure (MDP) is not exceeded during the Launch to Activation phase. During this time the element loop is a stand alone closed individual system. The solution approach for accommodating thermal expansion was affected by interactions of system components and their particular limitations. The mathematical solution approach was challenged by the presence of certain unknown or not readily obtainable physical and thermodynamic characteristics of some system components and processes. The purpose of this paper is to provide a brief description of a few of the solutions that evolved over time, a novel mathematical solution to eliminate some of the unknowns or derive the unknowns experimentally, and the testing and methods undertaken.

  8. Assessment and Accommodation of Thermal Expansion of the Internal Active Thermal Control System Coolant During Launch to On-Orbit Activation of International Space Station Elements

    NASA Technical Reports Server (NTRS)

    Edwards, Darryl; Ungar, Eugene K.; Holt, James M.

    2002-01-01

    The International Space Station (ISS) employs an Internal Active Thermal Control System (IATCS) comprised of several single-phase water coolant loops. These coolant loops are distributed throughout the ISS pressurized elements. The primary element coolant loops (i.e. U.S. Laboratory module) contain a fluid accumulator to accomodate thermal expansion of the system. Other element coolant loops are parasitic (i.e. Airlock), have no accumulator, and require an alternative approach to insure that the system maximum design pressure (MDP) is not exceeded during the Launch to Activation (LTA) phase. During this time the element loops is a stand alone closed system. The solution approach for accomodating thermal expansion was affected by interactions of system components and their particular limitations. The mathematical solution approach was challenged by the presence of certain unknown or not readily obtainable physical and thermodynamic characteristics of some system components and processes. The purpose of this paper is to provide a brief description of a few of the solutions that evolved over time, a novel mathematical solution to eliminate some of the unknowns or derive the unknowns experimentally, and the testing and methods undertaken.

  9. The Effect of Saline Coolant on Temperature Levels during Decortication with a Midas Rex: An in Vitro Model Using Sheep Cervical Vertebrae.

    PubMed

    Livingston, Asher; Wang, Tian; Christou, Chris; Pelletier, Matthew H; Walsh, William R

    2015-01-01

    Decortication of bone with a high-speed burr in the absence of coolant may lead to local thermal necrosis and decreased healing ability, which may negatively impact clinical outcome. Little data are available on the impact of applying a coolant during the burring process. This study aims to establish an in vitro model to quantitatively assess peak temperatures during endplate preparation with a high-speed burr. Six sheep cervical vertebrae were dissected and mounted. Both end plates were used to give a total of 12 sites. Two thermocouples were inserted into each vertebra, 2 mm below the end plate surface and a thermal camera set up to measure surface temperature. A 3 mm high-pneumatic speed burr (Midas Rex, Medtronic, Fort Worth, TX, USA) was used to decorticate the bone in a side to side sweeping pattern, using a matchstick burr (M-8/9MH30) with light pressure. This procedure was repeated while dripping saline onto the burr and bone. Data were compared between groups using a Student's t-test. Application of coolant at the bone-burr interface during decortication resulted in a significant decrease in final temperature. Without coolant, maximum temperatures 2 mm from the surface were not sufficient to cause thermal osteonecrosis, although peak surface temperatures would cause local damage. The use of a high-speed burr provides a quick and an effective method of vertebral end plate preparation. Thermal damage to the bone can be minimized through the use of light pressure and saline coolant. This has implications for any bone preparation performed with a high-speed burr. PMID:26284253

  10. The Effect of Saline Coolant on Temperature Levels during Decortication with a Midas Rex: An in Vitro Model Using Sheep Cervical Vertebrae

    PubMed Central

    Livingston, Asher; Wang, Tian; Christou, Chris; Pelletier, Matthew H.; Walsh, William R.

    2015-01-01

    Decortication of bone with a high-speed burr in the absence of coolant may lead to local thermal necrosis and decreased healing ability, which may negatively impact clinical outcome. Little data are available on the impact of applying a coolant during the burring process. This study aims to establish an in vitro model to quantitatively assess peak temperatures during endplate preparation with a high-speed burr. Six sheep cervical vertebrae were dissected and mounted. Both end plates were used to give a total of 12 sites. Two thermocouples were inserted into each vertebra, 2 mm below the end plate surface and a thermal camera set up to measure surface temperature. A 3 mm high-pneumatic speed burr (Midas Rex, Medtronic, Fort Worth, TX, USA) was used to decorticate the bone in a side to side sweeping pattern, using a matchstick burr (M-8/9MH30) with light pressure. This procedure was repeated while dripping saline onto the burr and bone. Data were compared between groups using a Student’s t-test. Application of coolant at the bone–burr interface during decortication resulted in a significant decrease in final temperature. Without coolant, maximum temperatures 2 mm from the surface were not sufficient to cause thermal osteonecrosis, although peak surface temperatures would cause local damage. The use of a high-speed burr provides a quick and an effective method of vertebral end plate preparation. Thermal damage to the bone can be minimized through the use of light pressure and saline coolant. This has implications for any bone preparation performed with a high-speed burr. PMID:26284253

  11. An Experimental Investigation and Correlation of the Heat Reduction to Nonporous Surfaces Behind a Porous Leading Edge Through Which Coolant is Ejected

    NASA Technical Reports Server (NTRS)

    Witte, William G.; Rashis, Bernard

    1960-01-01

    A configuration of a wing segment having constant chord thickness, 0 deg. sweep, a porous steel semicircular leading edge, and solid Inconel surfaces was tested in a Mach number 2.0 ethlyene-heated high-temperature air jet. Measurements were made of the wing surface temperatures at chordwise stations for several rates of helium flow through the porous leading edge. The investigation was conducted at stagnation temperatures ranging from 500 F to 2,400 F, at Reynolds numbers per foot ranging from 0.3 x 10(exp 7) to 1.2 x 10(exp 7), and at angles of attack of 0, +/- 5, and +/- 15 deg. The results indicated that the reduction of wing surface temperatures with respect to their values for no coolant flow, depended on the helium coolant flow rates and the distance behind the area of injection. The results were correlated in terms of the wall cooling parameter and the coolant flow-rate parameter, where the nondimensional flow rate was referenced to the cooled area up to the downstream position. For the same coolant flow rate, lower surface temperatures are achieved with a porous-wall cooling system. However, since flow-rate requirements decrease with increasing allowable surface temperatures, the higher allowable wall temperatures of the solid wall as compared to the structurally weaker porous wall- sharply reduce the flow-rate requirements of a downstream cooling system. Thus, for certain flight conditions it is possible to compensate for the lower efficiency of the downstream or solid-wall cooling system. For example, a downstream cooling system using solid walls that must be maintained at 1,800 F would require less coolant for Mach numbers up to 5.5 than would a porous-wall cooling system for which the walls must be maintained at temperatures less than or equal to 9000 F.

  12. Heat transfer in a two-pass internally ribbed turbine blade coolant channel with cylindrical vortex generators

    SciTech Connect

    Hibbs, R.G.; Acharya, S.; Chen, Y.; Nikitopoulos, D.E.; Myrum, T.A.

    1998-07-01

    The effect of vortex generators on the mass (heat) transfer from the ribbed passage of a two-pass turbine blade coolant channel is investigated with the intent of optimizing the vortex generator geometry so that significant enhancements in mass/heat transfer can be achieved. In the experimental configuration considered, ribs are mounted on two opposite walls; all four walls along each pass are active and have mass transfer from their surfaces but the ribs are nonparticipating. Mass transfer measurements, in the form of Sherwood number ratios, are made along the centerline and in selected interrib modules. Results are presented for Reynolds number in the range of 5000 to 40,000 pitch to rib height ratios of 10.5 and 21, and vortex generator-rib spacing to rib height ratios of 0.55 and 1.5. Centerline and spanwise-averaged Sherwood number ratios are presented along with contours of the Sherwood number ratios. Results indicate that the vortex generators lead to substantial increases in the local mass transfer rates, particularly along the side walls, and modest increases in the average mass transfer rates. The vortex generators have the effect of making the interrib profiles along the ribbed walls more uniform. Along the side walls, vortices that characterize the vortex generator wake are associated with significant mass transfer enhancements. The wake effects and the levels of enhancement decrease somewhat with increasing Reynolds number and decreasing pitch.

  13. The Effect of the UO{sub 2}/ZrO{sub 2} Composition on Fuel/Coolant Interaction

    SciTech Connect

    Song, Jin Ho; Kim, Jong Hwan

    2005-03-15

    A series of experiments on fuel/coolant interaction (FCI) was performed in the TROI facility, where the composition of the mixture was varied. The compositions of the UO{sub 2} and ZrO{sub 2} mixture in weight percent were 50:50, 70:30, 80:20, and pure ZrO{sub 2}. The responses of the system including the temperature of the pool of water in the test vessel, pressure and temperature of the containment vessel, and dynamic pressures and force were measured. In addition, high-speed movies were taken through the windows. The tests using corium with a 70:30 composition and pure zirconia resulted in a spontaneous energetic steam explosion, while the tests with other compositions did not lead to an energetic FCI. The debris size distribution and pressure and temperature responses clearly indicated the cases with an energetic explosion and the cases without an explosion. The high-speed movie taken during the FCI through the visible window clearly disclosed the outstanding phases of the FCI, which were the melt entry phase, the triggering phase, and the continued melt jet and expansion of the mixing zone phase.

  14. Analysis of Kuosheng Large-Break Loss-of-Coolant Accident with MELCOR 1.8.4

    SciTech Connect

    Wang, T.-C.; Wang, S.-J.; Chien, C.-S

    2000-09-15

    The MELCOR code, developed by Sandia National Laboratories, is capable of simulating the severe accident phenomena of light water reactor nuclear power plants (NPPs). A specific large-break loss-of-coolant accident (LOCA) for Kuosheng NPP is simulated with the use of the MELCOR 1.8.4 code. This accident is induced by a double-ended guillotine break of one of the recirculation pipes concurrent with complete failure of the emergency core cooling system. The MELCOR input deck for the Kuosheng NPP is established based on the design data of the Kuosheng NPP and the MELCOR users' guides. The initial steady-state conditions are generated with a developed self-initialization algorithm. The effect of the MELCOR 1.8.4-provided initialization process is demonstrated. The main severe accident phenomena and the corresponding fission product released fractions associated with the large-break LOCA sequences are simulated. The MELCOR 1.8.4 predicts a longer time interval between the core collapse and vessel failure and a higher source term. This MELCOR 1.8.4 input deck will be applied to the probabilistic risk assessment, the severe accident analysis, and the severe accident management study of the Kuosheng NPP in the near future.

  15. Proposed damage evolution model for large-scale finite element modeling of the dual coolant US-ITER TBM

    NASA Astrophysics Data System (ADS)

    Sharafat, S.; El-Awady, J.; Liu, S.; Diegele, E.; Ghoniem, N. M.

    2007-08-01

    Large-scale finite element modeling (FEM) of the US Dual Coolant Lead Lithium ITER Test Blanket Module including damage evolution is under development. A comprehensive rate-theory based radiation damage creep deformation code was integrated with the ABACUS FEM code. The advantage of this approach is that time-dependent in-reactor deformations and radiation damage can now be directly coupled with 'material properties' of FEM analyses. The coupled FEM-Creep damage model successfully simulated the simultaneous microstructure and stress evolution in small tensile test-bar structures. Applying the integrated Creep/FEM code to large structures is still computationally prohibitive. Instead, for thermo-structural analysis of the DCLL TBM structure the integrated FEM-creep damage model was used to develop true stress-strain behavior of F82H ferritic steel. Based on this integrated damage evolution-FEM approach it is proposed to use large-scale FEM analysis to identify and isolate critical stress areas for follow up analysis using detailed and fully integrated creep-FEM approach.

  16. Evaluation of effects of LWR coolant environments on fatigue life of carbon and low-alloy steels

    SciTech Connect

    Chopra, O.K.; Shack, W.J.

    1996-02-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figure I-90 of Appendix I to Section III of the Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Recent test data indicate a significant decrease in fatigue life of carbon and low-alloy steels in LWR environments when five conditions are satisfied simultaneously, viz., applied strain range, temperature, dissolved oxygen in the water, and sulfur content of the steel are above a minimum threshold level, and the loading strain rate is below a threshold value. Only a moderate decrease in fatigue life is observed when any one of these conditions is not satisfied. This paper summarizes available data on the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, and sulfur content on the fatigue life of carbon and low-alloy steels. The data have been analyzed to define the threshold values of the five critical parameters. Methods for estimating fatigue lives under actual loading histories are discussed.

  17. A fracture mechanics approach for estimating fatigue crack initiation in carbon and low-alloy steels in LWR coolant environments

    SciTech Connect

    Park, H. B.; Chopra, O. K.

    2000-04-10

    A fracture mechanics approach for elastic-plastic materials has been used to evaluate the effects of light water reactor (LWR) coolant environments on the fatigue lives of carbon and low-alloy steels. The fatigue life of such steel, defined as the number of cycles required to form an engineering-size crack, i.e., 3-mm deep, is considered to be composed of the growth of (a) microstructurally small cracks and (b) mechanically small cracks. The growth of the latter was characterized in terms of {Delta}J and crack growth rate (da/dN) data in air and LWR environments; in water, the growth rates from long crack tests had to be decreased to match the rates from fatigue S-N data. The growth of microstructurally small cracks was expressed by a modified Hobson relationship in air and by a slip dissolution/oxidation model in water. The crack length for transition from a microstructurally small crack to a mechanically small crack was based on studies on small crack growth. The estimated fatigue S-N curves show good agreement with the experimental data for these steels in air and water environments. At low strain amplitudes, the predicted lives in water can be significantly lower than the experimental values.

  18. The effect of coolant orificing on the core performance of a heterogeneous liquid-metal fast breeder reactor

    SciTech Connect

    Mamoru, K.; Shigehiro, A.; Yoshiaki, O.

    1983-04-01

    The effect of orificing on the core performance of a commercial-size heterogeneous liquid-metal fast breeder reactor was studied analytically. The thermal power output was flattened at beginning of life, and the coolant flow rate was chosen such that the maximum inner cladding temperature of a driver fuel and a blanket fuel was less than or equal to 620/sup 0/C at both beginning of equilibrium life (BOEL) and end of equilibrium life (EOEL). The difference between reactor outlet temperatures at BOEL and EOEL was then calculated for six core configurations: one homogeneous core configuration and five heterogeneous ones. The results showed that the core outlet temperature variation due to the change of the power profile of the radial heterogeneous core configurations is similar to that of the homogeneous one, even when a single type of orificing is used in each core zone, and it will not be necessary to use the more detailed orificing in each zone of a heterogeneous core configuration. The study concludes that for the present design, especially the thermal design, of some heterogeneous core configurations, it is feasible to control the change of the reactor outlet temperature with burnup, even when a single type of orificing is used in each core zone.

  19. Effect of surface oxidation on the onset of nucleate boiling in a materials test reactor coolant channel

    DOE PAGESBeta

    Forrest, Eric C.; Don, Sarah M.; Hu, Lin -Wen; Buongiorno, Jacopo; McKrell, Thomas J.

    2016-02-29

    The onset of nucleate boiling (ONB) serves as the thermal-hydraulic operating limit for many research and test reactors. However, boiling incipience under forced convection has not been well-characterized in narrow channel geometries or for oxidized surface conditions. This study presents experimental data for the ONB in vertical upflow of deionized (DI) water in a simulated materials test reactor (MTR) coolant channel. The channel gap thickness and aspect ratio were 1.96 mm and 29:1, respectively. Boiling surface conditions were carefully controlled and characterized, with both heavily oxidized and native oxide surfaces tested. Measurements were performed for mass fluxes ranging from 750more » to 3000 kg/m2s and for subcoolings ranging from 10 to 45°C. ONB was identified using a combination of high-speed visual observation, surface temperature measurements, and channel pressure drop measurements. Surface temperature measurements were found to be most reliable in identifying the ONB. For the nominal (native oxide) surface, results indicate that the correlation of Bergles and Rohsenow, when paired with the appropriate single-phase heat transfer correlation, adequately predicts the ONB heat flux. Furthermore, incipience on the oxidized surface occurred at a higher heat flux and superheat than on the plain surface.« less

  20. Heat transfer in a two-pass internally ribbed turbine blade coolant channel with cylindrical vortex generators

    SciTech Connect

    Hibbs, R.; Chen, Y.; Nikitopoulos, D.

    1995-10-01

    The effect of vortex generators on the mass (heat) transfer from the ribbed passage of a two pass turbine blade coolant channel is investigated with the intent of optimizing the vortex generator geometry so that significant enhancements in mass/heat transfer can be achieved. In the experimental configuration considered, ribs are mounted on two opposite walls; all four walls along each pass are active and have mass transfer from their surfaces but the ribs are non-participating. Mass transfer measurements, in the form of Sherwood number ratios, are made along the centerline and in selected inter-rib modules. Results are presented for Reynolds number in the range of 5,000 to 40,000, pitch to rib height ratios of 10.5 and 21, and vortex generator-rib spacing to rib height ratios of 0.55, and 1.5. Centerline and spanwise averaged Sherwood number ratios are presented along with contours of the Sherwood number ratios. Results indicate that the vortex generators induce substantial increases in the local mass transfer rates, particularly along the side walls, and modest increases in the average mass transfer rates. The vortex generators have the effect of making the inter-rib profiles along the ribbed walls more uniform. Along the side walls, horse-shoe vortices that characterize the vortex generator wake are associated with significant mass transfer enhancements. The wake effects and the levels of enhancement decrease somewhat with increasing Reynolds number and decreasing pitch.

  1. Application of 3-dimensional radiation transport codes to the analysis of the CRBR prototypic coolant pipe chaseway neutron streaming experiment

    SciTech Connect

    Chatani, K. )

    1992-08-01

    This report summarizes the calculational results from analyses of a Clinch River Breeder Reactor (CRBR) prototypic coolant pipe chaseway neutron streaming experiment Comparisons of calculated and measured results are presented, major emphasis being placed on results at bends in the chaseway. Calculations were performed with three three-dimensional radiation transport codes: the discrete ordinates code TORT and the Monte Carlo code MORSE, both developed by the Oak Ridge National Laboratory (ORNL), and the discrete ordinates code ENSEMBLE, developed by Japan. The calculated results from the three codes are compared (1) with previously-calculated DOT3.5 two-dimensional results, (2) among themselves, and (3) with measured results. Calculations with TORT used both the weighted-difference and nodal methods. Only the weighted-difference method was used in ENSEMBLE. When the calculated results were compared to measured results, it was found that calculation-to-experiment (C/E) ratios were good in the regions of the chaseway where two-dimensional modeling might be difficult and where there were no significant discrete ordinates ray effects. Excellent agreement was observed for responses dominated by thermal neutron contributions. MORSE-calculated results and comparisons are described also, and detailed results are presented in an appendix.

  2. Heat transfer in a two-pass internally ribbed turbine blade coolant channel with cylindrical vortex generators

    SciTech Connect

    Hibbs, R.; Acharya, S.; Chen, Y.

    1995-12-31

    The effect of vortex generators on the mass (heat) transfer from the ribbed passage of a two pass turbine blade coolant channel is investigated with the intent of optimizing the vortex generator geometry so that significant enhancements in mass/heat transfer can be achieved. In the experimental configuration considered, ribs are mounted on two opposite walls; all four walls along each pass are active and have mass transfer from their surfaces but the ribs are non-participating. Mass transfer measurements, in the form of Sherwood number ratios, are made along the centerline and in selected inter-rib modules. Results are presented for Reynolds number in the range of 5,000 to 40,000, pitch to rib height ratios of 10.5 and 21, and vortex generator-rib spacing to rib height ratios of 0.55, and 1.5. Centerline and spanwise averaged Sherwood number ratios are presented along with contours of the Sherwood number ratios. Results indicate that the vortex generators induce substantial increases in the local mass transfer rates, particularly along the side walls, and modest increases in the average mass transfer rates. The vortex generators have the effect of making the inter-rib profiles along the ribbed walls more uniform. Along the side walls, horse-shoe vortices that characterize the vortex generator wake are associated with significant mass transfer enhancements. The wake effects and the levels of enhancement decrease somewhat with increasing Reynolds number and decreasing pitch.

  3. Regulatory instrument review: Aging management of LWR cables, containment and basemat, reactor coolant pumps, and motor-operated valves

    SciTech Connect

    Werry, E.V.; Somasundaram, S.

    1995-09-01

    The results of Stage 2 of the Regulatory Instrument Review are presented in this volume. Selected regulatory instruments, such as the Code of Federal Regulations (CFR), US Nuclear Regulatory Commission (NRC), Regulatory Guides, and ASME Codes, were investigated to determine the extent to which these regulations apply aging management to selected safety-related components in nuclear power plants. The Regulatory Instrument Review was funded by the NRC under the Nuclear Plant Aging Research (NPAR) program. Stage 2 of the review focused on four safety-related structures and components; namely, cables, containment and basemat, reactor coolant pumps, and motor-operated valves. The review suggests that the primary-emphasis of the regulatory instruments was on the design, construction, start-up, and operation of a nuclear power plant, and that aging issues were primarily addressed after an aging-related problem was recognized. This Stage 2 review confirms the results of the prior review; (see Regulatory Instrument Review: Management of Aging of LWR Major Safety-Related Components NUREG/CR-5490. The observations indicate that the regulations generally address management of age-related degradation indirectly. Specific age-related degradation phenomena frequently are dealt with in bulletins and notices or through generic issues, letters, etc. The major recommendation of this report, therefore, is that the regulatory instruments should more directly and explicitly address the aging phenomenon and the management of the age-related degradation process.

  4. Probability of pipe failure in the reactor coolant loops of Westinghouse PWR plants. Volume 4. Pipe failure induced by crack growth in west coast plants

    SciTech Connect

    Chinn, D.J.; Holman, G.S.; Lo, T.Y.; Mensing, R.W.

    1985-07-01

    The US Nuclear Regulatory Commission contracted with the Lawrence Livermore National Laboratory to conduct a study to determine if the probability of occurrence of a double-ended guillotine break (DEGB) in primary coolant piping warrants the current design requirements that safeguard against the effecs of such a break. This report assesses the reactor-coolant-loop piping system of west coast Westinghouse plants. The results indicate that directly induced DEGB is an unlikely event in the west coast Westinghouse plants. For the Trojan plant, leak is far more likely than a direct DEGB. Further, earthquakes have very little effect on the probabilities of leak and direct DEGB. At the Diablo Canyon plant, the increase in postulated seismic levels due to reevaluation of the site to account for the Hosgri Fault has caused directly induced DEGB failure probability to be dependent on earthquake occurrences. The resulting direct DEGB failure probability is still much lower than the indirect DEGB failure probability for Diablo Canyon.

  5. Probability of pipe failure in the reactor coolant loops of Combustion Engineering PWR plants. Volume 2. Pipe failure induced by crack growth

    SciTech Connect

    Lo, T.Y.; Mensing, R.W.; Woo, H.H.; Holman, G.S.

    1984-09-01

    A study was conducted to determine if the probability of occurrence of a double-ended guillotine break (DEGB) in the primary coolant piping warrants the current design requirements that safeguard against the effect of DEGB. This report describes the results of an assessment of reactor coolant loop piping systems designed by Combustion Engineering, Inc. A probabilistic fracture mechanics approach was used to estimate the crack growth and to assess the crack stability in the piping throughout the lifetime of the plant. The results of the assessment indicate that the probability of occurrence of DEGB due to crack growth and instability is extremely small, which supports the argument that the postulation of DEGB in design should be eliminated and replaced with more reasonable criteria.

  6. Seismic Safety Margins Research Programs. Assessment of potential increases in risk due to degradation of steam generator and reactor coolant pump supports. [PWR

    SciTech Connect

    Bohn, M. P.; Wells, J. E.; Shieh, L. C.; Cover, L. E.; Streit, R. L.

    1983-08-01

    During the NRC licensing review for the North Anna Units 1 and 2 pressurized-water reactors (PWRs), questions were raised regarding the potential for low-fracture toughness of steam-generator and reactor-coolant-pump supports. Because other PWRs may face similar problems, this issue was incorporated into the NRC Program for Resolution of Generic Issues. The work described in this report was performed to provide the NRC with a quantitative evaluation of the value/impact implications of the various options of resolving the fracture-toughness question. This report presents an assessment of the probabilistic risk associated with nil-ductility failures of steam-generator and reactor-coolant-pump structural-support systems during seismic events, performed using the Seismic Safety Margins Research Program codes and data bases.

  7. Evolution of a Two-Layer Oxide Coating on the Steel Surface of the Primary Coolant Circuit in the Course of Nuclear Power Plant Operation

    NASA Astrophysics Data System (ADS)

    Alekseev, V. V.; Orlova, E. A.; Kozlov, F. A.; Varseev, E. V.

    2016-01-01

    An analysis of the laws governing mass transfer in a uniform oxide coating on the surface of a circuit is presented. As a result of calculations, the distribution of the fluxes of magnetite and of the particles of corrosion product suspensions depositing on the surface of the throughput section, as well as the distribution of the fluxes of dissolved iron, emerging from steel, along the length of the hydraulic system of the primary coolant circuit of the BREST-type plant, have been refined. The laws governing the process of mass transfer in the BREST-type primary coolant circuit with account for the oxide coating thickness variable in time have been obtained. The distribution of the thicknesses of the magnetite and spinel layers on the steel surface along the length of the BREST-type circulation loop after 365 days from the start of operation in a nominal regime is shown.

  8. PARTICLE IMAGE VELOCIMETRY MEASUREMENTS IN A REPRESENTATIVE GAS-COOLED PRISMATIC REACTOR CORE MODEL: FLOW IN THE COOLANT CHANNELS AND INTERSTITIAL BYPASS GAPS

    SciTech Connect

    Thomas E. Conder; Richard Skifton; Ralph Budwig

    2012-11-01

    Core bypass flow is one of the key issues with the prismatic Gas Turbine-Modular Helium Reactor, and it refers to the coolant that navigates through the interstitial, non-cooling passages between the graphite fuel blocks instead of traveling through the designated coolant channels. To determine the bypass flow, a double scale representative model was manufactured and installed in the Matched Index-of-Refraction flow facility; after which, stereo Particle Image Velocimetry (PIV) was employed to measure the flow field within. PIV images were analyzed to produce vector maps, and flow rates were calculated by numerically integrating over the velocity field. It was found that the bypass flow varied between 6.9-15.8% for channel Reynolds numbers of 1,746 and 4,618. The results were compared to computational fluid dynamic (CFD) pre-test simulations. When compared to these pretest calculations, the CFD analysis appeared to under predict the flow through the gap.

  9. Experimental study of the interaction of pulsations of the neutron flux and the coolant flow in a boiling-water reactor

    SciTech Connect

    Leppik, P.A.

    1984-12-01

    This paper presents results of a study designed to confirm that the interaction of the neutron flux and the coolant flow plays an important role in the mechanism of high-frequency (HF) resonant instability of the VK-50 boiling water reactor. To do this and to check the working model, signals from probes measuring the flow rate of the coolant and the neutron flux were recorded simultaneously (with the help of a magnetograph) in experiments performed in 1981 on driving the VK-50 reactor into the HF reonant instability regimes. Estimates were then obtained for the statistical characteristics of the pulsations of the flow rate and of the neutron flux, including the cross-correlation functions and coherence functions. The basic results of these studies are reported here.

  10. Blade-to-coolant heat-transfer results and operating data from a natural-convection water-cooled single-stage turbine

    NASA Technical Reports Server (NTRS)

    Diaguila, Anthony J; Freche, John C

    1951-01-01

    Blade-to-coolant heat-transfer data and operating data were obtained with a natural-convection water-cooled turbine over range of turbine speeds and inlet-gas temperatures. The convective coefficients were correlated by the general relation for natural-convection heat transfer. The turbine data were displaced from a theoretical equation for natural convection heat transfer in the turbulent region and from natural-convection data obtained with vertical cylinders and plates; possible disruption of natural convection circulation within the blade coolant passages was thus indicated. Comparison of non dimensional temperature-ratio parameters for the blade leading edge, midchord, and trailing edge indicated that the blade cooling effectiveness is greatest at the midchord and least at the trailing edge.

  11. Experimental studies of local coolant hydrodynamics using a scaled model of cassette-type fuel assembly of a KLT-40S reactor

    NASA Astrophysics Data System (ADS)

    Dmitriev, S. M.; Barinov, A. A.; Varentsov, A. V.; Doronkov, D. V.; Solntsev, D. N.; Khrobostov, A. E.

    2016-08-01

    The results of experimental studies of local hydrodynamic and mass exchange characteristics of the coolant flow behind the spacer grid in the fuel assembly of a KLT-40S reactor are presented. The experiments were aimed at the investigation of representative domains of the fuel assembly with three tracer injection regions. The studies were performed at the aerodynamic test facility using the tracer gas diffusion method. According to the theory of hydrodynamic similarity, the obtained experimental results can be transferred to full-scale coolant flow conditions in standard fuel assemblies. The analysis of the tracer concentration propagation made it possible to determine in detail the flow pattern and find the main regularities and specific features of the coolant flow behind the plate spacer grid of KLT-40S fuel assembly. The hydraulic resistance coefficient of the spacer grid was experimentally determined. The coefficients of mass exchange between cells for representative cells of the displacer region in the KLT-40S fuel assembly were calculated for the first time; these results are presented in the form of the "mixing matrix." The results of studies of local coolant flow hydrodynamics in the KLT-40S fuel assembly are used at AO Afrikantov OKBM for estimation of thermotechnical reliability of active cores for reactors of floating nuclear power stations. The experimental data on hydrodynamic and mass exchange characteristics are included in the database for verification of CDF codes and detailed cell-wise calculation of the active core for KLT-40S reactor installation. The results of these studies can be used at FSUE RFNC-VNIIEF for testing and verification of domestic three-dimensional hydrodynamic CFD codes ("Logos") that are applied for substantiation of newly designed reactor installations. Practical recommendations on the application of the obtained results in thermohydraulic calculations of the active core for the KLT-40S reactor will be worked out. Proposals

  12. Probability of failure in BWR (Boiling Water Reactor) reactor coolant piping: Volume 2, Pipe failure induced by crack growth and failure of intermediate supports

    SciTech Connect

    Lo, T.; Bumpus, S.E.; Chinn, D.J.; Mensing, R.W.; Holman, G.S.

    1989-03-01

    The US Nuclear Regulatory Commission (NRC) contracted with the Lawrence Livermore National Laboratory (LLNL) to conduct a study to determine if the probability of occurrence of a double-ended guillotine break (DEGB) in the major coolant piping systems of nuclear power plants is large enough to warrant the current stringent design requirements of designing against the postulated effects of a DEGB. The study includes both the PWR (Pressurized Water Reactor) and the BWR (Boiling Water Reactor) plants in the United States. Following the study of PWR plants, a study of BWR reactor coolant piping was performed. The Brunswick Steam Electric Plant at Southport, North Carolina was selected as the pilot plant for the BWR evaluation. The probability of pipe failure in three major coolant pipings was assessed: the recirculation loops, the primary steam lines, and the main feedwater lines. In the case of recirculation loops, both the existing and a proposed replacement system were studied. A probabilistic fracture mechanics approach was used in this study to estimate the crack growth and to assess the crack stability in the piping systems throughout the lifetime of the plant. The effects of the failure of intermediate pipe supports were also examined. The results of the assessment indicated that the probability of occurrence of DEGB due to crack growth and instability is small if the problem of intergranular stress corrosion cracking (IGSCC) is resolved by the use of the replacement system. 33 refs., 41 figs., 32 tabs.

  13. Analysis of loss-of-coolant and loss-of-flow accidents in the first wall cooling system of NET/ITER

    NASA Astrophysics Data System (ADS)

    Komen, E. M. J.; Koning, H.

    1994-03-01

    This paper presents the thermal-hydraulic analysis of potential accidents in the first wall cooling system of the Next European Torus or the International Thermonuclear Experimental Reactor. Three ex-vessel loss-of-coolant accidents, two in-vessel loss-of-coolant accidents, and three loss-of-flow accidents have been analyzed using the thermal-hydraulic system analysis code RELAP5/MOD3. The analyses deal with the transient thermal-hydraulic behavior inside the cooling systems and the temperature development inside the nuclear components during these accidents. The analysis of the different accident scenarios has been performed without operation of emergency cooling systems. The results of the analyses indicate that a loss of forced coolant flow through the first wall rapidly causes dryout in the first wall cooling pipes. Following dryout, melting in the first wall starts within about 130 s in case of ongoing plasma burning. In case of large break LOCAs and ongoing plasma burning, melting in the first wall starts about 90 s after accident initiation.

  14. Analysis of loss-of-coolant accident for a fast-spectrum lithium-cooled nuclear reactor for space-power applications

    NASA Technical Reports Server (NTRS)

    Turney, G. E.; Petrik, E. J.; Kieffer, A. W.

    1972-01-01

    A two-dimensional, transient, heat-transfer analysis was made to determine the temperature response in the core of a conceptual space-power nuclear reactor following a total loss of reactor coolant. With loss of coolant from the reactor, the controlling mode of heat transfer is thermal radiation. In one of the schemes considered for removing decay heat from the core, it was assumed that the 4 pi shield which surrounds the core acts as a constant-temperature sink (temperature, 700 K) for absorption of thermal radiation from the core. Results based on this scheme of heat removal show that melting of fuel in the core is possible only when the emissivity of the heat-radiating surfaces in the core is less than about 0.40. In another scheme for removing the afterheat, the core centerline fuel pin was replaced by a redundant, constant temperature, coolant channel. Based on an emissivity of 0.20 for all material surfaces in the core, the calculated maximum fuel temperature for this scheme of heat removal was 2840 K, or about 90 K less than the melting temperature of the UN fuel.

  15. Assessment of a large break loss of coolant accident scenario requiring operator action to initiate safety injection

    SciTech Connect

    Grendys, R.C.; Nissley, M.E.; Baker, D.C.

    1996-11-01

    As part of the licensing basis for a nuclear power plant, the acceptability of the Emergency Core Cooling Systems (ECCS) following a postulated Loss-of-Coolant Accident (LOCA) as described in the Code of Federal Regulations (CFR), Title 10, Chapter 1, Part 50.46, must be verified. The LOCA analysis is performed with an acceptable ECCS Evaluation Model and results must show compliance with the 10 CFR 50.46 acceptance criteria. Westinghouse Electric Corporation performs Large and Small Break LOCA and LOCA-related analyses to support the licensing basis of various nuclear power plants and also performs evaluations against the licensing basis analyses as required. Occasionally, the need arises for the holder of an operating license of a nuclear power plant to submit a Licensee Event Report (LER) to the US Nuclear Regulatory Commission (USNRC) for any event of the type described in the Code of Federal Regulations, Title 10, Chapter 1, Part 50.73. To support the LER, a Justification for Past Operation (JPO) may be performed to assess the safety consequences and implications of the event based on previous operating conditions. This paper describes the work performed for the Large Break LOCA to assess the impact of an event discovered by Florida Power and Light and reported in LER-94-005-02. For this event, it was determined that under certain circumstances, operator action would have been required to initiate safety injection (SI), thus challenging the acceptability of the ECCS. This event was specifically addressed for the Large Break LOCA by using an advanced thermal hydraulic analysis methodology with realistic input assumptions.

  16. Fuel-Coolant-Interaction modeling and analysis work for the High Flux Isotope Reactor Safety Analysis Report

    SciTech Connect

    Taleyarkhan, R.P.; Georgevich, V.; Nestor, C.W.; Chang, S.J.; Freels, J.; Gat, U.; Lepard, B.L.; Gwaltney, R.C.; Luttrell, C.; Kirkpatrick, J.

    1993-07-01

    A brief historical background and a description of short- and long-term task plan development for effective closure of this important safety issue for the HFIR are given. Short-term aspects deal with Fuel-Coolant-Interaction (FCI) issues experimentation, modeling, and analysis for the flow-blockage-induced steam explosion events in direct support of the SAR. Long-term aspects deal with addressing FCI issues resulting from other accidents in conjunction with issues dealing with aluminum ignition, which can result in an order of magnitude increase in overall energetics. Problem formulation, modeling, and computer code simulation for the various phases of steam explosions are described. The evaluation of core melt initiation propagation, and melt superheat are described. Core melt initiation and propagation have been studied using simple conservative models as well as from modeling and analysis using RELAP5. Core debris coolability, heatup, and melting/freezing aspects have been studied by use of the two-dimensional melting/freezing analysis code 2DKO, which was also benchmarked with MELCOR code predictions. Descriptions are provided for the HM, BH, FCIMOD, and CTH computer codes that have been implemented for studying steam explosion energetics from the standpoint of evaluating bounding loads by thermodynamic models or best-estimate loads from one- and two-dimensional simulations of steam explosion energetics. Vessel failure modeling and analysis was conducted using the principles of probabilistic fracture mechanics in conjunction with ADINA code calculations. Top head bolts failure modeling has also been conducted where the failure criterion was based upon stresses in the bolts exceeding the material yield stress for a given time duration. Missile transport modeling and analysis was conducted by setting up a one-dimensional mathematical model that accounts for viscous dissipation, virtual mass effects, and material inertia.

  17. Phased Array Ultrasonic Examination of Reactor Coolant System (Carbon Steel-to-CASS) Dissimilar Metal Weld Mockup Specimen

    SciTech Connect

    Crawford, S. L.; Cinson, A. D.; Diaz, A. A.; Anderson, M. T.

    2015-11-23

    In the summer of 2009, Pacific Northwest National Laboratory (PNNL) staff traveled to the Electric Power Research Institute (EPRI) NDE Center in Charlotte, North Carolina, to conduct phased-array ultrasonic testing on a large bore, reactor coolant pump nozzle-to-safe-end mockup. This mockup was fabricated by FlawTech, Inc. and the configuration originated from the Port St. Lucie nuclear power plant. These plants are Combustion Engineering-designed reactors. This mockup consists of a carbon steel elbow with stainless steel cladding joined to a cast austenitic stainless steel (CASS) safe-end with a dissimilar metal weld and is owned by Florida Power & Light. The objective of this study, and the data acquisition exercise held at the EPRI NDE Center, were focused on evaluating the capabilities of advanced, low-frequency phased-array ultrasonic testing (PA-UT) examination techniques for detection and characterization of implanted circumferential flaws and machined reflectors in a thick-section CASS dissimilar metal weld component. This work was limited to PA-UT assessments using 500 kHz and 800 kHz probes on circumferential flaws only, and evaluated detection and characterization of these flaws and machined reflectors from the CASS safe-end side only. All data were obtained using spatially encoded, manual scanning techniques. The effects of such factors as line-scan versus raster-scan examination approaches were evaluated, and PA-UT detection and characterization performance as a function of inspection frequency/wavelength, were also assessed. A comparative assessment of the data is provided, using length-sizing root-mean-square-error and position/localization results (flaw start/stop information) as the key criteria for flaw characterization performance. In addition, flaw signal-to-noise ratio was identified as the key criterion for detection performance.

  18. Neutron Imaging Investigations of the Secondary Hydriding of Nuclear Fuel Cladding Alloys during Loss of Coolant Accidents

    NASA Astrophysics Data System (ADS)

    Grosse, M.; Roessger, C.; Stuckert, J.; Steinbrueck, M.; Kaestner, A.; Kardjilov, N.; Schillinger, B.

    The hydrogen concentration and distribution at both sides of the burst opening of cladding tubes used in three QUENCH-LOCA simulation bundle experiments were investigated by means of neutron radiography and tomography. The quantitative correlation between the total macroscopic neutron cross-section and the atomic number density ratio between hydrogen and zirconium was determined by testing calibration specimens with known hydrogen concentrations. Hydrogen enrichments located at the end of the ballooning zone of the tested tubes were detected in the inner rods of the test bundles. Nearly all of the peripheral claddings exposed to lower temperatures do not show such enrichments. This implies that under the conditions investigated a threshold temperature exists below which no hydrogen enrichments can be formed. In order to understand the hydrogen distribution a model was developed describing the processes occurring during loss of coolant accidents after rod burst. The general shape of the hydrogen distributions with a peak each side of the ballooning region is well predicted by this model whereas the absolute concentrations are underestimated compared to the results of the neutron tomography investigations. The model was also used to discuss the influence of the alloy composition on the secondary hydrogenation. Whereas the relations for the maximal hydrogen concentrations agree well for one and the same alloy, the agreement for tests with different alloys is less satisfying, showing that material parameters such as oxidation kinetics, phase transition temperature for the zirconium oxide, and yield strength and ductility at high temperature have to be taken into account to reproduce the results of neutron imaging investigations correctly.

  19. Computational fluid dynamics analyses of lateral heat conduction, coolant azimuthal mixing and heat transfer predictions in a BR2 fuel assembly geometry.

    SciTech Connect

    Tzanos, C. P.; Dionne, B.

    2011-05-23

    To support the analyses related to the conversion of the BR2 core from highly-enriched (HEU) to low-enriched (LEU) fuel, the thermal-hydraulics codes PLTEMP and RELAP-3D are used to evaluate the safety margins during steady-state operation (PLTEMP), as well as after a loss-of-flow, loss-of-pressure, or a loss of coolant event (RELAP). In the 1-D PLTEMP and RELAP simulations, conduction in the azimuthal and axial directions is not accounted. The very good thermal conductivity of the cladding and the fuel meat and significant temperature gradients in the lateral directions (axial and azimuthal directions) could lead to a heat flux distribution that is significantly different than the power distribution. To evaluate the significance of the lateral heat conduction, 3-D computational fluid dynamics (CFD) simulations, using the CFD code STAR-CD, were performed. Safety margin calculations are typically performed for a hot stripe, i.e., an azimuthal region of the fuel plates/coolant channel containing the power peak. In a RELAP model, for example, a channel between two plates could be divided into a number of RELAP channels (stripes) in the azimuthal direction. In a PLTEMP model, the effect of azimuthal power peaking could be taken into account by using engineering factors. However, if the thermal mixing in the azimuthal direction of a coolant channel is significant, a stripping approach could be overly conservative by not taking into account this mixing. STAR-CD simulations were also performed to study the thermal mixing in the coolant. Section II of this document presents the results of the analyses of the lateral heat conduction and azimuthal thermal mixing in a coolant channel. Finally, PLTEMP and RELAP simulations rely on the use of correlations to determine heat transfer coefficients. Previous analyses showed that the Dittus-Boelter correlation gives significantly more conservative (lower) predictions than the correlations of Sieder-Tate and Petukhov. STAR-CD 3-D

  20. Vapor pressures of mixtures of CFC-114 with the potential replacement coolants C{sub 4}F{sub 10} and c-C{sub 4}F{sub 8}

    SciTech Connect

    Trowbridge, L.D.; Otey, M.G.

    1994-09-01

    The U.S. Enrichment Corporation`s production of isotopically enriched uranium depends solely on two plants which utilize the gaseous diffusion process. This process uses large quantities of CFC-114 as an evaporative coolant. CFC-114, however, will be phased out of production at the end of 1995 due to its potential to deplete stratospheric ozone. A search has been underway for substitutes for a number of years. The initial search (1988-89) for an ozone-friendly, commercially available, chemically compatible substitute yielded two candidates, FC-c318 (c-C{sub 4}F{sub 8}) and FC-3110 (C{sub 4}F{sub 10}). The intended mode of replacing coolant was to stage the new coolant into independent subsystems of the plants, so that some systems would continue to operate on CFC-114, and an increasing number would operate on the new coolant. During that changeover process, the possibility of coolant mixing arises in variety of scenarios. This work was intended to generate sufficient experimental information to be able to predict the vapor pressure of coolant mixtures over the range of operating conditions likely to be found in the diffusion plants. Specifically, vapor pressures were measured over the temperature range 322 to 355 K (120{degrees}F to 180{degrees}F) and over the full range of mole fractions for binary mixtures of CFC-114 with FC-3110, and of CFC-114 with FC-c318.

  1. Coolant controversy heats up

    SciTech Connect

    Shanley, A.

    1997-11-01

    In 1987, nations of the world banded together under the Montreal Protocol to help protect the earth`s ozone layer. Now, ten years and $2.4 billion in new fluorocarbon R and D later, the ozone hole is still with us, as a black market in illegal chlorofluorocarbons thrives and legal trade in recycled CFCs continues. Unfortunately, each alternative to CFCs poses tradeoffs. Some hydrochlorofluorocarbons present lubricant compatibility problems, and, because they are also somewhat ozone depleting, they`ll be phased out in the US between 2010 and 2030, and earlier in Europe. Chlorine-free hydrofluorocarbons contribute to global warming and some require lubricant changes and retooling, while the processes that make them emit halogenated organics. This paper discusses compatibility, flammability, global warming, and supply problems.

  2. COOLANT STABILIZED WITH KETONE

    DOEpatents

    Burr, J.G.; Strong, J.D.

    1961-11-21

    sion). U. S. Patent 3,009.88l. Nov. 2l, l06l. A method is given for inhibiting radiolytic damage to a polyphenyl composition undergoing nuclear irradiation, the polyphenyl composition initially consisting essentially of polyphenyls having 2 to 4 phenyl rings. The method comprises adding approximately 5 mole% anthraquinone to the polyphenyl composition. (AEC)

  3. REACTOR COOLANT TUBE SEAL

    DOEpatents

    Morris, W.J.

    1958-12-01

    A plle-flattenlng control element and a fluid seal therefore to permit movement of the element into a liquld contnining region of a neutronlc reactor are described. The device consists of flattened, thin-walled aluminum tubing contalnlng a uniform mixture of thermal neutron absorbing material, and a number of soft rubber closures for the process tubes, having silts capable of passing the flattened elements therethrough, but effectively sealing the process tubes against fluld leaknge by compression of the rubber. The flattened tubing is sufficiently flexible to enable it to conform to the configuratlon of the annular spacing surrounding the fuel elements ln the process tubes.

  4. Data center coolant switch

    SciTech Connect

    Iyengar, Madhusudan K.; Parida, Pritish R.; Schultz, Mark D.

    2015-10-06

    A data center cooling system is operated in a first mode; it has an indoor portion wherein heat is absorbed from components in the data center, and an outdoor heat exchanger portion wherein outside air is used to cool a first heat transfer fluid (e.g., water) present in at least the outdoor heat exchanger portion of the cooling system during the first mode. The first heat transfer fluid is a relatively high performance heat transfer fluid (as compared to the second fluid), and has a first heat transfer fluid freezing point. A determination is made that an appropriate time has been reached to switch from the first mode to a second mode. Based on this determination, the outdoor heat exchanger portion of the data cooling system is switched to a second heat transfer fluid, which is a relatively low performance heat transfer fluid, as compared to the first heat transfer fluid. It has a second heat transfer fluid freezing point lower than the first heat transfer fluid freezing point, and the second heat transfer fluid freezing point is sufficiently low to operate without freezing when the outdoor air temperature drops below a first predetermined relationship with the first heat transfer fluid freezing point.

  5. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    SciTech Connect

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  6. Handling interferences in ⁸⁹Sr and ⁹⁰Sr measurements of reactor coolant water: a method based on strontium separation chemistry.

    PubMed

    Holmgren, S; Tovedal, A; Jonsson, S; Nygren, U; Ramebäck, H

    2014-08-01

    Measurements for determining the activity of (89)Sr and (90)Sr in reactor coolant water are associated with limitations due to interferences from radionuclides with similar chemical properties and β(-)-energies. From a measurement bias point-of-view these interferences would result mainly in an overestimated activity concentration of (90)Sr. In order to address the interference problem, a common and well-known method was used in order to show the need for sufficient decontamination. An improvement was achieved by taking the sample through two initial strontium separations in order to increase the decontamination factor. This method determines the activity concentration of (89)Sr and (90)Sr, via its daughter nuclide (90)Y, by Cherenkov counting. This work is primarily based on theoretical calculations of strontium-, yttrium- and other potential interfering radionuclide ratios after instant fission. The work done to confirm the theoretical calculations were carried through on spiked strontium standard solutions and fresh reactor coolant water. The reactor coolant water was known to contain interferences at a composition resembling that of instant fission. The detection limit for double separation was calculated to 0.17 Bq/kg for (90)Sr and 0.38 Bq/kg for (89)Sr. When using methods that solely rely on strontium resins, this paper shows that the decontamination factor (DF) is significantly higher when using double separation than that of a single separation. The paper also shows that the DF of an initial double strontium separation is as effective when it comes to removing high ratio interferences as separations done with both Sr- and Ln-resin (EiChrom Technologies, Inc., 2003; Tovedal et al., 2009b). However, 'old' samples, where e.g. (140)Ba and (89)Sr has decayed, does not benefit from double separation. Furthermore, samples with low ratios of interfering radionuclides does not benefit from using this method either, seeing as this is a more time consuming method

  7. The effects of CO2, Nd:YAG and Er:YAG lasers with and without surface coolant on tooth root surfaces. An in vitro study.

    PubMed

    Israel, M; Cobb, C M; Rossmann, J A; Spencer, P

    1997-09-01

    The objective of this study was to compare and contrast the morphologic changes in tooth root surfaces treated in vitro by scaling and root planing followed by irradiation with the Er:YAG laser using air/water surface cooling and the CO2 and Nd:YAG lasers, both with and without surface coolant. The experimental unit consisted of 42 freshly extracted teeth which were divided equally and randomly assigned to the following 7 treatment groups: untreated control, S/RP only, CO2 laser with and without air/water surface cooling, Nd:YAG laser with and without/air water surface cooling, and Er:YAG laser with air/water surface coolant. Specimens treated with CO2 laser irradiation were subjected to energy densities ranging from 100 to 400 J/cm2; those treated with the Nd:YAG from 286 to 1857 J/cm2; and the Er:YAG was used within a range of 20 to 120 J/cm2. The degree of morphologic change following CO2 and Nd:YAG irradiation appeared directly related to energy density but unrelated to the use of surface coolant. Laser induced surface changes included cavitation, globules of melted and resolidified mineral, surface crazing, and production of a superficial char layer. In contrast, the Er:YAG laser produced root surface changes that might be expected from acid etching, i.e., removal of the smear layer and exposure of the collagen matrix. In addition, sharply defined microfractures of the mineralized structure were noted and unlike the CO2 and Nd:YAG lasers, there was no evidence of melting or surface char. Given the parameters of this study, it appears that both the CO2 and Nd:YAG lasers alter the root surface in an undesirable manner. The Er:YAG laser, however, when used at low energy densities shows sufficient potential for root surface modification to warrant further investigation. PMID:9378829

  8. Predicting the conditions under which vibroacoustic resonances with external periodic loads occur in the primary coolant circuits of VVER-based NPPs

    NASA Astrophysics Data System (ADS)

    Proskuryakov, K. N.; Fedorov, A. I.; Zaporozhets, M. V.

    2015-08-01

    The accident at the Japanese Fukushima Daiichi nuclear power plant (NPP) caused by an earthquake showed the need of taking further efforts aimed at improving the design and engineering solutions for ensuring seismic resistance of NPPs with due regard to mutual influence of the dynamic processes occurring in the NPP building structures and process systems. Resonance interaction between the vibrations of NPP equipment and coolant pressure pulsations leads to an abnormal growth of dynamic stresses in structural materials, accelerated exhaustion of equipment service life, and increased number of sudden equipment failures. The article presents the results from a combined calculation-theoretical and experimental substantiation of mutual amplification of two kinds of external periodic loads caused by rotation of the reactor coolant pump (RCP) rotor and an earthquake. The data of vibration measurements at an NPP are presented, which confirm the predicted multiple amplification of vibrations in the steam generator and RCP at a certain combination of coolant thermal-hydraulic parameters. It is shown that the vibration frequencies of the main equipment may fall in the frequency band corresponding to the maximal values in the envelope response spectra constructed on the basis of floor accelerograms. The article presents the results from prediction of conditions under which vibroacoustic resonances with external periodic loads take place, which confirm the occurrence of additional earthquake-induced multiple growth of pressure pulsation intensity in the steam generator at the 8.3 Hz frequency and additional multiple growth of vibrations of the RCP and the steam generator cold header at the 16.6 Hz frequency. It is shown that at the elastic wave frequency equal to 8.3 Hz in the coolant, resonance occurs with the frequency of forced vibrations caused by the rotation of the RCP rotor. A conclusion is drawn about the possibility of exceeding the design level of equipment vibrations

  9. Exploratory Investigation of Transpiration Cooling of a 40 deg Double Wedge using Nitrogen and Helium as Coolants at Stagnation Temperatures from 1,295 deg F to 2,910 deg F

    NASA Technical Reports Server (NTRS)

    Rashis, Bernard

    1961-01-01

    An investigation of transpiration cooling has been conducted in the preflight jet of the Langley Pilotless Aircraft Research Station at Wallops Island, Va. The model consisted of a double wedge of 40 deg included angle having a porous stainless-steel specimen inserted flush with the top surface of the wedge. The tests were conducted at a free-stream Mach number of 2.0 for stagnation temperatures ranging from 1,295 F to 2,910 F. Nitrogen and helium were used as coolants and tests were conducted for values ranging from approximately 0.03 to 0.30 percent of the local weight flow rate. The data for both the nitrogen and helium coolants indicated greater cooling effectiveness than that predicted by theory and were in good agreement with the results for an 8 deg cone tested at a stagnation temperature of 600 F. The results indicate that the helium coolant, for the same amount of heat-transfer reduction, requires only about one-fourth to one-fifth the coolant flow weight as the nitrogen coolant.

  10. PACER -- A fast running computer code for the calculation of short-term containment/confinement loads following coolant boundary failure. Volume 2: User information

    SciTech Connect

    Sienicki, J.J.

    1997-06-01

    A fast running and simple computer code has been developed to calculate pressure loadings inside light water reactor containments/confinements under loss-of-coolant accident conditions. PACER was originally developed to calculate containment/confinement pressure and temperature time histories for loss-of-coolant accidents in Soviet-designed VVER reactors and is relevant to the activities of the US International Nuclear Safety Center. The code employs a multicompartment representation of the containment volume and is focused upon application to early time containment phenomena during and immediately following blowdown. PACER has been developed for FORTRAN 77 and earlier versions of FORTRAN. The code has been successfully compiled and executed on SUN SPARC and Hewlett-Packard HP-735 workstations provided that appropriate compiler options are specified. The code incorporates both capabilities built around a hardwired default generic VVER-440 Model V230 design as well as fairly general user-defined input. However, array dimensions are hardwired and must be changed by modifying the source code if the number of compartments/cells differs from the default number of nine. Detailed input instructions are provided as well as a description of outputs. Input files and selected output are presented for two sample problems run on both HP-735 and SUN SPARC workstations.

  11. Regulatory analysis for the resolution of Generic Safety Issue 105: Interfacing system loss-of-coolant accident in light-water reactors

    SciTech Connect

    Not Available

    1993-07-01

    An interfacing systems loss of coolant accident (ISLOCA) involves failure or improper operation of pressure isolation valves (PIVs) that compose the boundary between the reactor coolant system and low-pressure rated systems. Some ISLOCAs can bypass containment and result in direct release of fission products to the environment. A cost/benefit evaluation, using three PWR analyses, calculated the benefit of two potential modifications to the plants. Alternative 1 is improved plant operations to optimize the operator`s performance and reduce human error probabilities. Alternative 2 adds pressure sensing devices, cabling, and instrumentation between two PIVs to provide operators with continuous monitoring of the first PIV. These two alternatives were evaluated for the base case plants (Case 1) and for each plant, assuming the plants had a particular auxiliary building design in which severe flooding would be a problem if an ISLOCA occurred. The auxiliary building design (Case 2) was selected from a survey that revealed a number of designs with features that provided less than optimal resistance to ECCS equipment loss caused by a ISLOCA-induced environment. The results were judged not to provide sufficient basis for generic requirements. It was concluded that the most viable course of action to resolve Generic Issue 105 is licensee participation in individual plant examinations (IPEs).

  12. Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel

    SciTech Connect

    Fauzia, A. F.; Waris, A.; Novitrian

    2010-06-22

    High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950 deg. C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of {sup 235}U enrichment.

  13. Measuring flow and pressure of lithium coolant under developmental testing of a high-temperature cooling system of a space nuclear power plant

    NASA Astrophysics Data System (ADS)

    Sobolev, V. Ya.; Sinyavsky, V. V.

    2014-12-01

    Sub-megawatt space NPP use lithium as a coolant and niobium alloy as a structural material. In order to refine the lithium-niobium technology of the material and design engineering, lithium-niobium loops were worked out in RSC Energia, and they were tested at a working temperature of lithium equal to 1070-1300 K. In order to measure the lithium flow and pressure, special gauges were developed, which made possible the calibration and checkout of the loops without their dismantling. The paper describes the architecture of the electromagnetic flowmeter and the electromagnetic vibrating-wire pressure transducer (gauge) for lithium coolant in the nuclear power plant cooling systems. The operating principles of these meters are presented. Flowmeters have been developed for channel diameters ranging from 10 to 100 mm, which are capable of measuring lithium flows in the range of 0.1 to 30 L/s with the error of 3% for design calibration and 1% for volume graduation. The temperature error of the pressure transducers does not exceed 0.4% per 100 K; the nonlinearity and hysteresis of the calibration curve do not exceed 0.3 and 0.4%, respectively. The transducer applications are illustrated by the examples of results obtained from tests on the NPP module mockup and heat pipes of a radiation cooler.

  14. Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel

    NASA Astrophysics Data System (ADS)

    Fauzia, A. F.; Waris, A.; Novitrian

    2010-06-01

    High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950° C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of 235 U enrichment.

  15. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    NASA Astrophysics Data System (ADS)

    Tyapkov, V. F.

    2014-07-01

    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  16. Probability of pipe fracture in the primary coolant loop of a PWR Plant. Volume 6. Failure mode analysis load combination program. Project I, final report

    SciTech Connect

    Streit, R.D.

    1981-06-01

    Material properties and failure criteria were evaluated to assess the requirements for double-ended guillotine break in the primary coolant loop of the Zion Unit 1 pressurized water reactor. The properties of the 316 stainless steel piping materials were obtained from the literature. Statistical distributions of both the tensile and fracture properties at room and operating temperatures were developed. Yield and ultimate strength tensile properties were combined to estimate the material flow strength. The flow strength and fracture properties were used in the various failure models analyzed. Linear-elastic, elastic-plastic, and fully plastic fracture models were compared, and the governing fracture criterion was determined. For the particular case studied, the fully plastic requirement was found to be the controlling fracture criterion leading to a double-ended guillotine pipe break.

  17. Shuttle active thermal control system development testing. Volume 8: Tube anomaly investigation. [failure of coolant fluid tubes containing Freon 21 refrigerant

    NASA Technical Reports Server (NTRS)

    Reed, M. W.

    1974-01-01

    A modular radiator system projected for use on the space shuttle has been tested. During thermal vacuum testing of the radiator panels leaks developed in the coolant fluid tubes containing the Freon 21 refrigerant. An analysis of thermal vacuum test conditions revealed that the test anomaly consisted of trapped Freon 21 fluid between frozen tube corners colder than the remainder of the panel. The trapped fluid expanded on further warming, causing high pressures to develop in the tubes. The pressure was sufficient to induce tube ruptures. Metallurgical analysis showed that concentric extrusion holes and heat treatment of the aluminum alloy to the -T6 condition would have prevented the ruptures. Thermal analysis indicates that attaching the tube corners to the radiator fin could eliminate the fluid trapping.

  18. Results of Semiscale Mod-2C small-break (5%) loss-of-coolant accident. Experiments S-LH-1 and S-LH-2

    SciTech Connect

    Loomis, G G; Streit, J E

    1985-11-01

    Two experiments simulating small break (5%) loss-of-coolant accidents (5% SBLOCAs) were performed in the Semiscale Mod-2C facility. These experiments were identical except for downcomer-to-upper-head bypass flow (0.9% in Experiment S-LH-1 and 3.0% in Experiment S-LH-2) and were performed at high pressure and temperature (15.6 MPa (2262 psia) system pressure; 37 K (67F) core differential temperature; 595 K(610F) hot leg fluid temperature). From the experimental results, the signature response and transient mass distribution are determined for a 5% SBLOCA. The core thermal-hydraulic response is characterized, including core void distribution maps, and the effect of core bypass flow on transient severity is assessed. Comparisons are made between postexperiment RELAP5 calculations and the experimental results, and the capability of RELAP5 to calculate the phenomena is assessed. 115 figs.

  19. Detailed heat/mass transfer distributions in a rotating two pass coolant channel with engine-near cross section and smooth walls.

    PubMed

    Rathjen, L; Hennecke, D K; Bock, S; Kleinstück, R

    2001-05-01

    This paper shows results obtained by experimental and numerical investigations concerning flow structure and heat/mass transfer in a rotating two-pass coolant channel with engine-near geometry. The smooth two passes are connected by a 180 degrees U-bend in which a 90 degrees turning vane is mounted. The influence of rotation number, Reynolds number and geometry is investigated. The results show a detailed picture of the flow field and distributions of Sherwood number ratios determined experimentally by the use of the naphthalene sublimation technique as well as Nusselt number ratios obtained from the numerical work. Especially the heat/mass transfer distributions in the bend and in the region after the bend show strong gradients, where several separation zones exist and the flow is forced to follow the turbine airfoil shape. Comparisons of numerical and experimental results show only partly good agreement. PMID:11460658

  20. Development and performance of a large-scale, transonic turbine blade cascade facility for aerodynamic studies of merging coolant-mainstream flows

    NASA Astrophysics Data System (ADS)

    Al-Sayeh, Amjad Isaaf

    1998-11-01

    A new, large scale, linear cascade facility of turbine blades has been developed for the experimental exploration of the aerodynamic aspects of film cooling technology. Primary interest is in the mixing of the ejected coolant with the mainstream, at both subsonic and supersonic mainstream Mach numbers at the cascade exit. In order to achieve a spatial resolution adequate for the exploration of details on the scale of the coolant ejection holes, the cascade dimensions were maximized, within the limitations of the air supply system. The cascade contains four blades (three passages) with 14.05 cm axial chord, 17.56 cm span and a design total turning angle of 130.6 degrees. Exit Mach numbers range from 0.6 to 1.5 and Reynolds numbers from 0.5 to 1.5 million. The air supply system capacity allows run times up to five minutes at maximum flow rates. A coolant supply system has been built to deliver mixtures of SFsb6 and air to simulate coolant/mainstream density ratios up to 2. The cascade contains several novel features. A full-perimeter bleed slot upstream of the blades is used to remove the approach boundary layer from all four walls, to improve the degree of two-dimensionality. The exit flow is bounded by two adjustable tailboards that are hinged at the trailing edges and actuated to set the exit flow direction according to the imposed pressure ratio. The boards are perforated and subjected to mass removal near the blades, to minimize the undesirable reflection of shocks and expansion waves. A probe actuator is incorporated that allows continuous positioning of probes in the exhaust stream, in both the streamwise and pitchwise directions. Diagnostic methods include extensive surface pressure taps on the approach and exhaust ducts and on the blade surfaces. The large size permitted as many as 19 taps on the trailing edge itself. Shadowgraph and schlieren are available. A three-prong wake probe has been constructed to simultaneously measure total and static pressures