WorldWideScience

Sample records for coolant circuit components

  1. Reactor pressure vessel and reactor coolant circuit cast duplex stainless steel components contribution of the expertise for life management studies

    International Nuclear Information System (INIS)

    Bezdikian, Georges

    2006-09-01

    The life management of French Nuclear Power Plants is a major stake from an economic and a technical point of view considering the aging management assessment of the key components of the plant. The actual life evaluation is the result of prediction of life assessment from important program of expertise for the 3-loop PWR and 4-loop PWR plants in operation. To optimize the strategic policy in order to achieve the best possible performance and to prepare the technical and economical choice and decision, the paper presents the association of life management strategy and the program of expertise considering: - the identification of degradation for different components and prediction criteria proposed; - the large database from cast reactor coolant and component removed from nuclear power plants and expertise studies to confirm the prediction; - the life evaluation of RPV with radiation surveillance program based on the expertise of irradiation capsules, it is particularly shown how the expertise is in the center of the strategic choice. The French utility has organized the life management of nuclear plant as a function of several programs of expertise of knowledge on the long term experience feedback and the maintenance program for life. This paper shows updated on RPV and reactor coolant equipment activities engaged by utility on: - periodic maintenance and volume of expertise; - Alternative maintenance actions; - Large volume of expertise and how are managed these results to predict the aging management. (author)

  2. Continuous surveillance of reactor coolant circuit integrity

    International Nuclear Information System (INIS)

    1986-01-01

    Continuous surveillance is important to assuring the integrity of a reactor coolant circuit. It can give pre-warning of structural degradation and indicate where off-line inspection should be focussed. These proceedings describe the state of development of several techniques which may be used. These involve measuring structural vibration, core neutron noise, acoustic emission from cracks, coolant leakage, or operating parameters such as coolant temperature and pressure. Twenty three papers have been abstracted and indexed separately for inclusion in the data base

  3. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  4. Monitoring of coolant temperature stratification on piping components in WWER-440 NPPs

    International Nuclear Information System (INIS)

    Hudcovsky, S.; Slanina, M.; Badiar, S.

    2001-01-01

    The presentation deals with the aims of non-standard temperature measurements installed on primary and secondary circuit in WWER-440 NPPs, explains reasons of coolant temperature stratification on the piping components. It describes methods of the measurements on pipings, range of installation of the temperature measurements in EBO and EMO units and illustrates results of measurements of coolant temperature stratification. (Authors)

  5. Feeding and purge systems of coolant primary circuit and coolant secondary circuit control of the I sup(123) target

    International Nuclear Information System (INIS)

    Almeida, G.L. de.

    1986-01-01

    The Radiation Protection Service of IEN (Brazilian-CNEN) detected three faults in sup(123)I target cooling system during operation process for producing sup(123)I: a) non hermetic vessel containing contaminated water from primary coolant circuit; possibility of increasing radioactivity in the vessel due to accumulation of contaminators in cooling water and; situation in region used for personnels to arrange and adjust equipments in nuclear physics area, to carried out maintenance of cyclotron and target coupling in irradiation room. The primary circuit was changed by secondary circuit for target coolant circulating through coil of tank, which receive weater from secondary circuit. This solution solved the three problems simultaneously. (M.C.K.)

  6. Modelling nonstationary thermohydrodynamic processes in heat-exchange circuits with a two-phase coolant

    International Nuclear Information System (INIS)

    Blinkov, V.N.

    1993-01-01

    This paper presents a mathematical model and a open-quotes fastclose quotes computer program for analyzing nonstationary thermohydrodynamic processes in distributed multi-element circuits containing a two-phase coolant. The author's approach is based on representing the distributed multi-element circuits with the two-phase coolant (such as cooling circuits of the reactor of an atomic power station) in the form of equivalent thermohydrodynamic chains composed of idealized elements with the intrinsic properties of the structure elements of real systems. The author has developed the nomenclature of such conceptual elements for objects which can be modelled; the nomenclature encompasses the control volumes (with a single-phase or two-phase coolant or a moving boundary of boiling/condensation) and the branch lines (type of tube and connections in dependence on the inertia of the coolant being taken into account) for a hydrodynamic submodel and the thermal components and lines for a thermal submodel. The mathematical models which have been developed and the program using them are designated for various forms of calculating slow thermohydrodynamic processes in multi-element coolant circuits in reactors and modeling test stands. The program facilitates calculation of the range of stable operation, detailed studies of stationary and nonstationary modes of operation, and forecasts of effective engineering measures to obtain stability with the aid of microcomputers

  7. Inkjet deposited circuit components

    Science.gov (United States)

    Bidoki, S. M.; Nouri, J.; Heidari, A. A.

    2010-05-01

    All-printed electronics as a means of achieving ultra-low-cost electronic circuits has attracted great interest in recent years. Inkjet printing is one of the most promising techniques by which the circuit components can be ultimately drawn (i.e. printed) onto the substrate in one step. Here, the inkjet printing technique was used to chemically deposit silver nanoparticles (10-200 nm) simply by ejection of silver nitrate and reducing solutions onto different substrates such as paper, PET plastic film and textile fabrics. The silver patterns were tested for their functionality to work as circuit components like conductor, resistor, capacitor and inductor. Different levels of conductivity were achieved simply by changing the printing sequence, inks ratio and concentration. The highest level of conductivity achieved by an office thermal inkjet printer (300 dpi) was 5.54 × 105 S m-1 on paper. Inkjet deposited capacitors could exhibit a capacitance of more than 1.5 nF (parallel plate 45 × 45 mm2) and induction coils displayed an inductance of around 400 µH (planar coil 10 cm in diameter). Comparison of electronic performance of inkjet deposited components to the performance of conventionally etched items makes the technique highly promising for fabricating different printed electronic devices.

  8. The electrochemistry of IGSCC mitigation in BWR coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, D.D. [Center for Electrochemical Science and Technology, The Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    A brief review is presented of the electrochemical mitigation of IGSCC in water-cooled reactor heat transport circuit structural materials. Electrochemical control and mitigation is possible, because of the existence of a critical potential for IGSCC and by the feasibility of modifying the environment to displace the corrosion potential (ECP) to a value that is more negative than the critical value. However, even in cases where the ECP cannot be displaced sufficiently in the negative direction to become more negative than the critical potential, considerable advantage is accrued, because of the roughly exponential dependence of crack growth rate on potential. The most important parameters in affecting electrochemical control over the ECP and crack growth rate are the kinetic parameters (exchange current densities and Tafel constants) for the redox reactions involving the principal radiolysis products of water (O{sub 2}, H{sub 2}, H{sub 2}O{sub 2}), external solution composition (concentrations of O{sub 2}, H{sub 2}O{sub 2}, and H{sub 2}), flow velocity, and the conductivity of the bulk environment. The kinetic parameters for the redox reactions essentially determine the charge transfer impedance of the steel surface, which is shown to be one of the key parameters in affecting the magnitude of the coupling current and hence the crack growth rate. The exchange current densities, in particular, are amenable to control by catalysis or inhibition, with the result that surface modification techniques are highly effective in controlling and mitigating IGSCC in reactor coolant circuit materials. (authors)

  9. The corrosion products in the coolant circuits of pressurized water nuclear power plants

    International Nuclear Information System (INIS)

    Darras, R.

    1983-01-01

    The characteristics of the corrosion products formed in the primary and secondary coolant circuits of light-water pressurized reactors are reviewed. The problem induced by the pollution of coolants and metallic surface are examined. Then, the recommendations to follow to minimize the disturbing effects of this pollution by the corrosion products are indicated [fr

  10. Predicted Variations of Water Chemistry in the Primary Coolant Circuit of a Supercritical Water Reactor

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya; Liu, Hong-Ming; Lee, Min

    2012-09-01

    In response to the demand over a higher efficiency for a nuclear power plant, various types of Generation IV nuclear reactors have been proposed. One of the new generation reactors adopts supercritical light water as the reactor coolant. While current in-service light water reactors (LWRs) bear an average thermal efficiency of 33%, the thermal efficiency of a supercritical water reactor (SCWR) could generally reach more than 44%. For LWRs, the coolants are oxidizing due to the presence of hydrogen peroxide and oxygen, and the degradation of structural materials has mainly resulted from stress corrosion cracking. Since oxygen is completely soluble in supercritical water, similar or even worse degradation phenomena are expected to appear in the structural and core components of an SCWR. To ensure proper designs of the structural components and suitable selections of the materials to meet the requirements of operation safety, it would be of great importance for the design engineers of an SCWR to be fully aware of the state of water chemistry in the primary coolant circuit (PCC). Since SCWRs are still in the stage of conceptual design and no practical data are available, a computer model was therefore developed for analyzing water chemistry variation and corrosion behavior of metallic materials in the PCC of a conceptual SCWR. In this study, a U.S. designed SCWR with a rated thermal power of 3575 MW and a coolant flow rate of 1843 kg/s was selected for investigating the variations in redox species concentration in the PCC. Our analyses indicated that the [H 2 ] and [H 2 O 2 ] at the core channel were higher than those at the other regions in the PCC of this SCWR. Due to the self-decomposition of H 2 O 2 , the core channel exhibited a lower [O 2 ] than the upper plenum. Because the middle water rod region was in parallel with the core channel region with relatively high dose rates, the [H 2 ] and [H 2 O 2 ] in this region were higher than those in the other regions

  11. Corrosion products in the coolant circuits of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Darras, R.

    1984-01-01

    The characteristics of corrosion products formed in the primary and secondary circuits of pressurized light water nuclear power plants are first briefly recalled. The problem set by the pollution of coolants and metallic surfaces is then examined. Finally, the measures of precaution to take and the possible solutions to minimize the disturbing effects of this pollution by corrosion products are presented [fr

  12. Secondary coolant circuit for a liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Brachet, A.; Figuet, J.; Guidez, J.; Lions, N.

    1984-01-01

    The invention can be applied for electric power generation with a fast neutron reactor cooled by liquid sodium. Each loop of the main circuit comprises a steam generator, a pump, and at least one heat exchanger disposed in the reactor vessel. A downstream buffer tank is disposed in the tube which is between the steam generator and the pump; the upper buffer tank can be disposed either in to the steam generator or out of this one. The invention allows to suppress the surge tank and the pump can be set in low place without needing a high argon-pressure for the circuit and without putting the storage tank in the secondary loop. It involves a diminution of dimensions and of the installation cost, and an improvement for the safety of heat exchangers in case of water/sodium reaction in the steam generator and, of the reliability for the pump operation [fr

  13. Hard alloys testing-machine for values of PWR primary coolant circuits

    International Nuclear Information System (INIS)

    Campan, J.L.; Sauze, A.

    1980-01-01

    Testing of valve parts or material used in valve fabrication and particularly seizing conditions in friction of plane surfaces coated with hard alloys of the type stellite. The testing equipment called Marguerite is composed of a hot pressurized water loop in conditions similar to PWR primary coolant circuits (320 0 C, 150 bars) and a testing-machine with measuring instruments. Testing conditions and samples are described [fr

  14. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

    2012-09-01

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE - ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

  15. Integrated equipment for increasing and maintaining coolant pressure in primary circuit of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Sykora, D.

    1986-01-01

    An open heat pump circuit is claimed connected to the primary circuit. The pump circuit consists of a steam pressurizer with a built-in steam distributor, a compressor, an expander, a reducing valve, an auxiliary pump, and of water and steam pipes. The operation is described and a block diagram is shown of integrated equipment for increasing and maintaining pressure in the nuclear power plant primary circuit. The appropriate entropy diagram is also shown. The advantage of the open pump circuit consists in reducing the electric power input and electric power consumption for the steam pressurizers, removing entropy loss in heat transfer with high temperature gradient, in the possibility of inserting, between the expander and the auxiliary pump, a primary circuit coolant treatment station, in simplified design and manufacture of the high-pressure steam pressurizer vessel, reducing the weight of the steam pressurizer by changing its shape from cylindrical to spherical, increasing the rate of pressure growth in the primary circuit. (E.S.)

  16. Active components for integrated plasmonic circuits

    DEFF Research Database (Denmark)

    Krasavin, A.V.; Bolger, P.M.; Zayats, A.V.

    2009-01-01

    We present a comprehensive study of highly efficient and compact passive and active components for integrated plasmonic circuit based on dielectric-loaded surface plasmon polariton waveguides.......We present a comprehensive study of highly efficient and compact passive and active components for integrated plasmonic circuit based on dielectric-loaded surface plasmon polariton waveguides....

  17. Materials, coolants, and components for perspective transmutation power units of VVER spent fuel

    International Nuclear Information System (INIS)

    Matal, O.; Hosnedl, P.

    2004-01-01

    The aim of the contribution is provide information on the selection and development of materials and molten salt coolants for high temperature applications for the new nuclear power units and discuss briefly design problems of selected components such as reactor and salt-salt and salt-gas heat exchangers. The following concepts and topics are discussed : Liquid metal reactor concepts (liquid sodium coolant concepts, lead-bismuth coolant concepts, lead coolant concept), CO 2 -cooled reactor concepts, molten-salt fueled and cooled reactor concepts, existing and prospective molten-salt fuels and coolants and structural materials, structural material development for MSR concepts, experimental studies on material corrosion in fluoride molten-salts, and R and D of some components for MSR. (P.A.)

  18. Nuclear power plant with a closed multiple-loop gas coolant circuit. Kernkraftwerk mit geschlossenem, mehrstraengig ausgefuehrtem Gaskuehlkreislauf

    Energy Technology Data Exchange (ETDEWEB)

    Arndt, E.; Brodmann, N.; Roeser, H.

    1972-08-23

    The dimension of the prestressed concrete pressure vessel of the HTR with a closed multiple-loop gas coolant circuit is kept as small as possible by special measures. In each loop several heat exchanger units are supplied by a ring segment-type piping system. This principle is applied to the supply of turbine gas to the recuperators as well as to the supply of the cold, compressed gas to the tube bundles of the recuperators and to the removal of the cold gas from the precoolers. Short stub lines connect the ring segment channels with the corresponding outlets of the gas turbine set. The arrangement of the components may be kept for each increase of power.

  19. Borated coolant decontamination and water-chemical regime of the primary coolant circuit of WWER-1 type reactor

    International Nuclear Information System (INIS)

    Matskevich, G.V.; Klement'eva, E.M.; Plotnikov, I.M.

    1974-01-01

    Treatment of the boron concentrate from the evaporation plant by coagulation and mechanical filtration and H-cation exchange, Bo 3 3- anion exchange makes it possible to return the boron concentrate to the primary loop. The duration of the filter cycle, during the course of which the boron concentrate is very effectively treated with respect to all the 'standard' characteristics, is determined by the start of breakthrough of potassium and ammonium cations from the H-cation exchange filter. After regenerating the cation filter with acid the effectiveness of treatment by the anion exchange filter is restored. The investigation leads to the recommendation of a technology for regenerating boric acid not only from the boron-containing drain waters from the primary loop under conditions of liquid boron control but also, if necessary, from the non-design leakage flows to the special drain system of the station. The optimal water conditions for the primary loop of the WWER-1 reactor when effecting control by a boron-containing coolant are those in which the water is treated with caustic potash and a volatile alkali, ammonia or hydrazine hydrate. When working with a coolant which has a high concentration of boric acid (up to 7-8g/l), corrosion-free conditions for the materials of the high- and low-temperature parts of the primary loop can be achieved only by an additional dose of hydrogen-containing alkali

  20. Circuit bridging of components by smoke

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.J.; Nowlen, S.P.; Anderson, D.J. [Sandia National Labs., Albuquerque, NM (United States)

    1996-10-01

    Smoke can adversely affect digital electronics; in the short term, it can lead to circuit bridging and in the long term to corrosion of metal parts. This report is a summary of the work to date and component-level tests by Sandia National Laboratories for the Nuclear Regulatory Commission to determine the impact of smoke on digital instrumentation and control equipment. The component tests focused on short-term effects such as circuit bridging in typical components and the factors that can influence how much the smoke will affect them. These factors include the component technology and packaging, physical board protection, and environmental conditions such as the amount of smoke, temperature of burn, and humidity level. The likelihood of circuit bridging was tested by measuring leakage currents and converting those currents to resistance in ohms. Hermetically sealed ceramic packages were more resistant to smoke than plastic packages. Coating the boards with an acrylic spray provided some protection against circuit bridging. The smoke generation factors that affect the resistance the most are humidity, fuel level, and burn temperature. The use of CO{sub 2} as a fire suppressant, the presence of galvanic metal, and the presence of PVC did not significantly affect the outcome of these results.

  1. Circuit bridging of components by smoke

    International Nuclear Information System (INIS)

    Tanaka, T.J.; Nowlen, S.P.; Anderson, D.J.

    1996-10-01

    Smoke can adversely affect digital electronics; in the short term, it can lead to circuit bridging and in the long term to corrosion of metal parts. This report is a summary of the work to date and component-level tests by Sandia National Laboratories for the Nuclear Regulatory Commission to determine the impact of smoke on digital instrumentation and control equipment. The component tests focused on short-term effects such as circuit bridging in typical components and the factors that can influence how much the smoke will affect them. These factors include the component technology and packaging, physical board protection, and environmental conditions such as the amount of smoke, temperature of burn, and humidity level. The likelihood of circuit bridging was tested by measuring leakage currents and converting those currents to resistance in ohms. Hermetically sealed ceramic packages were more resistant to smoke than plastic packages. Coating the boards with an acrylic spray provided some protection against circuit bridging. The smoke generation factors that affect the resistance the most are humidity, fuel level, and burn temperature. The use of CO 2 as a fire suppressant, the presence of galvanic metal, and the presence of PVC did not significantly affect the outcome of these results

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Drilled insertions of temperature measurement gauge heads in the main coolant circuits for the installation of instrument pockets

    International Nuclear Information System (INIS)

    Millat, H.; Daemmig, S.

    1990-01-01

    Present day resistance temperature sensors in the primary circuit of the pressurized water reactors units I and II at Beznau nuclear power station are directly inserted in the main coolant circuit. This imposes very high requirements on the pressure and vibration resistance of these measuring elements which manifests itself through relatively high failure rate. New developments in the electronics field today permit sensors to be installed even in the instrument pockets. The application of instrument pockets has had the consequence that the borings in the gauge heads have had to be increased. This modification has required calculated proof as to the strength of the associated pipework to an extent that it is in compliance with the licence. (orig.) [de

  4. Improvements to secondary coolant circuits of a liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Brachet, Alain.

    1981-01-01

    This invention concerns improvements to secondary coolant-systems for sodium cooled nuclear reactors. It further concerns a protective device for a free level mechanical pump which prevents any gas bubbles due to leaks of the working gas of the pump from entering the secondary system of the nuclear reactor [fr

  5. Verification results of methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Saunin, Yuri V.; Dobrotvorski, Alexander N.; Semenikhin, Alexander V.; Korolev, Alexander S. [JSC ' ' Atomtechenergo' ' , Novovoronezh (Russian Federation). Novovoronezh Filial ' ' Novovoronezhatomtechenergo' ' ; Ryasny, Sergei I. [JSC ' ' Atomtechenergo' ' , Moscow (Russian Federation)

    2017-09-15

    The JSC ''Atomtechenergo'' experts have developed a new methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants. The necessity for developing the new methodology was determined by the need to decrease the calculation error of the weighted mean coolant temperature in the hot legs because of the coolant temperature stratification. The methodology development was based on the findings of experimental and calculating research executed by the authors. The methodology verification was fulfilled through comparison of calculation results obtained with and without the methodology use in various operational states and modes of several WWER-1000 power units. The obtained verification results have confirmed that the use of the new methodology provides objective error decrease in determining the weighted mean coolant temperature in the primary circuit hot legs. The decrease value depends on the stratification character which is various for different objects and conditions.

  6. AGING MANAGMENT OF REACTOR COOLANT SYSTEM MECHANICAL COMPONENTS FOR LICENSE RENEWAL

    International Nuclear Information System (INIS)

    SUBUDHI, M.; MORANTE, R.; LEE, A.D.

    2002-01-01

    The reactor coolant system (RCS) mechanical components that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer. steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions. determination of the effects of aging on their intended safety functions. and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. In addition, this paper discusses time-limited aging analyses associated with neutron embrittlement of the reactor vessel beltline region and thermal fatigue

  7. Circuit arrangement of an electronic component for the design of fail-safe protective circuits

    International Nuclear Information System (INIS)

    Centmaier, W.; Bernhard, U.; Friederich, B.; Heisecke, I.

    1974-01-01

    The critical parameters of reactors are controlled by safety circuits. These circuits are controlled designed as logic modules operating by the 'n-out-of-m' selection principle. In most cases, a combination of a '1-out-of-3' circuit with a '2-out-of-3' circuit and separate indication is sufficient for a dynamic fail-safe circuit. The basic logic elements are AND and OR gate circuits, respectively, which are triggered by pulse trains and in which the failure of a pulse train is indicated as an error at the output. The module allows the design of safety circuits offering various degrees of safety. If the indication of an error is made on the modules, faulty components can be exchanged by the maintenance crew right away. (DG) [de

  8. Perceived importance of components of asynchronous music in circuit training

    OpenAIRE

    Crust, Lee

    2008-01-01

    This study examined regular exercisers’ perceptions of specific components of music during circuit training. Twenty-four men (38.8 years, s = 11.8 years) and 31 women (32.4 years, s = 9.6 years) completed two questionnaires immediately after a circuit training class. Participants rated the importance of 13 components of music (rhythm, melody, etc.) in relation to exercise enjoyment, and each completed the Affect Intensity Measure (Larsen, 1984) to measure emotional reactivit...

  9. Microwaves photonic links components and circuits

    CERN Document Server

    Rumelhard, Christian; Billabert, Anne-Laure

    2013-01-01

    This book presents the electrical models for the different elements of a photonic microwave link like lasers, external modulators, optical fibers, photodiodes and phototransistors. The future trends of these components are also introduced: lasers to VCSEL, external modulators to electro-absorption modulators, glass optical fibers to plastic optical fibers, photodiodes to UTC photodiodes or phototransistors. It also describes an original methodology to evaluate the performance of a microwave photonic link, based on the developed elcetrical models, that can be easily incorporated in

  10. Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  11. Generic aging management programs for license renewal of BWR reactor coolant systems components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  12. Requirements on cast steel for the primary coolant circuit of water cooled reactors

    International Nuclear Information System (INIS)

    The most important requirements placed on the structural components of water cooled nuclear reactors include corrosion resistance and mechanical materials properties. Intercrystalline corrosion resistance was tested using the Strauss Test in compliance with the DIN 50914 Standard. Following sensitization between 600 to 700 degC with a dwell time between 15 minutes and 100 hours, a specimen homogeneously annealed with the casting and rapidly water cooled showed no intercrystalline corrosion. Specimens cooled from 1050 degC at a rate of 100 degC per hour showed no unambiguous tendency for intercrystalline corrosion after sensitization; in some cases, however, an initial attack of intercrystalline corrosion was found. It was found that austenitic Cr-Ni cast steel containing 2.5% Mo and about 15% ferrite showed the sensitive intercrystalline corrosion range at higher temperatures and longer dwell times than rolled Cr-Ni steels. In plating the ferritic cast steel with a corrosion resistant plating material, annealing temperature after welding must not exceed 600 to 620 degC otherwise the resistance of the plated layer against intercrystalline corrosion would not be safeguarded, and following annealing for stress removal at a temperature of 600 to 620 degC all requirements must be satisfied by the weld metal and weld transition placed on the initial material. Martensite materials are used for the manufacture of components which are not used under pressure, such as alloys with 13% Cr and 1% to 6% Ni and alloys with 17% Cr and 4% Ni. Carbon content is maintained below 0.10% to guarantee good weldability and the highest corrosion resistance. Cast steels with 13% Cr and 4% Ni after a dwell of 2500 hours in fully desalinated water without oxygen and with 3600 ppm of boron at a test temperature of 95 to 300 degC showed a surface reduction of 0.005 mm annually. In identical conditions except for the water containing oxygen the reduction in surface was 0.05 mm per year. (J.B.)

  13. Modeling of Hydrodynamic Processes at a Large Leak of Water into Sodium in the Fast Reactor Coolant Circuit

    Directory of Open Access Journals (Sweden)

    Sergey Perevoznikov

    2016-10-01

    Full Text Available In this paper, we describe a physicomathematical model of the processes that occur in a sodium circuit with a variable flow cross-section in the case of a water leak into sodium. The application area for this technique includes the possibility of analyzing consequences of this leak as applied to sodium–water steam generators in fast neutron reactors. Hydrodynamic processes that occur in sodium circuits in the event of a water leak are described within the framework of a one-dimensional thermally nonequilibrium three-component gas–liquid flow model (sodium–hydrogen–sodium hydroxide. Consideration is given to the results of a mathematical modeling of experiments involving steam injection into the sodium loop of a circulation test facility. That was done by means of the computer code in which the proposed model had been implemented.

  14. Reverse osmosis and its use at the nuclear power plants. Purification of primary circuit coolant by the means of reverse osmosis

    International Nuclear Information System (INIS)

    Kus, Pavel; Vonkova, Katerina; Kunesova, Katerina; Bartova, Sarka; Skala, Martin; Moucha, Tomáš

    2014-01-01

    This contribution is focused on the use of membrane technologies (e.g. reverse osmosis) for the primary coolant purification at the nuclear power plants. Currently, boric acid present in the primary coolant is preconcentrated at the evaporators, but their operation is very inefficient and expensive. Therefore, reverse osmosis was proposed as one of promising methods possibly replacing evaporators. The aim of the purification process is to achieve boric acid solution of a defined concentration (40 g/l) in the retentate stream in order to recycle it and reuse it in the primary circuit. Additionally, permeate flow should consist solely of pure water. To study the efficiency of several reverse osmosis modulus in the boric acid removal form the water solutions, experimental apparatus was constructed in our laboratory. It consists of the solution reservoir, pump and reverse osmosis modulus. The arrangement of experiments was batch and the retentate flow was refluxed to the feed solution. Several modulus of commercial reverse osmosis membranes were tested. The feed solution contained various concentrations of H 3 BO 3 , KOH, LiOH and NH 3 in order to simulate real primary coolant composition. Based on the experimental results, mathematical model was developed in order to optimize experimental conditions for the best results in primary coolant purification and boric acid preconcentration. (author)

  15. Mathematical Model-Based Temperature Preparation of Liquid-Propellant Components Cooled by Liquid Nitrogen in the Heat Exchanger with a Coolant

    Directory of Open Access Journals (Sweden)

    S. K. Pavlov

    2014-01-01

    Full Text Available Before fuelling the tanks of missiles, boosters, and spacecraft with liquid-propellant components (LPC their temperature preparation is needed. The missile-system ground equipment performs this operation during prelaunch processing of space-purpose missiles (SPM. Usually, the fuel cooling is necessary to increase its density and provide heat compensation during prelaunch operation of SPM. The fuel temperature control systems (FTCS using different principles of operation and types of coolants are applied for fuel cooling.To determine parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is cooled by liquid nitrogen upon contact heat exchange in the coolant reservoir, a mathematical model of this process and a design technique are necessary. Both allow us to determine design parameters of the cooling system and the required liquid nitrogen reserve to cool LPC to the appropriate temperature.The article presents an overview of foreign and domestic publications on cooling processes research and implementation using cryogenic products such as liquid nitrogen. The article draws a conclusion that it is necessary to determine the parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is liquid nitrogen-cooled upon contact heat exchange in the coolant reservoir allowing to define rational propellant cooling conditions to the specified temperature.The mathematical model describes the set task on the assumption that a heat exchange between the LPC and the coolant in the heat exchanger and with the environment through the walls of tanks and pipelines of circulation loops is quasi-stationary.The obtained curves allow us to calculate temperature changes of LPC and coolant, cooling time and liquid nitrogen consumption, depending on the process parameters such as a flow rate of liquid nitrogen, initial coolant temperature, pump characteristics, thermal

  16. Perceived importance of components of asynchronous music during circuit training.

    Science.gov (United States)

    Crust, Lee

    2008-12-01

    This study examined regular exercisers' perceptions of specific components of music during circuit training. Twenty-four men (38.8 years, s = 11.8 years) and 31 women (32.4 years, s = 9.6 years) completed two questionnaires immediately after a circuit training class. Participants rated the importance of 13 components of music (rhythm, melody, etc.) in relation to exercise enjoyment, and each completed the "Affect Intensity Measure" (Larsen, 1984, Dissertation Abstracts International, 5, 2297B. (University microfilms No. 84-22112)) to measure emotional reactivity. Independent t-tests were used to evaluate gender differences in perceptions of musical importance. Pearson correlations were computed to evaluate the relationships between affect intensity, age and importance of musical components. Consistent with previous research and theoretical predictions, rhythm response components (rhythm, tempo, beat) were rated as most important. Women rated the importance of melody significantly higher than did men, whereas men gave more importance to music associated with sport. Affect intensity was found to be positively and significantly related to the perceived importance of melody, lyrical content, musical style, personal associations and emotional content. Results suggest that exercise leaders need to be sensitive to personal factors when choosing music to accompany exercise. Qualitative research that focuses on the personal meaning of music is encouraged.

  17. NGNP Reactor Coolant Chemistry Control Study

    International Nuclear Information System (INIS)

    Castle, Brian

    2010-01-01

    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. The 'NGNP Reactor Coolant Chemistry Control Study' has identified the necessary components and the approach used to control the coolant chemistry in high-temperature gas-cooled reactors (HTGR) for eventual use in the Next Generation Nuclear Plant (NGNP). The focus of this study is to determine what technical developments have taken place since previous helium purification systems (HPSs) were operated in previous HTGRs. Since previous HTGR operational experience, a considerable number of corrosion studies have been performed on various high temperature materials in helium environments with various chemistries at elevated temperatures. This newer work has been reviewed and incorporated into this corrosion study to determine how this new information impacts the design requirements for modern HPSs. Research shows that there are ratios of key impurities that determine whether or not a protective oxide layer is formed or if a deleterious relationship is formed between the material and the coolant. These and other influencing factors are taken into consideration when considering the design requirements for the HPS for the NGNP. In addition to the metallic components in the primary circuit, there are other important material considerations that greatly impact the chemistry requirements for the helium coolant, which are the graphite core support structures. While the metallic components need a small amount of moisture in the coolant to form a protective oxide layer, the graphite needs a very dry environment to prevent degradation. Recommendations and conclusion are made that identify important HPS parameters that should be considered in the design of the HPS for the NGNP. Following the coolant chemistry

  18. Mass analysis of the components separated from printed circuit boards

    Directory of Open Access Journals (Sweden)

    Hana Charvátová

    2010-06-01

    Full Text Available Methods of effective and ecological recycling of printed circuit boards (PCBs are searched all over the world at this time.The material composition and temperature properties of PCB are necessary to be known for an optimal recycling technology. For thispurpose we analyzed weight ratio of the electronic components moulded on the selected kinds of PCBs and next we formulatedmathematic model of temperature field in PCB during a grinding process in that the metal layers are separated from the plasticelements. We present the obtained results in this paper.

  19. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Science.gov (United States)

    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.

  20. Dividing of metal and plastic components of printed circuit boards

    Directory of Open Access Journals (Sweden)

    Krenek Jiri

    2017-01-01

    Full Text Available This paper contents one possibility of PCB separation due to the temperature due to the different thermal expansion of the metal path and the plastic from which the boards are made. Using the knowledge from the literary study and the simulation environments used, we have reached the maximum analysis of the problem. The separation of metal and plastic may occur due to temperature changes if the temperature difference is sufficient. We have carried out a study of printed circuit board production so we can choose the appropriate path of separation. Furthermore, we calculated the shear stress size required to tear the conductive copper paths from the epoxy resin. The temperature field in a two-layer board was modeled in the FEMLAB and Pro/ENGINEER programming environments. From the simulated temperature field simulations, conclusions can be drawn that accurately describe the condition and characteristics of materials subjected to heat shock. They derived the resulting relationship for calculating the resulting shear stress needed to separate the conductive paths and plastic materials of the PCB. In our own experiments, we used several ways to heat PCBs. The temperature is also sufficient for the separation of tin. After using the mechanical separation of the components, they were dropped from the PCB. Mechanical separation was also used when removing conductive paths. This separation is effective, but in the newer types of PCBs, the cyclical effects of thermal shock have to be applied to the separation of copper paths. Laboratory tests have demonstrated the viability of the proposed solution. The proposed method of recycling could lead to industrial use, which requires consistent sorting of waste electrical and electronic equipment.

  1. The influence of EI-21 redox ion-exchange resins on the secondary-coolant circuit water chemistry of vehicular nuclear power installations

    Science.gov (United States)

    Moskvin, L. N.; Rakov, V. T.

    2015-06-01

    The results obtained from testing the secondary-coolant circuit water chemistry of full-scale land-based prototype bench models of vehicular nuclear power installations equipped with water-cooled water-moderated and liquid-metal reactor plants are presented. The influence of copper-containing redox ionexchange resins intended for chemically deoxygenating steam condensate on the working fluid circulation loop's water chemistry is determined. The influence of redox ion-exchange resins on the water chemistry is evaluated by generalizing an array of data obtained in the course of extended monitoring using the methods relating to physicochemical analysis of the quality of condensate-feedwater path media and the methods relating to metallographic analysis of the state of a faulty steam generator's tube system surfaces. The deoxygenating effectiveness of the normal state turbine condensate vacuum deaeration system is experimentally determined. The refusal from applying redox ion-exchange resins in the condensate polishing ion-exchange filters is formulated based on the obtained data on the adverse effect of copper-containing redox ionexchange resins on the condensate-feedwater path water chemistry and based on the data testifying a sufficient effect from using the normal state turbine condensate vacuum deaeration system. Data on long-term operation of the prototype bench model of a vehicular nuclear power installation without subjecting the turbine condensate to chemical deoxygenation are presented.

  2. Thermal response of core and central-cavity components of a high-temperature gas-cooled reactor in the absence of forced convection coolant flow. [NATCON code

    Energy Technology Data Exchange (ETDEWEB)

    Whaley, R.L.; Sanders, J.P.

    1976-09-01

    A means of determining the thermal responses of the core and the components of a high-temperature gas-cooled reactor after loss of forced coolant flow is discussed. A computer program, using a finite-difference technique, is presented together with a solution of the confined natural convection. The results obtained are reasonable and demonstrate that the computer program adequately represents the confined natural convection.

  3. Durability evaluation method for contact component interconnections in printed circuit boards under thermal loads

    Science.gov (United States)

    Azin, Anton; Zhukov, Andrey A.; Ponomarev, Sergey A.; Ponomarev, Sergey V.

    2017-11-01

    In designing sophisticated printed circuit boards, required evaluation of the product life cycle is relevant in terms of presumably applied load conditions during operation. The paper describes the durability evaluation method of printed circuit board components with contact output such as ball grid array (BGA) and pitch grid array (PGA) under thermal loads. Experiment data and numerical simulation results of soldered connections have been obtained. This method is demonstrated by a practical application example-evaluating of printed circuit board durability under cyclic thermal loads. The application of proposed method for printed circuit boards would make it possible to predict the service life of these designed products.

  4. Voltage-Mode All-Pass Filters Including Minimum Component Count Circuits

    Directory of Open Access Journals (Sweden)

    Sudhanshu Maheshwari

    2007-01-01

    Full Text Available This paper presents two new first-order voltage-mode all-pass filters using a single-current differencing buffered amplifier and four passive components. Each circuit is compatible to a current-controlled current differencing buffered amplifier with only two passive elements, thus resulting in two more circuits, which employ a capacitor, a resistor, and an active element, thus using a minimum of active and passive component counts. The proposed circuits possess low output impedance, and hence can be easily cascaded for voltage-mode systems. PSPICE simulation results are given to confirm the theory.

  5. Stochastic Simulation of Integrated Circuits with Nonlinear Black-Box Components via Augmented Deterministic Equivalents

    Directory of Open Access Journals (Sweden)

    MANFREDI, P.

    2014-11-01

    Full Text Available This paper extends recent literature results concerning the statistical simulation of circuits affected by random electrical parameters by means of the polynomial chaos framework. With respect to previous implementations, based on the generation and simulation of augmented and deterministic circuit equivalents, the modeling is extended to generic and ?black-box? multi-terminal nonlinear subcircuits describing complex devices, like those found in integrated circuits. Moreover, based on recently-published works in this field, a more effective approach to generate the deterministic circuit equivalents is implemented, thus yielding more compact and efficient models for nonlinear components. The approach is fully compatible with commercial (e.g., SPICE-type circuit simulators and is thoroughly validated through the statistical analysis of a realistic interconnect structure with a 16-bit memory chip. The accuracy and the comparison against previous approaches are also carefully established.

  6. Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

  7. Neural circuit components of the Drosophila OFF motion vision pathway.

    Science.gov (United States)

    Meier, Matthias; Serbe, Etienne; Maisak, Matthew S; Haag, Jürgen; Dickson, Barry J; Borst, Alexander

    2014-02-17

    Detecting the direction of visual motion is an essential task of the early visual system. The Reichardt detector has been proven to be a faithful description of the underlying computation in insects. A series of recent studies addressed the neural implementation of the Reichardt detector in Drosophila revealing the overall layout in parallel ON and OFF channels, its input neurons from the lamina (L1→ON, and L2→OFF), and the respective output neurons to the lobula plate (ON→T4, and OFF→T5). While anatomical studies showed that T4 cells receive input from L1 via Mi1 and Tm3 cells, the neurons connecting L2 to T5 cells have not been identified so far. It is, however, known that L2 contacts, among others, two neurons, called Tm2 and L4, which show a pronounced directionality in their wiring. We characterized the visual response properties of both Tm2 and L4 neurons via Ca(2+) imaging. We found that Tm2 and L4 cells respond with an increase in activity to moving OFF edges in a direction-unselective manner. To investigate their participation in motion vision, we blocked their output while recording from downstream tangential cells in the lobula plate. Silencing of Tm2 and L4 completely abolishes the response to moving OFF edges. Our results demonstrate that both cell types are essential components of the Drosophila OFF motion vision pathway, prior to the computation of directionality in the dendrites of T5 cells. Copyright © 2014 Elsevier Ltd. All rights reserved.

  8. Structural safety of coolant channel components under excessively high pressure tube diametral expansion rate at garter spring location

    Energy Technology Data Exchange (ETDEWEB)

    Aravind, M. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Sinha, S.K., E-mail: sunilks@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2013-08-15

    Structural safety of coolant channel assembly in the event of high diametral expansion of pressure tube in a 220 MWe pressurised heavy water reactor was investigated using axisymmetric and 3-D finite element models. The axisymmetric analyses were performed and stresses were evaluated for pressure tube, girdle wire and calandria tube at different point of time for diametral expansion rates of 0.2%, 0.25% and 0.3% per year of the pressure tube inside diameter. The results of this study indicated that for the case of 0.3% per year of diametral expansion rate (worst case scenario), occurrence of complete circumferential interference of garter spring with calandria tube at the location of maximum expansion would take place much earlier at around 14 years or 4.2% of the total expansion of pressure tube as opposed to its anticipated design life (30 years). This fact was further corroborated by 3-D finite element analysis performed for the actual assembly configuration under actual loadings. The latter analysis revealed that net section yielding of calandria tube occurs in just 1 year after the occurrence of total circumferential interference between calandria tube and garter spring spacer. It has also been observed that the maximum stress intensity in girdle wire does not increase beyond the ultimate tensile strength even when maximum stress intensity in calandria tube reaches its yield strength. These analyses also revealed that the structural as well as functional integrity of pressure tube and the garter spring is not affected as result of this interference.

  9. Structural safety of coolant channel components under excessively high pressure tube diametral expansion rate at garter spring location

    Science.gov (United States)

    Aravind, M.; Sinha, S. K.

    2013-08-01

    Structural safety of coolant channel assembly in the event of high diametral expansion of pressure tube in a 220 MWe pressurised heavy water reactor was investigated using axisymmetric and 3-D finite element models. The axisymmetric analyses were performed and stresses were evaluated for pressure tube, girdle wire and calandria tube at different point of time for diametral expansion rates of 0.2%, 0.25% and 0.3% per year of the pressure tube inside diameter. The results of this study indicated that for the case of 0.3% per year of diametral expansion rate (worst case scenario), occurrence of complete circumferential interference of garter spring with calandria tube at the location of maximum expansion would take place much earlier at around 14 years or 4.2% of the total expansion of pressure tube as opposed to its anticipated design life (30 years). This fact was further corroborated by 3-D finite element analysis performed for the actual assembly configuration under actual loadings. The latter analysis revealed that net section yielding of calandria tube occurs in just 1 year after the occurrence of total circumferential interference between calandria tube and garter spring spacer. It has also been observed that the maximum stress intensity in girdle wire does not increase beyond the ultimate tensile strength even when maximum stress intensity in calandria tube reaches its yield strength. These analyses also revealed that the structural as well as functional integrity of pressure tube and the garter spring is not affected as result of this interference.

  10. ‘Forcing the issue’ - Aromatic tuning facilitates stimulus-independent modulation of a two-component signaling circuit

    DEFF Research Database (Denmark)

    Nørholm, Morten; von Heijne, Gunnar; Draheim, Roger R.

    2014-01-01

    Two-component signaling circuits allow bacteria to detect and respond to external stimuli. Unfortunately, the input stimulus remains unidentified for the majority of these circuits. Therefore, development of a synthetic method for stimulus-independent modulation of these circuits is highly desira...

  11. Irradiation of electronic components and circuits at the Portuguese Research Reactor: Lessons learned

    Energy Technology Data Exchange (ETDEWEB)

    Marques, J.G.; Ramos, A.R.; Fernandes, A.C.; Santos, J.P. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade de Lisboa, Estrada Nacional 10, 2695-066 Bobadela LRS (Portugal)

    2015-07-01

    The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, since the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor

  12. Predicting the conditions under which vibroacoustic resonances with external periodic loads occur in the primary coolant circuits of VVER-based NPPs

    Science.gov (United States)

    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

  13. Assembling surface mounted components on ink-jet printed double sided paper circuit board

    International Nuclear Information System (INIS)

    Andersson, Henrik A; Manuilskiy, Anatoliy; Haller, Stefan; Sidén, Johan; Nilsson, Hans-Erik; Hummelgård, Magnus; Olin, Håkan; Hummelgård, Christine

    2014-01-01

    Printed electronics is a rapidly developing field where many components can already be manufactured on flexible substrates by printing or by other high speed manufacturing methods. However, the functionality of even the most inexpensive microcontroller or other integrated circuit is, at the present time and for the foreseeable future, out of reach by means of fully printed components. Therefore, it is of interest to investigate hybrid printed electronics, where regular electrical components are mounted on flexible substrates to achieve high functionality at a low cost. Moreover, the use of paper as a substrate for printed electronics is of growing interest because it is an environmentally friendly and renewable material and is, additionally, the main material used for many packages in which electronics functionalities could be integrated. One of the challenges for such hybrid printed electronics is the mounting of the components and the interconnection between layers on flexible substrates with printed conductive tracks that should provide as low a resistance as possible while still being able to be used in a high speed manufacturing process. In this article, several conductive adhesives are evaluated as well as soldering for mounting surface mounted components on a paper circuit board with ink-jet printed tracks and, in addition, a double sided Arduino compatible circuit board is manufactured and programmed. (paper)

  14. Evaluation of blood components exposed to coated arterial filters in extracorporeal circuits.

    Science.gov (United States)

    Hussaini, Bader E; Treanor, Patrick R; Healey, Nancy A; Tilahun, Daniel; Srey, Rithy; Lu, Xiu-Gui; Khuri, Shukri F; Thatte, Hemant S

    2009-09-01

    Biocompatible surfaces play an important role in the inflammatory response during cardiopulmonary bypass (CBP), with the arterial filter contributing a large surface area of the circuit. Different filter-coating materials designed to improve blood-filter biocompatibility are currently used in CPB circuits. This study evaluates eight biocompatible coatings used for arterial filters and their effects on blood components during circulation. Arterial filters were randomly assigned in eight independent heparin-bonded tubing loops and perfused by a single swine (n=8). Arterial blood was routed simultaneously, but separately, into each circuit and circulated for 30 minutes at 37 degrees C. Blood samples were drawn for CBC, ACT, and TAT III measurements at baseline, post-heparinization and post-circulation. At study completion, filters were imaged using multiphoton microscopy. RBC, platelet, and WBC counts, and TAT III complex were all decreased after 30 minutes of circulation; however, WBC count was the only parameter that showed statistically significant differences between the filters. Circulating WBC reduction ranged from 6% (Carmeda and Trillium) to 41% (Terumo-X-coating) with corresponding microscopic confirmation of increased WBC entrapment. All eight filter coatings altered the blood components to varying degrees. Selection of the most effective filter, in conjunction with a heparin-bonded circuit for CPB, may decrease the intraoperative foreign-surface activation of blood cells.

  15. Field analysis and enhancement of multi-pole magnetic components fabricated on printed circuit board

    International Nuclear Information System (INIS)

    Chiu, K.-C.; Chen, C.-S.

    2007-01-01

    A multi-pole magnetic component magnetized with a fine magnetic pole pitch of less than 1 mm is very difficult to achieve by using traditional methods. Moreover, it requires a precise mechanical process and a complicated magnetization system. Different fine magnetic pole pitches of 300, 350 and 400 μm have been accomplished on 9-pole magnetic components through the printed circuit board (PCB) manufacturing technology. Additionally, another fine magnetic pole pitch of 500 μm was also fabricated on a dual-layered (DL) wire circuit structure to investigate the field enhancement. After measurements, a gain factor of 1.37 was obtained in the field strength. The field variations among different magnetic pole pitches were analyzed in this paper

  16. Problems of heavy coolant technology for accelerator-driven systems

    International Nuclear Information System (INIS)

    Gromov, B.F.; Orlov, Yu.I.; Gulevskij, V.A.; Martinov, P.N.; Efimov, E.I.

    1999-01-01

    The title topic is discussed, and the activities of nuclides in the target gas system by 1 MW power after 6 months of work are tabulated. The following facts are stressed: Realization of traditional technology based mainly on cover gas replacement, release and feeding of gas mixtures into the loop is complicated and needs developing and applying devices for purification of the gases removed and radiation shielding of these devices at least. Such developments are especially urgent for ADS projects where coolant circulation is intensified by gas lift. Traditional technology is unable to solve a number of problems associated with the elimination of electropositive impurities (Au, Ag, Hg and others) that make a significant contribution to change compound and radio activity of ADS coolants. The task to remove such impurities was not set for traditional nuclear power plants, so that appropriate developments began only recently. New processes that could replace the developed processes have not been revealed in the course of investigations, however there is an opportunity of using new methods of their realization. For example, the processes of purification and regulation of coolant quality using chemically active gases can be implemented without using external gas systems. The solid hydride, metallooxide and other chemical compounds located in the gas volume or in the coolant loops of ADS can be the sources and absorbents of chemically active gas components. Applying to ADS conditions the emphasis of some processes goals can be slightly changed. For example: regulation of dissolved oxygen activity can be performed as well (probably, mainly) for oxygen refining of coolant by means of electronegative impurities transition into solid-phase oxide compounds; gas and coolant filtering can also be performed to localize radioactive impurities in the ADS circuit. Probably, the coolant technology can be employed to solve a number of tasks: to localize radioactive impurities in the

  17. Increasing of prediction reliability of calcium carbonate scale formation in heat exchanger of secondary coolant circuits of thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Tret'yakov, O.V.; Kritskij, V.G.; Styazhkin, P.S.

    1991-01-01

    Calcium carbonate scale formation in the secondary circuit heat exchanger of thermal and nuclear power plants is investigated. A model of calcium-carbonate scale formation providing quite reliable prediction of process running and the possibility of its control affecting the parameters of hydrochemical regime (HCR) is developed. The results can be used when designing the automatic-control system of HCR

  18. Influence of hydrazine primary water chemistry on corrosion of fuel cladding and primary circuit components

    International Nuclear Information System (INIS)

    Iourmanov, V.; Pashevich, V.; Bogancs, J.; Tilky, P.; Schunk, J.; Pinter, T.

    1999-01-01

    Earlier at Paks 1-4 NPP standard ammonia chemistry was in use. The following station performance indicators were improved when hydrazine primary water chemistry was introduced: occupational radiation exposures of personnel; gamma-radiation dose rates near primary system components during refuelling and maintenance outages. The reduction of radiation exposures and radiation fields were achieved without significant expenses. Recent results of experimental studies allowed to explain the mechanism of hydrazine dosing influence on: corrosion rate of structure materials in primary coolant; behaviour of soluble and insoluble corrosion products including long-life corrosion-induced radionuclides in primary system during steady-state and transient operation modes; radiolytic generation of oxidising radiolytic products in core and its corrosion activity in primary system; radiation situation during refuelling and maintenance outages; foreign material degradation and removal (including corrosion active oxidant species) from primary system during abnormal events. Operational experience and experimental data have shown that hydrazine primary water chemistry allows to reduce corrosion wear and thereby makes it possible to extend the life-time of plant components in primary system. (author)

  19. Process technology and effects of spallation products: Circuit components, maintenance, and handling

    International Nuclear Information System (INIS)

    Sigg, B.; Haines, S.J.; Dressler, R.; McManamy, T.

    1996-01-01

    Working Session D included an assessment of the status of the technology and components required to: (1) remove impurities from the liquid metal (mercury or Pb-Bi) target flow loop including the effects of spallation products, (2) provide the flow parameters necessary for target operations, and (3) maintain the target system. A series of brief presentations were made to focus the discussion on these issues. The subjects of these presentations, and presenters were: (1) Spallation products and solubilities - R. Dressler; (2) Spallation products for Pb-Bi - Y. Orlov; (3) Clean/up/impurity removal components - B. Sigg; (4) open-quotes Road-Mapclose quotes and remote handling needs - T. McManamy; (5) Remote handling issues and development - M. Holding. The overall conclusion of this session was that, with the exception of (i) spallation product related processing issues, (ii) helium injection and clean-up, and (iii) specialized remote handling equipment, the technology for all other circuit components (excluding the target itself) exists. Operating systems at the Institute of Physics in Riga, Latvia (O. Lielausis) and at Ben-Gurion University in Beer Shiva, Israel (S. Lesin) have demonstrated that other liquid metal circuit components including pumps, heat exchangers, valves, seals, and piping are readily available and have been reliably used for many years. In the three areas listed above, the designs and analysis are not judged to be mature enough to determine whether and what types of technology development are required. Further design and analysis of the liquid metal target system is therefore needed to define flow circuit processing and remote handling equipment requirements and thereby identify any development needs

  20. The operating reliability of the reactor coolant pump

    International Nuclear Information System (INIS)

    Grancy, W.

    1996-01-01

    There is a strong tendency among operating companies and manufacturers of nuclear power stations to further increase safety and operating availability of the plant and of its components. This applies also and particularly to reactor coolant pumps for the primary circuit of nuclear power stations of the type PWR. For 3 decades, ANDRITZ has developed and built such pumps and has attached great importance to the design of the complete pump rotor and of its essential surrounding elements, such as bearing and shaft seal. Apart from questions connected with design functioning of the pump there is one question of top priority: the operating reliability of the reactor coolant pump. The pump rotor (together with the rotor of the drive motor) is the only component within the primary system that permanently rotates at high speed during operation of the reactor plant. Many questions concerning design and configuration of such components cannot be answered purely theoretically, or they can only be answered partly. Therefore comprehensive development work and testing was necessary to increase the operating reliability of the pump rotor itself and of its surrounding elements. This contribution describes the current status of development and, as a focal point, discusses shaft sealing solutions elaborated so far. In this connection also a sealing system will be presented which aims for the first time at using a two-stage mechanical seal in reactor coolant pumps

  1. Microwave Photonic Architecture for Direction Finding of LPI Emitters: Front End Analog Circuit Design and Component Characterization

    Science.gov (United States)

    2016-09-01

    PHOTONIC ARCHITECTURE FOR DIRECTION FINDING OF LPI EMITTERS: FRONT-END ANALOG CIRCUIT DESIGN AND COMPONENT CHARACTERIZATION by Chew K. Tan... PHOTONIC ARCHITECTURE FOR DIRECTION FINDING OF LPI EMITTERS: FRONT-END ANALOG CIRCUIT DESIGN AND COMPONENT CHARACTERIZATION 5. FUNDING NUMBERS 6. AUTHOR...miniature microwave- photonic phase-sampling DF technique is investigated in this thesis. This front-end design uses a combination of integrated optical

  2. Extended Life Coolant Testing

    Science.gov (United States)

    2016-06-06

    military vehicles. Newer vehicles come factory- filled with ELC, while the Army continues to use traditional supplemental coolant additives (SCA)-based...UNCLASSIFIED TABLE OF CONTENTS EXTENDED LIFE COOLANT TESTING INTERIM REPORT TFLRF No. 478 by Gregory A. T. Hansen Edwin A...longer needed. Do not return it to the originator. UNCLASSIFIED UNCLASSIFIED EXTENDED LIFE COOLANT TESTING INTERIM REPORT TFLRF No

  3. Estimative of core damage frequency in IPEN's IEA-R1 research reactor (PSA level 1) due to the initiating event of loss of coolant caused by large rupture in the pipe of the primary circuit

    International Nuclear Information System (INIS)

    Hirata, Daniel Massami

    2009-01-01

    This work applies the methodology of probabilistic safety assessment level 1 to the research reactor IEA-R1 IPEN-CNEN/SP. Two categories of identified initiating events of accidents in the reactor are studied: loss of flow and loss of primary coolant. Among the initiating events, blockage of flow channel and loss of cooling fluid by major pipe rupture in the primary circuit are chosen for a detailed analysis. The event tree technique is used to analyze the evolution of the accident, including the actuation or the fail of actuation of the safety systems and the reactor damages. Using the fault tree the reliability of the following reactor safety systems is evaluated: reactor shutdown system, isolation of the reactor pool, emergency core cooling system (ECCS) and the electric system. Estimative for the frequency of damage to the reactor core and the probability of failure of the analyzed systems are calculated. The estimated values for the frequencies of core damage are within the expected margins and are of the same order of magnitude as those found for similar reactors. The reliability of the reactor shutdown system, isolation of the reactor pool and ECCS are satisfactory for the conditions these systems are required. However, for the electric system it is suggested an upgrade to increase its reliability. (author)

  4. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A. [St. Petersburg State Technical Univ. (Russian Federation)

    1997-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  5. Musical molecules: the molecular junction as an active component in audio distortion circuits

    Science.gov (United States)

    Bergren, Adam Johan; Zeer-Wanklyn, Lucas; Semple, Mitchell; Pekas, Nikola; Szeto, Bryan; McCreery, Richard L.

    2016-03-01

    Molecular junctions that have a non-linear current-voltage characteristic consistent with quantum mechanical tunneling are demonstrated as analog audio clipping elements in overdrive circuits widely used in electronic music, particularly with electric guitars. The performance of large-area molecular junctions fabricated at the wafer level is compared to currently standard semiconductor diode clippers, showing a difference in the sound character. The harmonic distributions resulting from the use of traditional and molecular clipping elements are reported and discussed, and differences in performance are noted that result from the underlying physics that controls the electronic properties of each clipping component. In addition, the ability to tune the sound using the molecular junction is demonstrated. Finally, the hybrid circuit is compared to an overdriven tube amplifier, which has been the standard reference electric guitar clipped tone for over 60 years. In order to investigate the feasibility of manufacturing molecular junctions for use in commercial applications, devices are fabricated using a low-density format at the wafer level, where 38 dies per wafer, each containing two molecular junctions, are made with exceptional non-shorted yield (99.4%, representing 718 out of 722 tested devices) without requiring clean room facilities.

  6. Musical molecules: the molecular junction as an active component in audio distortion circuits

    International Nuclear Information System (INIS)

    Bergren, Adam Johan; Zeer-Wanklyn, Lucas; Pekas, Nikola; Szeto, Bryan; McCreery, Richard L; Semple, Mitchell

    2016-01-01

    Molecular junctions that have a non-linear current–voltage characteristic consistent with quantum mechanical tunneling are demonstrated as analog audio clipping elements in overdrive circuits widely used in electronic music, particularly with electric guitars. The performance of large-area molecular junctions fabricated at the wafer level is compared to currently standard semiconductor diode clippers, showing a difference in the sound character. The harmonic distributions resulting from the use of traditional and molecular clipping elements are reported and discussed, and differences in performance are noted that result from the underlying physics that controls the electronic properties of each clipping component. In addition, the ability to tune the sound using the molecular junction is demonstrated. Finally, the hybrid circuit is compared to an overdriven tube amplifier, which has been the standard reference electric guitar clipped tone for over 60 years. In order to investigate the feasibility of manufacturing molecular junctions for use in commercial applications, devices are fabricated using a low-density format at the wafer level, where 38 dies per wafer, each containing two molecular junctions, are made with exceptional non-shorted yield (99.4%, representing 718 out of 722 tested devices) without requiring clean room facilities. (paper)

  7. Enrichment of the metallic components from waste printed circuit boards by a mechanical separation process using a stamp mill

    International Nuclear Information System (INIS)

    Yoo, Jae-Min; Jeong, Jinki; Yoo, Kyoungkeun; Lee, Jae-chun; Kim, Wonbaek

    2009-01-01

    Printed circuit boards incorporated in most electrical and electronic equipment contain valuable metals such as Cu, Ni, Au, Ag, Pd, Fe, Sn, and Pb. In order to employ a hydrometallurgical route for the recycling of valuable metals from printed circuit boards, a mechanical pre-treatment step is needed. In this study, the metallic components from waste printed circuit boards have been enriched using a mechanical separation process. Waste printed circuit boards shredded to 5.0 mm. The fractions of milled printed circuit boards of size 5.0 mm fraction and the heavy fraction were subjected to two-step magnetic separation. Through the first magnetic separation at 700 Gauss, 83% of the nickel and iron, based on the whole printed circuit boards, was recovered in the magnetic fraction, and 92% of the copper was recovered in the non-magnetic fraction. The cumulative recovery of nickel-iron concentrate was increased by a second magnetic separation at 3000 Gauss, but the grade of the concentrate decreased remarkably from 76% to 56%. The cumulative recovery of copper concentrate decreased, but the grade increased slightly from 71.6% to 75.4%. This study has demonstrated the feasibility of the mechanical separation process consisting of milling/size classification/gravity separation/two-step magnetic separation for enriching metallic components such as Cu, Ni, Al, and Fe from waste printed circuit boards

  8. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    Tingey, G.L.; Morgan, W.C.

    1975-01-01

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

  9. Study of heat and mass transfer in a steam generator with chemically reacting coolant

    International Nuclear Information System (INIS)

    Lemeshev, V.U.; Mikhalevich, A.A.; Nemtsev, V.A.; Nesterenko, V.B.

    1983-01-01

    A one-dimensional mathematical model is represented once-through type and heat and mass transfer steam generator with turbulent flow of chemically reacting N 2 O 4 -NO coolant is investigated. During development of the mathematical model it has been assumed that the process of heating and boiling of liquid N 2 O 4 -NO coolant as well as superheating of produced vapour at subcritical parameters or heating of pseudo-liquid and superheating of produced pseudovapour at supercritical parameters (the heated side) is carried out at the expense of gaseous N 2 O 4 -NO coolant cooling (the heating side). The process of heating and cooling of the N 2 O 4 -NO system is followed by N 2 O 4 reversible 2NO 2 (1); 2NO 2 reversible 2NO+O 2 (2); N 2 O 3 reVersible NO 2 +NO (3) reactions, whereas the reactions (1) and (3) are practically equilibrium and the reaction (2) proceeds for the time comparable with the coolant residence time in the reactor circuit and the reaction rate is to be taken into account at mathematical modelling of the heat and mass transfer processes in the equipment. The modelling of thermal and hydrodynamic processes in the elements of a powergenerating components is needed for developing power plants with a dissociating coolant

  10. Two Independent Mushroom Body Output Circuits Retrieve the Six Discrete Components of Drosophila Aversive Memory

    Directory of Open Access Journals (Sweden)

    Emna Bouzaiane

    2015-05-01

    Full Text Available Understanding how the various memory components are encoded and how they interact to guide behavior requires knowledge of the underlying neural circuits. Currently, aversive olfactory memory in Drosophila is behaviorally subdivided into four discrete phases. Among these, short- and long-term memories rely, respectively, on the γ and α/β Kenyon cells (KCs, two distinct subsets of the ∼2,000 neurons in the mushroom body (MB. Whereas V2 efferent neurons retrieve memory from α/β KCs, the neurons that retrieve short-term memory are unknown. We identified a specific pair of MB efferent neurons, named M6, that retrieve memory from γ KCs. Moreover, our network analysis revealed that six discrete memory phases actually exist, three of which have been conflated in the past. At each time point, two distinct memory components separately recruit either V2 or M6 output pathways. Memory retrieval thus features a dramatic convergence from KCs to MB efferent neurons.

  11. Utilisation of symmetrical components in a communication-based protection for loop MV feeders with variable short-circuit power

    DEFF Research Database (Denmark)

    Ciontea, Catalin-Iosif; Bak, Claus Leth; Blaabjerg, Frede

    2018-01-01

    Variability of the available short-circuit power also implies variation of the fault level, which can potentially cause several protection problems in the electric networks. In this paper, a novel protection method that is insensitive to the fault level changes caused by variable short......-circuit power is presented. It relies on utilisation of symmetrical components of the short-circuit currents and on communication between the protection relays. The proposed method addresses the Single Phase to Ground (SPG) faults occurring in directly grounded distribution networks, with focus on closed......-loop Medium Voltage (MV) feeders. Case studies are presented, which demonstrate that the proposed protection scheme is capable of effectively detecting the SPG faults in closed-loop feeders with variable short-circuit power....

  12. Leak detection system for RBMK coolant circuit

    International Nuclear Information System (INIS)

    Cherkashov, Ju.M.; Strelkov, B.P.; Korolev, Yu.V.; Eperin, A.P.; Kozlov, E.P.; Belyanin, L.A.; Vanukov, V.N.

    1996-01-01

    In report the description of an object of the control is submitted, requests to control of leak-tightness and functioning of system are formulated, analysis of a current status on NPP with RBMK is submitted, review of methods of the leak-tightness monitoring, their advantage and defects with reference to conditions and features of a design RBMK is indicated, some results of tests and operation of various monitoring methods are submitted, requests on interaction of operative staff, leak-tightness monitoring system and protection system of reactor are submitted. (author). 11 figs, 1 tab

  13. Active component modeling for analog integrated circuit design. Model parametrization and implementation in the SPICE-PAC circuit simulator

    International Nuclear Information System (INIS)

    Marchal, Xavier

    1992-01-01

    In order to use CAD efficiently in the analysis and design of electronic Integrated circuits, adequate modeling of active non-linear devices such as MOSFET transistors must be available to the designer. Many mathematical forms can be given to those models, such as explicit relations, or implicit equations to be solved. A major requirement in developing MOS transistor models for IC simulation is the availability of electrical characteristic curves over a wide range of channel width and length, including the sub-micrometer range. To account in a convenient way for bulk charge influence on I DS = f(V DS , V GS , v BS ) device characteristics, all 3 standard SPICE MOS models use an empirical fitting parameter called the 'charge sharing factor'. Unfortunately, this formulation produces models which only describe correctly either some of the short channel phenomena, or some particular operating conditions (low injection, avalanche effect, etc.). We present here a cellular model (CDM = Charge Distributed Model) implemented in the open modular SPICE-PAC Simulator; this model is derived from the 4-terminal WANG charge controlled MOSFET model, using the charge sheet approximation. The CDM model describes device characteristics in ail operating regions without introducing drain current discontinuities and without requiring a 'charge sharing factor'. A usual problem to be faced by designers when they simulate MOS ICs is to find a reliable source of model parameters. Though most models have a physical basis, some of their parameters cannot be easily estimated from physical considerations. It can also happen that physically determined parameters values do not produce a good fit to measured device characteristics. Thus it is generally necessary to extract model parameters from measured transistor data, to ensure that model equations approximate measured curves accurately enough. Model parameters extraction can be done in 2 different ways, exposed in this

  14. Organic coolant summary report

    International Nuclear Information System (INIS)

    Smee, J.L.; Puttagunta, V.R.; Robertson, R.F.S.; Hatcher, S.R.

    1975-08-01

    This report summarizes the development of the use of organic liquids, specifically the terphenyls, as heat transfer mediums in nuclear reactors. All phases of the development program are covered, including the choice of coolant, decomposition, coolant reprocessing, physical properties, fouling, heat transfer, and corrosion. Few experimental results are given, but rather the emphasis is on summarizing the state of the art as it exists within AECL at the end of 1973. (Author)

  15. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  16. Design and test of component circuits of an integrated quantum voltage noise source for Johnson noise thermometry

    International Nuclear Information System (INIS)

    Yamada, Takahiro; Maezawa, Masaaki; Urano, Chiharu

    2015-01-01

    Highlights: • We demonstrated RSFQ digital components of a new quantum voltage noise source. • A pseudo-random number generator and variable pulse number multiplier are designed. • Fabrication process is based on four Nb wiring layers and Nb/AlOx/Nb junctions. • The circuits successfully operated with wide dc bias current margins, 80–120%. - Abstract: We present design and testing of a pseudo-random number generator (PRNG) and a variable pulse number multiplier (VPNM) which are digital circuit subsystems in an integrated quantum voltage noise source for Jonson noise thermometry. Well-defined, calculable pseudo-random patterns of single flux quantum pulses are synthesized with the PRNG and multiplied digitally with the VPNM. The circuit implementation on rapid single flux quantum technology required practical circuit scales and bias currents, 279 junctions and 33 mA for the PRNG, and 1677 junctions and 218 mA for the VPNM. We confirmed the circuit operation with sufficiently wide margins, 80–120%, with respect to the designed bias currents.

  17. Design and test of component circuits of an integrated quantum voltage noise source for Johnson noise thermometry

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, Takahiro, E-mail: yamada-takahiro@aist.go.jp [Nanoelectronics Research Institute, National Institute of Advanced Industrial Science and Technology, Central 2, Umezono 1-1-1, Tsukuba, Ibaraki 305-8568 (Japan); Maezawa, Masaaki [Nanoelectronics Research Institute, National Institute of Advanced Industrial Science and Technology, Central 2, Umezono 1-1-1, Tsukuba, Ibaraki 305-8568 (Japan); Urano, Chiharu [National Metrology Institute of Japan, National Institute of Advanced Industrial Science and Technology, Central 3, Umezono 1-1-1, Tsukuba, Ibaraki 305-8563 (Japan)

    2015-11-15

    Highlights: • We demonstrated RSFQ digital components of a new quantum voltage noise source. • A pseudo-random number generator and variable pulse number multiplier are designed. • Fabrication process is based on four Nb wiring layers and Nb/AlOx/Nb junctions. • The circuits successfully operated with wide dc bias current margins, 80–120%. - Abstract: We present design and testing of a pseudo-random number generator (PRNG) and a variable pulse number multiplier (VPNM) which are digital circuit subsystems in an integrated quantum voltage noise source for Jonson noise thermometry. Well-defined, calculable pseudo-random patterns of single flux quantum pulses are synthesized with the PRNG and multiplied digitally with the VPNM. The circuit implementation on rapid single flux quantum technology required practical circuit scales and bias currents, 279 junctions and 33 mA for the PRNG, and 1677 junctions and 218 mA for the VPNM. We confirmed the circuit operation with sufficiently wide margins, 80–120%, with respect to the designed bias currents.

  18. Method of cleaning up coolants

    International Nuclear Information System (INIS)

    Asakura, Yamato; Kikuchi, Makoto; Uchida, Shunsuke.

    1979-01-01

    Purpose: To decrease the dose rates of the pipeways outside of reactors by charging adsorbents in a primary coolant circuits and recovering adsorbed solid impurities and ionic impurities in a reactor clean up system. Method: An addition device incorporating cleaning agents (adsorbents: metal oxide such as TiO 2 ) is connected to a feedwater pipe and a mixer for reacting the cleaning agent and solid impurities in feedwater is also provided. The cleaning agents added from the adding device into the feedwater are agitated in the mixer, adsorb the solid impurities in the feedwater and enter the reactor as reactor water while suspended in the feedwater. The reactor water is passed through reactor water clean up pipeway by a pump and removed with the cleaning agents and the solid impurities adsorbed thereon through the filter. (Seki, T.)

  19. Direct intertectal inputs are an integral component of the bilateral sensorimotor circuit for behavior in Xenopus tadpoles.

    Science.gov (United States)

    Gambrill, Abigail C; Faulkner, Regina L; Cline, Hollis T

    2018-02-14

    The circuit controlling visually-guided behavior in non-mammalian vertebrates, like Xenopus tadpoles, includes retinal projections to the contralateral optic tectum, where visual information is processed, and tectal motor outputs projecting ipsilaterally to hindbrain and spinal cord. Tadpoles have an intertectal commissure whose function is unknown, but it might transfer information between the tectal lobes. Differences in visual experience between the two eyes have profound effects on the development and function of visual circuits in animals with binocular vision, but the effects on animals with fully-crossed retinal projections are not clear. We tested the effect of monocular visual experience on the visuomotor circuit in Xenopus tadpoles. We show that cutting the intertectal commissure or providing visual experience to one eye (monocular visual experience) are both sufficient to disrupt tectally-mediated visual avoidance behavior. Monocular visual experience induces asymmetry in tectal circuit activity across the midline. Repeated exposure to monocular visual experience drives maturation of the stimulated retinotectal synapses, seen as increased AMPA/NMDA ratios, induces synaptic plasticity in intertectal synaptic connections and induces bilaterally asymmetric changes in the tectal excitation/inhibition ratio (E/I). We show that unilateral expression of peptides that interfere with AMPA or GABAA receptor trafficking alters E/I in the transfected tectum and is sufficient to degrade visuomotor behavior. Our study demonstrates that monocular visual experience in animals with fully-crossed visual systems produces asymmetric circuit function across the midline and degrades visuomotor behavior. The data further suggest that intertectal inputs are an integral component of a bilateral visuomotor circuit critical for behavior.

  20. The Plateau de Bure ASTEP Platform Test in natural radiation environment of electronic components and circuits

    International Nuclear Information System (INIS)

    Autran, J.L.; Munteanu, D.; Sauze, S.; Roche, Ph.; Gasiot, G.; Borel, J.

    2010-01-01

    Reducing the size of microelectronic devices and increasing the integration density of circuits lead (following the famous Moore's law) to an increased sensitivity of circuits to natural terrestrial radiation environment. - Such sensitivity to atmospheric particles (mainly neutrons) can cause non-destructive (soft-errors) or destructive (latch-up) failures in most electronic circuits, including volatile static memories (SRAM), object of the research work carried out since 2004 on the European Test Platform ASTER. - This paper presents in details the ASTEP platform, its location, the instruments (neutron monitor of the Plateau de Bure) and the experiences (memory tester) currently installed on the Plateau de Bure. In a second part, we also report a synthesis of the key results concerning the natural radiation sensitivity of SRAM fabricated in 130 nm and 65 nm bulk silicon technologies. (authors)

  1. Problems of hydrogen - water vapor - inert gas mixture use in heavy liquid metal coolant technology

    International Nuclear Information System (INIS)

    Ul'yanov, V.V.; Martynov, P.N.; Gulevskij, V.A.; Teplyakov, Yu.A.; Fomin, A.S.

    2014-01-01

    The reasons of slag deposit formation in circulation circuits with heavy liquid metal coolants, which can cause reactor core blockage, are considered. To prevent formation of deposits hydrogen purification of coolant and surfaces of circulation circuit is used. It consists in introduction of gaseous mixtures hydrogen - water vapor - rare gas (argon or helium) directly into coolant flow. The principle scheme of hydrogen purification and the processes occurring during it are under consideration. Measures which make it completely impossible to overlap of the flow cross section of reactor core, steam generators, pumps and other equipment by lead oxides in reactor facilities with heavy liquid metal coolants are listed [ru

  2. Reliability improvement of electronic circuits based on physical failure mechanisms in components

    NARCIS (Netherlands)

    Brombacher, A.C.; de Boer, H.A.; Bennion, M.; Fennema, P.H.; Hermann, O.E.

    1991-01-01

    Traditionally the position of reliability analysis in the design and production process of electronic circuits is a position of reliability verification. A completed design is checked on reliability aspects and either rejected or accepted for production. This paper describes a method to model

  3. Qualitative infrared spectral analysis of products adsorbed by silica gel from ditolylmethane coolant and their adsorption isotherm

    International Nuclear Information System (INIS)

    Ermakov, V.A.; Benderskaya, O.S.

    1987-01-01

    The IR-spectral analysis has been applied to study the products adsorbed from ditolylmethane first-circuit coolant, as well as from still bottoms after coolant distillation on silicagel of various makes. The qualitative study of desorbate IR-spectra has shown that they refer to the classes of arylaldehydes, diarylketones and carbonic acids. Under actual conditions first-circuit reactor coolant also has a wide set of products of its radiolysis, therefore the spectrum of coolant oxidaton products must be wider. It is noted that adsorption on silica gel, ASK of oxygen-bearing compounds which are present in ditolyl methane coolant has 2 stages

  4. Environmentally Friendly Coolant System

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Integrated circuits and molecular components for stress and feeding: implications for eating disorders.

    Science.gov (United States)

    Hardaway, J A; Crowley, N A; Bulik, C M; Kash, T L

    2015-01-01

    Eating disorders are complex brain disorders that afflict millions of individuals worldwide. The etiology of these diseases is not fully understood, but a growing body of literature suggests that stress and anxiety may play a critical role in their development. As our understanding of the genetic and environmental factors that contribute to disease in clinical populations like anorexia nervosa, bulimia nervosa and binge eating disorder continue to grow, neuroscientists are using animal models to understand the neurobiology of stress and feeding. We hypothesize that eating disorder clinical phenotypes may result from stress-induced maladaptive alterations in neural circuits that regulate feeding, and that these circuits can be neurochemically isolated using animal model of eating disorders. © 2014 John Wiley & Sons Ltd and International Behavioural and Neural Genetics Society.

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

    Science.gov (United States)

    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.

  7. Digital logic circuit based on two component molecular systems of BSA and salen

    Science.gov (United States)

    Hai-Bin, Lin; Feng, Chen; Hong-Xu, Guo

    2018-02-01

    A new fluorescent molecular probe 1 was designed and constructed by combining bovine serum albumin (BSA) and N,N‧-bis(salicylidene)ethylenediamine (salen). Stimulated by Zn2 +, tris, or EDTAH2Na2, the distance between BSA and salen was regulated, which was accompanied by an obvious change in the fluorescence intensity at 350 or 445 nm based on Förster resonance energy transfer. Moreover, based on the encoding binary digits in these inputs and outputs applying positive logic conventions, a monomolecular circuit integrating one OR, three NOT, and three YES gates, was successfully achieved.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

    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

  9. SUBSTATIONS OF DISTRICT HEATING SYSTEMS WITH PULSE COOLANT CIRCULATION

    Directory of Open Access Journals (Sweden)

    Andrey N. Makeev

    2017-01-01

    Full Text Available Abstract. Objectives The aim of the study is to generalise the results of the application of technologies and means for organising pulse coolant flow within a district heating system in order to increase its energy efficiency based on the organisation of local hydraulic shocks and the subsequent use of their energy to ensure the purification of heat energy equipment, intensify the heat transfer process and realise the possibility of transforming the available head from one hydraulic circuit to another. Methods Substations connecting the thermal power installations of consumers with heat networks via dependent and independent schemes are analytically generalised. The use of pulse coolant circulation is proposed as a means of overcoming identified shortcomings. Results Principal schemes of substations with pulse coolant circulation for dependent and independent connection of thermal power installations are detailed. A detailed description of their operation is given. The advantages of using pulse coolant circulation in substations are shown. The materials reflecting the results of the technical implementation and practical introduction of this technology are presented. Conclusion Theoretical analysis of the operation of the basic schemes of substations with pulse coolant circulation and the results of their practical application, as well as the materials of scientific works devoted to the use of the energy of a hydraulic impact and the study of the effect of pulse coolant flow on thermal and hydrodynamic processes, have yielded a combination of factors reflecting technical and economic rationality of application of pulse coolant circulation. 

  10. Investigation of Planar Waveguides and Components for Millimeter-Wave Integrated Circuits.

    Science.gov (United States)

    1985-03-01

    cntant of fin-line in a WR-28 shield. ..................................... 1A 16 rmodea. 3 .C am - - *’% *jrrode 2 0 M0 e I 2.8. . .4 8.0 3.8 4.6...line with a WR-62 shield. I1 L2- 7.773 mm, 2d -0.254 mm, e,. -2.2, 2b -3.95 mm w1 3 mm, m -17, n -33. 66 0. am - char. fn. 𔃺 0. am 0.36 0.35 6.40...communication trasmitter and receiver system using dielectric waveguide integrated circuits," IEEE Trans. Microwave Theory Tech, vol. M’T-24, pp. 797-803, Nov

  11. Water chemistry and corrosion control of cladding and primary circuit components. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1999-12-01

    Corrosion is the principal life limiting degradation mechanism in nuclear steam supply systems, especially taking into account the trends to increase fuel burnup, thermal rate and cycle length. Primary circuit components of water cooled power reactors have an impact on Zr-based alloys behaviour due to crud (primary circuit corrosion products) formation, transport and deposition on heat transfer surfaces. Crud deposits influence water chemistry, radiation and thermal hydraulic conditions near cladding surface, and by this way-Zr-based alloy corrosion. During the last decade, significant improvements were achieved in the reduction of the corrosion and dose rates by changing the cladding material for one more resistant to corrosion or by the improvement of water chemistry conditions. However, taking into account the above mentioned tendency for heavier fuel duties, corrosion and water chemistry, control will remain a serious task to work with for nuclear power plant operators and scientists, as well as development of generally accepted corrosion model of Zr-based alloys in a water environment in a new millennium. Upon the recommendation of the International Working Group on Water Reactor Fuel Performance and Technology, water chemistry and corrosion of cladding and primary circuit components are in the focus of the IAEA activities in the area of fuel technology and performance. At present the IAEA performs two co-ordinated research projects (CRPs): on On-line High Temperature Monitoring of Water Chemistry and Corrosion (WACOL) and on Activity Transport in Primary Circuits. Two CRPs deal with hydrogen and hydride degradation of the Zr-based alloys. A state-of-the-art review entitled: 'Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants' was published in 1998. Technical Committee meetings on the subject were held in 1985 (Cadarache, France), 1989 (Portland, USA), 1993 (Rez, Czech Republic). During the last few years extensive exchange of experience in

  12. Correlation between Ni base alloys surface conditioning and cation release mitigation in primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Clauzel, M.; Guillodo, M.; Foucault, M. [AREVA NP SAS, Technical Centre, Le Creusot (France); Engler, N.; Chahma, F.; Brun, C. [AREVA NP SAS, Chemistry and Radiochemistry Group, Paris La Defense (France)

    2010-07-01

    The mastering of the reactor coolant system radioactive contamination is a real stake of performance for operating plants and new builds. The reduction of activated corrosion products deposited on RCS surfaces allows minimizing the global dose integrated by workers which supports the ALARA approach. Moreover, the contamination mastering limits the volumic activities in the primary coolant and thus optimizes the reactor shutdown duration and environment releases. The main contamination sources on PWR are due to Co-60 and Co-58 nuclides which come respectively Co-59 and Ni-58, naturally present in alloys used in the RCS. Co is naturally present as an impurity in alloys or as the main component of hardfacing materials (Stellites™). Ni is released mainly by SG tubes which represent the most important surface of the RCS. PWR steam generators (SG), due to the huge wetted surface are the main source of corrosion products release in the primary coolant circuit. As corrosion products may be transported throughout the whole circuit, activated in the core, and redeposited all over circuit surfaces, resulting in an increase of activity buildup, it is of primary importance to gain a better understanding of phenomenon leading to corrosion product release from SG tubes before setting up mitigation measures. Previous studies have shown that SG tubing made of the same material had different release rates. To find the origin of these discrepancies, investigations have been performed on tubes at the as-received state and after exposure to a nominal primary chemistry in titanium recirculating loop. These investigations highlighted the existence of a correlation between the inner surface metallurgical properties and the release of corrosion products in primary coolant. Oxide films formed in nominal primary chemistry are always protective, their morphology and their composition depending strongly on the geometrical, metallurgical and physico-chemical state of the surface on which they

  13. Parametric Study of Time-Dependent Corrosion Product Activity due to 56Mn, 58Co, and 60Co in the Primary Coolant Circuit of a Typical Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Muhammad Rafique

    2015-01-01

    Full Text Available Results of a detailed study, based on the parametric analysis of activated corrosion products, in primary coolant of a typical pressurized water reactor (PWR are presented. The parameters influencing time dependent buildup of corrosion product activity (CPA in primary coolant loop of PWR were identified. The computer program CPAIR was used to accommodate for time dependent corrosion rates. The behaviors of 56Mn, 58Co, and 60Co were studied over the reactor operational time. During the course of normal operation of reactor, the CPA is dominated by 56Mn, while 58Co and 60Co are the predominant radionuclides after reactor shutdown. Parametric study suggests that the total CPA is most sensitive to ion-exchanger removal rates. For a removal rate of 300 cm3-s−1, the specific activity due to 56Mn has the maximum value of 3.552 × 104 Bq-m−3 after 1,000 hours of reactor operation. This value decreases drastically to 8.325 × 103 Bq-m−3 at removal rate of 900 cm3-s−1. Additionally, CPA due to 56Mn, 58Co, and 60Co shows strong dependence on removal rates from the core material surfaces. Variations in the values of radionuclide removal rates from piping surface and radionuclide removal rate from deposition on pipes showed only very small effects on CPA buildup.

  14. Estimative of core damage frequency in IPEN'S IEA-R1 research reactor due to the initiating event of loss of coolant caused by large rupture in the pipe of the primary circuit

    International Nuclear Information System (INIS)

    Hirata, Daniel Massami; Sabundjian, Gaiane; Cabral, Eduardo Lobo Lustosa

    2009-01-01

    The National Commission of Nuclear Energy (CNEN), which is the Brazilian nuclear regulatory commission, imposes safety and licensing standards in order to ensure that the nuclear power plants operate in a safe way. For licensing a nuclear reactor one of the demands of CNEN is the simulation of some accidents and thermalhydraulic transients considered as design base to verify the integrity of the plant when submitted to adverse conditions. The accidents that must be simulated are those that present large probability to occur or those that can cause more serious consequences. According to the FSAR (Final Safety Analysis Report) the initiating event that can cause the largest damage in the core, of the IEA-R1 research reactor at IPEN-CNEN/SP, is the LOCA (Loss of Coolant Accident). The objective of this paper is estimate the frequency of the IEA-R1 core damage, caused by this initiating event. In this paper we analyze the accident evolution and performance of the systems which should mitigate this event: the Emergency Coolant Core System (ECCS) and the isolated pool system. They will be analyzed by means of the event tree. In this work the reliability of these systems are also quantified using the fault tree. (author)

  15. Primary coolant pH for control of CANDU plant aging

    Energy Technology Data Exchange (ETDEWEB)

    Burrill, K.A.; Cheluget, E.L.; Miller, D.G.; Turner, C.W

    1998-12-01

    Plant aging can be defined as any degradation with time of system performance that increases the operator's difficulty in maintaining operation within design specification. Degradation can be a physical change in a component (e.g., surface roughness), or a change in operating condition (e.g., Reactor Inlet Header Temperature (RIHT) rise). This paper focuses on the corrosion of the carbon steel piping in the CANDU primary circuit and the aging issues that arise. In one approach, a small reduction in the coolant pH has been recommended to operating plants that will slow those aging issues driven by dissolved iron transport around the primary circuit. Secondly, chemical decontamination of the entire Heat Transport System (HTS) can be carried out as a single process application step, or it can be performed following cleaning of the steam generators. (author)

  16. Reactor coolant pump flywheel

    Science.gov (United States)

    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.

  17. Packaging strategies for printed circuit board components. Volume I, materials & thermal stresses.

    Energy Technology Data Exchange (ETDEWEB)

    Neilsen, Michael K. (Kansas City Plant, Kansas City, MO); Austin, Kevin N.; Adolf, Douglas Brian; Spangler, Scott W.; Neidigk, Matthew Aaron; Chambers, Robert S.

    2011-09-01

    Decisions on material selections for electronics packaging can be quite complicated by the need to balance the criteria to withstand severe impacts yet survive deep thermal cycles intact. Many times, material choices are based on historical precedence perhaps ignorant of whether those initial choices were carefully investigated or whether the requirements on the new component match those of previous units. The goal of this program focuses on developing both increased intuition for generic packaging guidelines and computational methodologies for optimizing packaging in specific components. Initial efforts centered on characterization of classes of materials common to packaging strategies and computational analyses of stresses generated during thermal cycling to identify strengths and weaknesses of various material choices. Future studies will analyze the same example problems incorporating the effects of curing stresses as needed and analyzing dynamic loadings to compare trends with the quasi-static conclusions.

  18. The use of the acoustic emission for the components of the primary circuit of the nuclear power plants

    International Nuclear Information System (INIS)

    Svoboda, V.

    1992-01-01

    Full text: The Modrany Engineering Works (Modranske strojirny) is a producer and a final supplier of the main connecting piping circuit systems and valves for the nuclear power plants (type VVER 440 and VVER 1000) built in Czechoslovakia. Besides the delivery and assembly of valves and components methods there were developed for a monitoring of the stated equipment ability of a service in the Material and Diagnostic Laboratory, which is a part of the company. An important object of this work is to obtain a sufficient set of data and to work out suitable methods, on the basis of which it would be possible to perform a serious estimation of residual service life of the main piping components after certain service operation of the nuclear power plant. During the operation of a nuclear power station a failure of the main piping circuit could happen in either of two possible modes: 1.) A sudden break - by an unstable defect propagation leading to a. final fracture of the piping; 2) A fatigue failure - which is characterised by a gradual subcritical growth of defect in relation to the loading parameters. This process is frequently accelerated by further processes, e.g. corrosion. It is therefore suitable to use such physical and mechanical quantities, which characterize the material damage. Acoustic emission signals belongs to these quantities. A knowledge of the response of these signals in relation to the damage of the material gives us the possibility to evaluate the residual life of the piping containing defects. The importance of this is increasing mainly after a long period of service. She paper deals in details with experience gained in application of acoustic emission, during pressure tests of primary circuit components (elbow, welds, T- junction etc) in laboratory conditions which imitate those in service. There are shown some results of cyclic fatigue tests by internal pressure on prototypes models and specimen. Acoustic emission method represents the

  19. Analytical modeling of multi-layered printed circuit board using multi-stacked via clusters as component heat spreader

    Directory of Open Access Journals (Sweden)

    Monier-Vinard Eric

    2016-01-01

    Full Text Available In order to help the electronic designer to early determine the limits of the power dissipation of electronic component, an analytical model was established to allow a fast insight of relevant design parameters of a multi-layered electronic board constitution. The proposed steady-state approach based on Fourier series method promotes a practical solution to quickly investigate the potential gain of multi-layered thermal via clusters. Generally, it has been shown a good agreement between the results obtained by the proposed analytical model and those given by electronics cooling software widely used in industry. Some results highlight the fact that the conventional practices for Printed Circuit Board modeling can be dramatically underestimate source temperatures, in particular with smaller sources. Moreover, the analytic solution could be applied to optimize the heat spreading in the board structure with a local modification of the effective thermal conductivity layers.

  20. The cooling of printed circuit board mounted components using copper ladder heat conduction to a cold wall

    Science.gov (United States)

    Dale, I. C.

    1982-09-01

    A series of experimental tests, designed to investigate the cooling of printed circuit board (PCB) mounted dual-in-line (DIL) components within an avionic box using the copper ladder/cold wall technique is described. Areas of investigation include avionic box orientation, side wall conduction, top plate finning, mixed air-wash, avionic power reduction, cooling air temperature reduction, cooling air mass flow rate reduction, cold wall heat pick-up and avionic box insulation. Results were obtained from thermocouple temperature measurements. The use of an aluminum alloy interplate to cool two adjacent PCBs is discussed. Results in graphic form are included together with a list of conclusions on the effects of all the major parameters considered.

  1. Component-Level Electronic-Assembly Repair (CLEAR) Spacecraft Circuit Diagnostics by Analog and Complex Signature Analysis

    Science.gov (United States)

    Oeftering, Richard C.; Wade, Raymond P.; Izadnegahdar, Alain

    2011-01-01

    The Component-Level Electronic-Assembly Repair (CLEAR) project at the NASA Glenn Research Center is aimed at developing technologies that will enable space-flight crews to perform in situ component-level repair of electronics on Moon and Mars outposts, where there is no existing infrastructure for logistics spares. These technologies must provide effective repair capabilities yet meet the payload and operational constraints of space facilities. Effective repair depends on a diagnostic capability that is versatile but easy to use by crew members that have limited training in electronics. CLEAR studied two techniques that involve extensive precharacterization of "known good" circuits to produce graphical signatures that provide an easy-to-use comparison method to quickly identify faulty components. Analog Signature Analysis (ASA) allows relatively rapid diagnostics of complex electronics by technicians with limited experience. Because of frequency limits and the growing dependence on broadband technologies, ASA must be augmented with other capabilities. To meet this challenge while preserving ease of use, CLEAR proposed an alternative called Complex Signature Analysis (CSA). Tests of ASA and CSA were used to compare capabilities and to determine if the techniques provided an overlapping or complementary capability. The results showed that the methods are complementary.

  2. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

    A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

  3. Development of an ultrasonic inspection technique for the Sizewell B PWR cast austenitic reactor coolant pump bowl

    International Nuclear Information System (INIS)

    Gilroy, K.S.

    1988-01-01

    The CEGB has recently received approval to build a PWR at Sizewell in Suffolk. The CEGB's Non-Destructive Testing Applications Centre is developing ultrasonic inspection techniques for those components of the primary circuit whose failure is claimed to be incredible. For the welds and castings that fall into this category, the ultrasonic inspections will be validated using test-blocks of representative macrostructure and geometry, and containing realistic defects. This paper describes the research leading to the development of an ultrasonic inspection technique for the near-surface regions of the cast austenitic reactor coolant pump bowls. (author)

  4. Development of lead-bismuth coolant technology for nuclear device

    International Nuclear Information System (INIS)

    Kamata, Kin-ya; Kitano, Teruaki; Ono, Mikinori

    2004-01-01

    Liquid lead-bismuth is a promising material as a future fast reactor coolant or an intensive neutron source material for accelerator driven transmutation system (ADS). To develop nuclear plants and their installations using lead-bismuth coolant for practical use, both coolant technologies, inhabitation process of steels and quality control of coolant, and total operation system for liquid lead-bismuth plants are required. Based on the experience of liquid metal coolant, Mitsui Engineering and Shipbuilding Co., Ltd. (MES) has completed the liquid lead-bismuth forced circulation loop and has acquired various engineering data on main components including economizer. As a result of tis operation, MES has developed key technologies of lead-bismuth coolant such as controlling of oxygen content in lead-bismuth and a purification of lead-bismuth coolant. MES participated in the national project, ''The Development of Accelerator Driven Transmutation System'', together with JAERI (Japan Atomic Energy Research Institute) and started corrosion test for beam window of ADS. (author)

  5. Transient behaviour of main coolant pump in nuclear power plants

    International Nuclear Information System (INIS)

    Delja, A.

    1986-01-01

    A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)

  6. Design and test of component circuits of an integrated quantum voltage noise source for Johnson noise thermometry

    Science.gov (United States)

    Yamada, Takahiro; Maezawa, Masaaki; Urano, Chiharu

    2015-11-01

    We present design and testing of a pseudo-random number generator (PRNG) and a variable pulse number multiplier (VPNM) which are digital circuit subsystems in an integrated quantum voltage noise source for Jonson noise thermometry. Well-defined, calculable pseudo-random patterns of single flux quantum pulses are synthesized with the PRNG and multiplied digitally with the VPNM. The circuit implementation on rapid single flux quantum technology required practical circuit scales and bias currents, 279 junctions and 33 mA for the PRNG, and 1677 junctions and 218 mA for the VPNM. We confirmed the circuit operation with sufficiently wide margins, 80-120%, with respect to the designed bias currents.

  7. Maintenance of the primary and secondary circuit chemistry - The key of the minimum corrosion of metallic components

    International Nuclear Information System (INIS)

    Radulescu, M.; Pirvan, I.

    2008-01-01

    The objective of this paper is the presentation of some chemistry regimes available in primary and secondary circuit of a CANDU Nuclear Power Station simultaneously with the establishment of some correlations between the corrosion behavior of some representative metallic materials and the specific parameters of the respective aqueous environments. First we reviewed the main chemical parameters of the aqueous environments from the primary and secondary circuit in a CANDU NPS and finally we established some correlations between these chemistry variations and the quality of superficial films formed on some representative alloys such as zirconium alloys and carbon steels from primary and secondary circuit, respectively. To characterize the filmed samples the following methods were used: the gravimetric, X-rays diffraction, metallographic microscopy method and some electrochemical methods such as: potentiodynamic and electrochemical impedance spectroscopy. Using these methods, there were established the corrosion kinetics and the characteristics of the films formed on the samples in the NPS operation conditions

  8. Sodium as a reactor coolant

    International Nuclear Information System (INIS)

    Cesar, S.B.G.

    1989-01-01

    This work is related to the use of sodium as a reactor coolant, to the advantages and problems related to its use, its mechanical, thermophysics, eletronical, magnetic and nuclear properties. It is mainly a bibliographic review, with the aim of gathering the necessary information to persons initiating in the study of sodium and also as reference source. (author) [pt

  9. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  10. Sodium coolant of fast reactors: Experience and problems

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Volchkov, L.G.; Drobyshev, A.V.; Nikulin, M.P.; Kochetkov, L.A.; Alexeev, V.V.

    1997-01-01

    In present report the following subjects are considered: state of the coolant and sodium systems under normal operating condition as well as under decommissioning, disclosing of sodium circuits and liquidation of its consequences, cleaning from sodium and decontamination under repairing works of equipment and circuits. Cleaning of coolant and sodium systems under normal operating conditions and under accident contamination. Cleaning of the equipment under repairing works and during decommissioning from sodium and products of its interaction with water and air. Treatment of sodium waste, taking into account a possibility of sodium fires. It is shown that the state of coolant, cover gas, surfaces of constructive materials which are in contact with them, cleaning systems, formed during installation operation require development of specific technologies. Developed technologies ensured safety operation of sodium cooled installations as in normal operating conditions so in abnormal situations. R and D activities in this field and experience gained provided a solid base for coping with problems arising during decommissioning. Prospective research problems are emphasized where the future efforts should be concentrated in order to improve characteristics of sodium cooled reactors and to make their decommissioning optimal and safe. (author)

  11. Automatic ultrasonic pre-service, and in-service inspection of pressurized components of the primary circuit of nuclear power stations

    International Nuclear Information System (INIS)

    Muller, G.P.; Hallermeier, L.; Heinrich, D.; Grabendorfer, W.; Rebrmann, M.

    1985-01-01

    Ultrasonic pre-service and especially in-service inspection activities on the primary circuit of nuclear power stations form an essential part of the maintenance work that must be performed throughout the lifetime to ensure plant integrity. Consequently, the equipment required to carry out these inspections must be continuously improved in respect of reliability, safety, accuracy and ease of handling in order to minimize disturbances and repairs and reduce radiation exposure of the personnel. The authors' discussion of technique, equipment and performance of automated ultrasonic inspection is based on 15 years of experience in the testing of components of the primary circuit in nuclear power stations. To cover all inspection areas of the RPV of a PWR, four different manipulators are required, two for the closure head, one for the studs and one for the cylindrical shell and bottom closure. The use of the newly developed equipment, which naturally meets all the recommendations of the licensing authorities, allows for the automatic inspection of the components of primary circuit of nuclear power stations and the thus helps to substantially decrease the radiation exposure of the personnel. All the manipulators and their control consoles were designed and manufactured by M.A.N., Nuremberg while the ultrasonic electronic system was developed by Krautkramer, Cologne

  12. Design, R and D, and manufacture of the components of secondary circuits for fast sodium power plant

    International Nuclear Information System (INIS)

    Avanzini, P.G.; Valentini, P.

    1978-01-01

    The efforts of Italian industry in the field of components for fast sodium power plants are presented. The main components involved are steam generators, sodium/air heat exchangers and large sodium valves. The development work carried out for these components is outlined and the main realizations are presented. A brief account is also given of the current situation of sodium plants supporting the development of these components. Research on materials and the development of special sodium components are also reported upon. (author)

  13. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  14. Full reactor coolant system chemical decontamination qualification programs

    Energy Technology Data Exchange (ETDEWEB)

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-03-01

    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  15. Analyses of Decrease in Reactor Coolant Flow Rate in SMART

    International Nuclear Information System (INIS)

    Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

    2011-01-01

    SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

  16. The circuit designer's companion

    CERN Document Server

    Williams, Tim

    1991-01-01

    The Circuit Designer's Companion covers the theoretical aspects and practices in analogue and digital circuit design. Electronic circuit design involves designing a circuit that will fulfill its specified function and designing the same circuit so that every production model of it will fulfill its specified function, and no other undesired and unspecified function.This book is composed of nine chapters and starts with a review of the concept of grounding, wiring, and printed circuits. The subsequent chapters deal with the passive and active components of circuitry design. These topics are foll

  17. Coolant clean-up and recycle systems

    International Nuclear Information System (INIS)

    Ito, Takao.

    1979-01-01

    Purpose: To increase the service life of mechanical seals in a shaft sealing device, eliminate leakages and improve the safety by providing a recycle pump for feeding coolants to a coolant clean-up device upon reactor shut-down and adapting the pump treat only low temperature and low pressure coolants. Constitution: The system is adapted to partially take out coolants from the pipeways of a recycling pump upon normal operation and feed them to a clean-up device. Upon reactor shut-down, the recycle pump is stopped and coolants are extracted by the recycle pump for shut-down into the clean-up device. Since the coolants are not fed to the clean-up device by the recycle pump during normal operation as conducted so far, high temperature and high pressure coolants are not directly fed to the recycle pump, thereby enabling to avoid mechanical problems in the pump. (Kamimura, M.)

  18. Water quality estimation method for primary coolant circuit

    International Nuclear Information System (INIS)

    Wada, Yoichi; Ibe, Hidefumi.

    1994-01-01

    The present invention is suitable to water quality diagnosis at each of the portions in a reactor upon hydrogen injection for preventing stress corrosion crackings (SCC) of a BWR type reactor. That is, a plurality of simulations are conducted how the water quality at each of the portions in the reactor is changed when hydrogen injection amount is changed depending on the design and operation conditions of the plant. The result of the calculation is stored in a memory device. A water quality distribution in a pressure vessel having a solution which agrees with a value actually measured by a water quality measuring device disposed at the outside of a reactor core is retrieved from the results of the calculation. If no agreeing solution can be found, water quality distribution containing the actually measured value is determined based on the result of the calculation by using interpolation. In the present invention, the result of the calculation obtained by the simulation and the actually measured value at the outside of the reactor core can be utilized, to map the distribution of reactor water ingredients on a screen, which can accurately estimate the water quality at the periphery of the reactor core on real time. As a result, an operational efficiency of a reactor which can control water quality upon hydrogen injection at an optimum condition. (I.S.)

  19. Decontamination between dismantling of the Rapsodie primary coolant circuit

    International Nuclear Information System (INIS)

    Costes, J.R.; Gauchon, J.P.; Antoine, P.

    1991-01-01

    The large-scale decontamination of FBR sodium loops is a novel task, as only a limited number of laboratory-scale results are available to date. The principal objective of this work is to develop a suitable decontamination procedure for application to the primary loops of the RAPSODIE fast breeder reactor as part of decommissionning to Stage 2

  20. Dual Regenerative Cooling Circuits for Liquid Rocket Engines (POSTPRINT)

    National Research Council Canada - National Science Library

    Naraghi, N. H; Dunn, S; Coats, D

    2006-01-01

    .... Two engines, the SSME and a RP1-LOX engine, are retrofitted with dual-circuits. It is shown that the maximum wall temperatures for both engines are substantially reduced while also lowering coolant pumping power...

  1. Test facility for investigation of heat transfer of promising coolants for the nuclear power industry

    Science.gov (United States)

    Belyaev, I. A.; Sviridov, V. G.; Batenin, V. M.; Biryukov, D. A.; Nikitina, I. S.; Manchkha, S. P.; Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, E. V.

    2017-11-01

    The results are presented of experimental investigations into liquid metal heat transfer performed by the joint research group consisting of specialist in heat transfer and hydrodynamics from NIU MPEI and JIHT RAS. The program of experiments has been prepared considering the concept of development of the nuclear power industry in Russia. This concept calls for, in addition to extensive application of water-cooled, water-moderated (VVER-type) power reactors and BN-type sodium cooled fast reactors, development of the new generation of BREST-type reactors, fusion power reactors, and thermonuclear neutron sources. The basic coolants for these nuclear power installations will be heavy liquid metals, such as lead and lithium-lead alloy. The team of specialists from NRU MPEI and JIHT RAS commissioned a new RK-3 mercury MHD-test facility. The major components of this test facility are a unique electrical magnet constructed at Budker Nuclear Physics Institute and a pressurized liquid metal circuit. The test facility is designed for investigating upward and downward liquid metal flows in channels of various cross-sections in a transverse magnetic field. A probe procedure will be used for experimental investigation into heat transfer and hydrodynamics as well as for measuring temperature, velocity, and flow parameter fluctuations. It is generally adopted that liquid metals are the best coolants for the Tokamak reactors. However, alternative coolants should be sought for. As an alternative to liquid metal coolants, molten salts, such as fluorides of lithium and beryllium (so-called FLiBes) or fluorides of alkali metals (so-called FLiNaK) doped with uranium fluoride, can be used. That is why the team of specialists from NRU MPEI and JIHT RAS, in parallel with development of a mercury MHD test facility, is designing a test facility for simulating molten salt heat transfer and hydrodynamics. Since development of this test facility requires numerical predictions and verification

  2. Organic coolant for ARIES-III

    International Nuclear Information System (INIS)

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

    1991-04-01

    ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 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

  3. Activity in neurons of a putative protocerebral circuit representing information about a ten component plant odour blend in Heliothis virescens.

    Directory of Open Access Journals (Sweden)

    Bjarte Bye Løfaldli

    2012-09-01

    Full Text Available The olfactory pathway in the insect brain is anatomically well described from the antennal lobe to the mushroom bodies and the lateral protocerebrum in several species. Less is known about the further connections of the olfactory network in protocerebrum and how information about relevant plant odorants and mixtures are represented in this network, resulting in output information mediated by descending neurons. In the present study we have recorded intracellularly followed by dye injections from neurons in the lateral- and superior protocerebrum of the moth, Heliothis virescens. As relevant stimuli, we have used selected primary plant odorants and mixtures of them. The results provide the morphology and physiological responses of neurons involved in a putative circuit connecting the mushroom body lobes, the superior and the lateral protocerebrum, as well as input to superior and lateral protocerebrum by one multiglomerular antennal lobe neuron and output from the lateral protocerebrum by one descending neuron. All neurons responded to one particular mixture of ten primary plant odorants, some of them also to single odorants of the mixture. Altogether, the physiological data indicate integration in protocerebral neurons of information from several of the receptor neuron types functionally described in this species.

  4. Physical properties of organic coolants

    International Nuclear Information System (INIS)

    Debbage, A.G.; Garton, D.A.; Kinneir, J.H.

    1963-03-01

    Density, viscosity, specific heat, vapour pressure and calorific value were measured within the temperature range 100 - 400 deg C for mixtures of Santowax R with pyrolytic high boiler and Santowax R with O.M.R.E. radiolytic high boiler; in addition measurements were made on Santowax OM, X-7 standard, X-7 loop coolant and O.M.R.E. coolant supplied by Atomic Energy of Canada Ltd. The accuracy of the measurements made were density (± 1/4%), viscosity (± 2%), specific heat (± 2%), vapour pressure (± 2%) and calorific value (± 1/2%). Thermal conductivity was calculated from an improved form of the Smiths equation with an accuracy within ± 6%. Equations fitted to the vapour pressure results were used to provide data outside the experimental range for burnout correlation purposes. The general effect of high boiler content on the specific heat and calorific values was small. The differences in physical property values for corresponding values of either pyrolytic or radiolytic high boiler were small for density (0.3%) and specific heat (2%), but quite large for viscosity (70%) with the pyrolytic high boiler mixture giving the higher value. The chemical analysis of all materials was based on gas chromatography and the relationship between this and an earlier distillation method established. (author)

  5. Cleaning of aluminum after machining with coolants

    International Nuclear Information System (INIS)

    Roop, B.

    1992-01-01

    An x-ray photoemission spectroscopic study was undertaken to compare the cleaning of the Advanced Photon Source (APS) aluminum extrusion storage ring vacuum chambers after machining with and without water soluble coolants. While there was significant contamination left by the coolants, the cleaning process was capable of removing the residue. The variation of the surface and near surface composition of samples machined either dry or with coolants was negligible after cleaning. The use of such coolants in the machining process is therefore recommended

  6. A Component-Minimized Single-Phase Active Power Decoupling Circuit with Reduced Current Stress to Semiconductor Switches

    DEFF Research Database (Denmark)

    Tang, Yi; Blaabjerg, Frede

    2015-01-01

    component, e.g. inductors or film capacitors for ripple energy storage because this task can be accomplished by the dc-link capacitors, and therefore its implementation cost can be minimized. Another unique feature of the proposed topology is that the current stress of power semiconductors can be reduced...

  7. Sodium confluent rates of flow values, on 0,5 mixer, of a sodium italian SS-050 circuit component

    International Nuclear Information System (INIS)

    Walsh, L.M.

    1987-01-01

    Sodium lines on different temperatures, during an emergency drainage on 0,5 mixer was found. To future valuation by DIMEC of tensions that occurs on that component of SS-050, the confluent rates of flow values were calculated. (L.M.J.) [pt

  8. Modular core component support for nuclear reactor

    International Nuclear Information System (INIS)

    Finch, L.M.; Anthony, A.J.

    1975-01-01

    The core of a nuclear reactor is made up of a plurality of support modules for containing components such as fuel elements, reflectors and control rods. Each module includes a component support portion located above a grid plate in a low-pressure coolant zone and a coolant inlet portion disposed within a module receptacle which depends from the grid plate into a zone of high-pressure coolant. Coolant enters the module through aligned openings within the receptacle and module inlet portion and flows upward into contact with the core components. The modules are hydraulically balanced within the receptacles to prevent expulsion by the upward coolant forces. (U.S.)

  9. SMART core power control method by coolant temperature variation

    International Nuclear Information System (INIS)

    Lee, Chung Chan; Cho, Byung Oh

    2001-08-01

    SMART is a soluble boron-free integral type pressurized water reactor. Its moderator temperature coefficient (MTC) is strongly negative throughout the cycle. The purpose of this report is how to utilize the primary coolant temperature as a second reactivity control system using the strong negative MTC. The reactivity components associated with reactor power change are Doppler reactivity due to fuel temperature change, moderator temperature reactivity and xenon reactivity. Doppler reactivity and moderator temperature reactivity take effects almost as soon as reactor power changes. On the other hand, xenon reactivity change takes more than several hours to reach an equilibrium state. Therefore, coolant temperature at equilibrium state is chosen as the reference temperature. The power dependent reference temperature line is limited above 50% power not to affect adversely in reactor safety. To compensate transient xenon reactivity, coolant temperature operating range is expanded. The suggested coolant temperature operation range requires minimum control rod motion for 50% power change. For smaller power changes such as 25% power change, it is not necessary to move control rods to assure that fuel design limits are not exceeded

  10. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

    Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

  11. Nonlinear dynamic analysis of nuclear reactor primary coolant systems

    International Nuclear Information System (INIS)

    Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

    1979-01-01

    The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

  12. Reactor having coolant recycling pump

    International Nuclear Information System (INIS)

    Goto, Tadashi; Karatsuka, Shigeki; Yamamoto, Hajime.

    1991-01-01

    In a coolant recycling pump for an LMFBR type reactor, vertical grooves are formed to a static portion which surrounds a pump shaft as far as the lower end thereof. Sodium mists present in an annular gap of the pump shaft form a rotational flow, lose its centrifugal force at the grooved portion and are collected positively to the grooved portion. Further, since the rotational flow in the grooved channel is in a state of a cavity flow, the pressure is released in the grooved portion and a secondary eddy current is formed thereby providing a depressurized state. Accordingly, by a synergestic effect of the centrifugal force and the cavity flow, sodium mists can be recovered completely. (T.M.)

  13. Method for investigation of various iodine species in the primary coolant of the nuclear power plant in Paks

    International Nuclear Information System (INIS)

    Volent, G.; Gimesi, O.; Solymosi, J.

    1996-01-01

    Iodine isotopes formed in the course of fission in nuclear reactors may be present in the primary coolant in different oxidation states, i.e., in different chemical forms. It is important to know the chemical forms and their proportions in order to asses the environmental effect of the emitted iodine and the performance of air filters used in the primary circuit for binding iodine, species, since both depend on the chemical forms in which it is present. Volatile components were separated from water samples taken separately from each block of the nuclear power station by purging with inert gas, then the aerosol, iodine vapour and alkyl iodides were selectively bound on the filter system of the 'KOMBI' sampler. I 3 - , I - , IO - , IO 3 - and IO 4 - left in the aqueous phase after purging were separated by consecutive physical and chemical procedures (extraction, isotope exchange, reduction). The results of the investigations have shown that the water technology used in the Nuclear Power Plant in Paks is appropriate with respect to the radioiodine balance. Iodine was found to be predominant species, and no volatile iodine species were found to be present in the primary coolant. Volatile iodine species sometimes appearing in emissions may be formed from leaching waters due to secondary effects. (author)

  14. Mathematical model of the reactor coolant pump

    International Nuclear Information System (INIS)

    Kozuh, M.

    1989-01-01

    The mathematical model of reactor coolant pump is described in this paper. It is based on correlations for centrifugal reactor coolant pumps. This code is one of the elements needed for the simulation of the whole NPP primary system. In subroutine developed according to this model we tried in every possible detail to incorporate plant specific data for Krsko NPP. (author)

  15. Analog circuits cookbook

    CERN Document Server

    Hickman, Ian

    2013-01-01

    Analog Circuits Cookbook presents articles about advanced circuit techniques, components and concepts, useful IC for analog signal processing in the audio range, direct digital synthesis, and ingenious video op-amp. The book also includes articles about amplitude measurements on RF signals, linear optical imager, power supplies and devices, and RF circuits and techniques. Professionals and students of electrical engineering will find the book informative and useful.

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Flow boiling test of GDP replacement coolants

    International Nuclear Information System (INIS)

    Park, S.H.

    1995-01-01

    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 4 F 10 and C 4 F 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 4 F 10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C 4 F 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

  20. Flow boiling test of GDP replacement coolants

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.H. [comp.

    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.

  1. Testing methods of gaseous admixtures in HLMC circuits

    International Nuclear Information System (INIS)

    Shelemet'ev, V.M.; Martynov, P.N.; Askhadullin, R.Sh.; Storozhenko, A.N.; Sadovnichij, R.P.; Ivanov, I.I.

    2014-01-01

    Control of gas phase is the effective method for state diagnostics of circuit of nuclear power facilities with heavy liquid metal coolants. Use of developing in IPPE solid electrolyte and conductometric oxygen and hydrogen sensors, which are set directly in gas system of the primary circuit, allows to maintain continuously control of oxygen and hydrogen content as well as operational efficiency and accuracy of these parameters determination under various situations related with oxygen and hydrogen insertion into circuit. Sensors ensure long-term safe operation under extreme conditions of high temperatures, pressures, humidity, etc., and are advanced devices for application in nuclear power facilities with heavy liquid metal coolants [ru

  2. Applied model of through-wall crack of coolant vessels of WWER-type reactors

    International Nuclear Information System (INIS)

    Petrosyan, V.; Hovakimyan, T.; Vardanyan, M.; Khachatryan, A.; Minasyan, K.

    2010-01-01

    We propose an applied-model of Through-Wall Crack (TWC) for WWER-type units primary vessels. The model allows to simulate the main morphological parameters of real TWC, i.e. length, area of inlet and outlet openings, channel depth and small and large size unevenness of the crack surface. The model can be used for developing and improving the coolant-leak detectors for the primary circuit vessels of WWER-units. Also, it can be used for research of the coolant two-phase leakage phenomenon through narrow cracks/channels and thermo-physical processes in heat-insulation layer of the Main Coolant Piping (MCP) during the leak

  3. Electronic devices and circuits

    CERN Document Server

    Pridham, Gordon John

    1968-01-01

    Electronic Devices and Circuits, Volume 1 deals with the design and applications of electronic devices and circuits such as passive components, diodes, triodes and transistors, rectification and power supplies, amplifying circuits, electronic instruments, and oscillators. These topics are supported with introductory network theory and physics. This volume is comprised of nine chapters and begins by explaining the operation of resistive, inductive, and capacitive elements in direct and alternating current circuits. The theory for some of the expressions quoted in later chapters is presented. Th

  4. Definition of loss-of-coolant accident radiation source

    International Nuclear Information System (INIS)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist

  5. A Collector Geometry Impact on the Coolant Flow Distribution in the Reactor Model Core

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2015-01-01

    Full Text Available In creating the reactor facility for the transport and energy module of a megawatt class the important task is to optimize a coolant flow path, i.e. to provide a moderate flow resistance and uniform distribution of a coolant. A kind of the chosen collector design to supply coolant significantly contributes to hydraulic losses, in particular, the porosity of the inlet lattice which may lead to uneven coolant rate at the inlet, flow pulsations, and hydraulic losses.For the first time in domestic practice the work examines an impact of the inlet lattices geometry on the averaged and pulsating flow both in a hemispherical collector and at the core inlet to the model paths of a reactor gas-cooled coolant, and gives advices on optimization of collector paths of the coolant flow.The paper presents the results of experiments carried out on the gas dynamic model of the coolant paths containing the inlet lattices of different porosity. It offers a numerical simulation of the flow in the two-parameter model using k-ε turbulence model and ANSYS CFX v14.0 software package and demonstrates a compliance of experimental data with numerical results.The obtained results show that the inlet lattice with a porosity of 0.25 allows relative leveling of the coolant flow directly at the core inlet, which for a uniform cross-sectional energy release reduces temperature of fuel elements. The considered options of design solutions allow you to select the inlet lattice structure, and the core, as well, according to the porosity parameter to solve the problem of reducing hydraulic losses in the coolant paths, reducing pulsating components of the flow in the core and length of the initial portion of flow stabilization. References

  6. Variable cooling circuit for thermoelectric generator and engine and method of control

    Science.gov (United States)

    Prior, Gregory P

    2012-10-30

    An apparatus is provided that includes an engine, an exhaust system, and a thermoelectric generator (TEG) operatively connected to the exhaust system and configured to allow exhaust gas flow therethrough. A first radiator is operatively connected to the engine. An openable and closable engine valve is configured to open to permit coolant to circulate through the engine and the first radiator when coolant temperature is greater than a predetermined minimum coolant temperature. A first and a second valve are controllable to route cooling fluid from the TEG to the engine through coolant passages under a first set of operating conditions to establish a first cooling circuit, and from the TEG to a second radiator through at least some other coolant passages under a second set of operating conditions to establish a second cooling circuit. A method of controlling a cooling circuit is also provided.

  7. Condition monitoring of main coolant pumps, Dhruva

    International Nuclear Information System (INIS)

    Prasad, V.; Satheesh, C.; Acharya, V.N.; Tikku, A.C.; Mishra, S.K.

    2002-01-01

    Full text: Dhruva is a 100 MW research reactor with natural uranium fuel, heavy water as moderator and primary coolant. Three Centrifugal pumps circulate the primary coolant across the core and the heat exchangers. Each pump is coupled to a flywheel (FW) assembly in order to meet operational safety requirements. All the 3 main coolant pump (MCP) sets are required to operate during operation of the reactor. The pump-sets are in operation since the year 1984 and have logged more than 1,00,000 hrs. Frequent breakdowns of its FW bearings were experienced during initial years of operation. Condition monitoring of these pumps, largely on vibration based parameters, was initiated on regular basis. Break-downs of main coolant pumps reduced considerably due to the fair accurate predictions of incipient break-downs and timely maintenance efforts. An effort is made in this paper to share the experience

  8. Coolant clean-up apparatus for reactors

    International Nuclear Information System (INIS)

    Yonekawa, Yasumasa; Akita, Minoru; Moritani, Kenji.

    1984-01-01

    Purpose: To prevent the occurrence of heat fatigue crackings due to the temperature difference between cleaned water and feed water at the portion on the pipeway where the two fluids are joined. Constitution: In the coolant clean-up system of a BWR type reactor, coolants are taken out from the side of the recycling system pump cooled by way of a regenerative heat exchanger and a non-regenerative heat exchanger and then flow into a filtering desalter for purification. Then, the purified water is joined with the feed water and returned to the inside of the reactor. Coolants for the auxiliary equipment cooler are recycled through non-regenerative heat exchanger and the flow control valve for the coolants is adjusted such that the temperature for the cleaned water accords with that of the feed water, whereby the temperature of the purified water is controlled to eliminate the temperature difference with the feed water. (Kamimura, M.)

  9. Capabilities and Testing of the Fission Surface Power Primary Test Circuit (FSP-PTC)

    Science.gov (United States)

    Garber, Anne E.

    2007-01-01

    An actively pumped alkali metal flow circuit, designed and fabricated at the NASA Marshall Space Flight Center, is currently undergoing testing in the Early Flight Fission Test Facility (EFF-TF). Sodium potassium (NaK), which was used in the SNAP-10A fission reactor, was selected as the primary coolant. Basic circuit components include: simulated reactor core, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, liquid metal flowmeter, load/drain reservoir, expansion reservoir, test section, and instrumentation. Operation of the circuit is based around a 37-pin partial-array core (pin and flow path dimensions are the same as those in a full core), designed to operate at 33 kWt. NaK flow rates of greater than 1 kg/sec may be achieved, depending upon the power applied to the EM pump. The heat exchanger provides for the removal of thermal energy from the circuit, simulating the presence of an energy conversion system. The presence of the test section increases the versatility of the circuit. A second liquid metal pump, an energy conversion system, and highly instrumented thermal simulators are all being considered for inclusion within the test section. This paper summarizes the capabilities and ongoing testing of the Fission Surface Power Primary Test Circuit (FSP-PTC).

  10. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    The invention deals with disengaging the coupling of a reactor coolant pump of a nuclear reactor feeding pressurized coolant. The disengaging coupling has two parts joined by bolts, at least one of them containing a driving agent within a bore. This is provided with a speed-depending ignition device in such manner that, if the critical speed is reached, the driving charge is ignited and the coupling is disengaged by destroying the bolts. (UWI) [de

  11. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Pothier, N.E.; Metcalfe, R.

    1986-06-01

    Large CANDU plants are benefitting from transient-resistant four-year reliable reactor coolant pump seal lifetimes, a direct result of AECL's 20-year comprehensive seal improvement program involving R and D staff, manufacturers, and plant designers and operators. An overview of this program is presented, which covers seal modification design, testing, post-service examination, specialized maintenance and quality control. The relevancy of this technology to Light Water Reactor Coolant Pump Seals is also discussed

  12. Troubleshooting analog circuits

    CERN Document Server

    Pease, Robert A

    1991-01-01

    Troubleshooting Analog Circuits is a guidebook for solving product or process related problems in analog circuits. The book also provides advice in selecting equipment, preventing problems, and general tips. The coverage of the book includes the philosophy of troubleshooting; the modes of failure of various components; and preventive measures. The text also deals with the active components of analog circuits, including diodes and rectifiers, optically coupled devices, solar cells, and batteries. The book will be of great use to both students and practitioners of electronics engineering. Other

  13. Electronic Circuit Analysis Language (ECAL)

    Science.gov (United States)

    Chenghang, C.

    1983-03-01

    The computer aided design technique is an important development in computer applications and it is an important component of computer science. The special language for electronic circuit analysis is the foundation of computer aided design or computer aided circuit analysis (abbreviated as CACD and CACA) of simulated circuits. Electronic circuit analysis language (ECAL) is a comparatively simple and easy to use circuit analysis special language which uses the FORTRAN language to carry out the explanatory executions. It is capable of conducting dc analysis, ac analysis, and transient analysis of a circuit. Futhermore, the results of the dc analysis can be used directly as the initial conditions for the ac and transient analyses.

  14. Secondary coolant circuit for liquid-metal cooled reactor and steam generator for such a circuit

    International Nuclear Information System (INIS)

    Brachet, A.; Figuet, J.; Guidez, J.; Lions, N.; Traiteur, R.; Zuber, T.

    1984-01-01

    An upper buffer tank and downstream buffer tank are disposed inside the steam generators. The downstream briffer tank is annular and it surrounds and communicates with a zone of the steam generator through which the liquid metal flows towards the bottom between the exchange zone and the outlet nozzle. The pressure of the inert gas blanket in the downstream buffer volume is more important than this one in the upper buffer volume. The invention applies to fast neutron nuclear reactor cooled by sodium [fr

  15. Diode, transistor & fet circuits manual

    CERN Document Server

    Marston, R M

    2013-01-01

    Diode, Transistor and FET Circuits Manual is a handbook of circuits based on discrete semiconductor components such as diodes, transistors, and FETS. The book also includes diagrams and practical circuits. The book describes basic and special diode characteristics, heat wave-rectifier circuits, transformers, filter capacitors, and rectifier ratings. The text also presents practical applications of associated devices, for example, zeners, varicaps, photodiodes, or LEDs, as well as it describes bipolar transistor characteristics. The transistor can be used in three basic amplifier configuration

  16. Water coolant supply in relation to different ultrasonic scaler systems, tips and coolant settings

    NARCIS (Netherlands)

    Koster, T.J.G.; Timmerman, M.F.; Feilzer, A.J.; van der Velden, U.; van der Weijden, F.A.

    2009-01-01

    Objective: This study evaluated "in vitro" the consistency of the water coolant supply for five ultrasonic scaler systems in relation to the tip type and different coolant settings. Material and Methods: The systems were: EMS PM-400, EMS PM-600, Satelec P-max, Dürr Vector and Dentsply Cavitron. For

  17. Calculational advance in the modeling of fuel-coolant interactions

    International Nuclear Information System (INIS)

    Bohl, W.R.

    1982-01-01

    A new technique is applied to numerically simulate a fuel-coolant interaction. The technique is based on the ability to calculate separate space- and time-dependent velocities for each of the participating components. In the limiting case of a vapor explosion, this framework allows calculation of the pre-mixing phase of film boiling and interpenetration of the working fluid by hot liquid, which is required for extrapolating from experiments to a reactor hypothetical accident. Qualitative results are compared favorably to published experimental data where an iron-alumina mixture was poured into water. Differing results are predicted with LMFBR materials

  18. Lead Coolant Test Facility Development Workshop

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

    Directory of Open Access Journals (Sweden)

    Veijo KAUPPINEN

    2012-11-01

    Full Text Available Strained environment is a global problem. In metal industries the use of coolant has become more problematic in terms of both employee health and environmental pollution. It is said that the use of coolant forms approximately 8 - 16 % of the total production costs.The traditional methods that use coolants are now obviously becoming obsolete. Hence, it is clear that using a dry cutting system has great implications for resource preservation and waste reduction. For this purpose, a new cooling system is designed for dry cutting. This paper presents the new eco-friendly cooling innovation and the benefits gained by using this method. The new cooling system relies on a unit for ionising ejected air. In order to compare the performance of using this system, cutting experiments were carried out. A series of tests were performed on a horizontal turning machine and on a horizontal machining centre.

  20. Main coolant pump testing at Ontario Hydro

    International Nuclear Information System (INIS)

    Hartlen, R.

    1991-01-01

    This article describes Ontario Hydro Research Division's experience with a computerized data acquisition and analysis system for monitoring mechanical vibration in reactor coolant pumps. The topics covered include bench-marking of the computer system and the coolant pumps, signatures of normal and malfunctioning pumps, analysis of data collected by the monitoring system, simulation of faults, and concerns that have been expressed about data interpretation, sensor types and locations, alarm/shutdown limits and confirmation of nondestructive examination testing. This presentation consists of overheads only

  1. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

  2. The propagation of pressure pulsations in the primary circuit of power plant A1

    International Nuclear Information System (INIS)

    Pecinka, L.

    1976-01-01

    A classification is made of the exciting forces of pressure pulsations in the primary coolant circuit with forced coolant circulation. A mathematical model is constructed of the propagation of pressure pulsations in the system and examples of measurements are given. The measurement methods used and the methods for the generalization of obtained data are assessed. The methods and results of the measurements of hydrodynamic pressure pulsations in a closed primary circuit with forced coolant circulation of the A-1 nuclear power plant are given. (F.M.)

  3. Composition and concentration of soluble and particulate matter in the coolant of the reactor primary cooling system of the Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Garcia Rodenas, Luis; La Gamma, Ana M.; Villegas, Marina; Fernandez, Alberto N.; Allemandi, Walter; Manera, Raul; Rosales, Hugo

    2000-01-01

    Nuclear power plants type PWR and PHWR (pressurized water reactor and pressurized heavy water reactor) have three coolant circuits which only exchange energy among them. The primary circuit, whose coolant extracts the reactor energy, the secondary circuit or water-steam cycle and the tertiary circuit which could be lake, river or sea water. The chemistry of the primary and secondary coolants is carefully controlled with the aim of minimizing the corrosion of structural materials. However, very low rates of corrosion are inevitable and one of the consequences of the corrosion processes is the presence of soluble and particulate matter in the coolant from where several problems associated with mass transfer arisen. In this way radioactive nuclides are transported out of the core to the steam generators, hydraulic resistance increases and heat transfer capability degrades. In the present paper some alternative techniques are proposed for the quantification of both, the particulate and soluble matter present in the coolant and their correspondent composition. Some results are also included and discussed. (author)

  4. Estimation of maximum pressure in small containments of PWR reactors due to loss of coolant accident in primary circuit; Estimativa da pressao maxima em contencoes de reatores PWR de pequeno porte devido a um acidente de perda de refrigerante no circuito primario

    Energy Technology Data Exchange (ETDEWEB)

    Mendes Neto, Teofilo [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil); Moreira, Joao Manoel Losada [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2000-07-01

    This work studies the problem of containment pressurization after a LOCA in reactors with small containment free volumes. The relationship between the reactor power and the containment free volume is described with the ratio between the volumes of the primary circuit and of the containment. The maximum pressure in a containment, following a LOCA, obtained after a correlation based on large containment PWR, is around 185 psia for a primary circuit and containment volumes ratio of 0.025. For the same problem, calculations with the CONTEMPT-LT code produced a maximum pressure of 162 psia. The behavior of the temperature after a LOCA to the containment, as a function of the ratio between the primary circuit and containment volume, is such that it increases reaching asymptotically to a maximum; differently, the pressure increases almost linearly with the ratio of volumes. (author)

  5. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

    The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

  6. Fission product release into the primary coolant

    International Nuclear Information System (INIS)

    Apperson, C.E.

    1977-01-01

    The analytic evaluation of steady state primary coolant activity is discussed. The reported calculations account for temperature dependent fuel failure in two particle types and arbitrary radioactive decay chains. A matrix operator technique implemented in the SUVIUS code is used to solve the simultaneous equations. Results are compared with General Atomic Company's published results

  7. Deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Parsons, P.D.; Hindle, E.D.; Mann, C.A.

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail.

  8. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  9. The deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Parsons, P.D.; Hindle, E.D.; Mann, C.A.

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail. (author)

  10. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    National Research Council Canada - National Science Library

    Can, Levent

    2006-01-01

    .... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

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

    Science.gov (United States)

    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.

  12. Use of Distribution Devices for Hydraulic Profiling of Coolant Flow in Core Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

    Full Text Available In setting up a reactor plant for the transportation-power module of the megawatt class an important task is to optimize the path of flow, i.e. providing moderate hydraulic resistance, uniform distribution of the coolant. Significant contribution to the hydraulic losses makes one selected design of the coolant supplies. It is, in particular, hemispherical or semi-elliptical shape of the supply reservoir, which is selected to reduce its mass, resulting in the formation of torusshaped vortex in the inlet manifold, that leads to uneven coolant velocity at the inlet into the core, the flow pulsations, hydraulic losses.To control the flow redistribution in the core according to the level of energy are used the switchgear - deflectors installed in a hemispherical reservoir supplying coolant to the fuel elements (FE of the core of gas-cooled reactor. This design solution has an effect on the structure of the flow, rate in the cooling duct, and the flow resistance of the collector.In this paper we present the results of experiments carried out on the gas dynamic model of coolant paths, deflectors, and core, comprising 55 fuel rod simulators. Numerical simulation of flow in two-parameter model, using the k-ε turbulence model, and the software package ANSYS CFX v14.0 is performed. The paper demonstrates that experimental results are in compliance with calculated ones.The results obtained suggest that the use of switchgear ensures a coolant flow balance directly at the core inlet, thereby providing temperature reduction of fuel rods with a uniform power release in the cross-section. Considered options to find constructive solutions for deflectors give an idea to solve the problem of reducing hydraulic losses in the coolant paths, to decrease pulsation components of flow in the core and length of initial section of flow stabilization.

  13. Resonance circuits for adiabatic circuits

    Directory of Open Access Journals (Sweden)

    C. Schlachta

    2003-01-01

    Full Text Available One of the possible techniques to reduces the power consumption in digital CMOS circuits is to slow down the charge transport. This slowdown can be achieved by introducing an inductor in the charging path. Additionally, the inductor can act as an energy storage element, conserving the energy that is normally dissipated during discharging. Together with the parasitic capacitances from the circuit a LCresonant circuit is formed.

  14. Internal twist drill coolant channel modelling using computational fluid dynamics

    OpenAIRE

    Johns, AS; Hewson, RW; Merson, E; Summers, JL; Thompson, HM

    2014-01-01

    Coolant technologies have become increasingly used within twist-drill machining. Internal helical coolant channels are a common means of delivering high-pressure coolant directly to the cutting edge to increase the transfer of heat away from the cutting edge and to aid with chip evacuation. However, the complex behaviour of Coriolis and centrifugal forces generated by large angular velocities and the curvature of the coiled channel govern the highly turbulent flow of coolant. This research us...

  15. Membrane systems and their use in nuclear power plants. Treatment of primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kus, Pavel; Bartova, Sarka; Skala, Martin; Vonkova, Katerina [Research Centre Rez, Husinec-Rez (Czech Republic). Technological Circuits Innovation Dept.; Zach, Vaclav; Kopa, Roman [CEZ a.s., Temelin (Czech Republic). Nuclear Power Plant Temelin

    2016-03-15

    In nuclear power plants, drained primary coolant containing boric acid is currently treated in the system of evaporators and by ion exchangers. Replacement of the system of evaporators by membrane system (MS) will result in lower operating cost mainly due to lower operation temperature. In membrane systems the feed primary coolant is separated into two output streams: retentate and permeate. Retentate stream consists of the concentrated boric acid solution together with other components, while permeate stream consists of purified water. Results are presented achieved by testing a pilot-plant unit of reverse osmosis in nuclear power plant (NPP) Temelin.

  16. The MIT in-pile loops for coolant chemistry and corrosion studies

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Harling, O.K.; Kohse, G.E.; Ballinger, R.G.

    1988-01-01

    A set of modular subsystems and components have been developed which can be combined in several ways to create in-pile loop facilities for experimental investigation of the effect of LWR reactor coolant chemistry on corrosion, both local and global, and on radionuclide transport and deposition. Care has been taken to scale the systems to achieve close similitude for parameters currently thought to control the phenomena of interest. Initial in-pile runs of the first of these facilities, a PWR coolant chemsitry loop, will take place soon. (author)

  17. Management of large scale coolant channel replacement programme for Indian PHWRs

    International Nuclear Information System (INIS)

    Bhatnagar, V.K.; Chadda, S.K.; Arya, R.C.

    1994-01-01

    Coolant channel assemblies form most important core components of pressurised heavy water reactors. Zirconium alloy pressure tube which form part of coolant channel assemblies are subjected to environment of high neutron flux, high pressure and temperature. Under those operating environmental conditions, the pressure tubes material undergoes degradation of metallurgical and mechanical properties in addition to dimensional changes. The coolant channels are subjected to an in-service inspection (ISI) programme for monitoring the health particularly of the pressure tubes. The en-mass replacement of pressure tubes is needed after most of the pressure tubes show unacceptable conditions for an assured safe and reliable operation. An overview of various issues pertaining to this aspect is presented. (author). 4 figs

  18. Fitness for service assessment of coolant channels of Indian PHWRs

    Science.gov (United States)

    Sinha, R. K.; Sinha, S. K.; Madhusoodanan, K.

    2008-12-01

    A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure tube, calandria tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure tube is rolled at both its ends to the end fittings and is located concentrically inside the calandria tube with the help of garter spring spacers. Pressure tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10 MPa and 573 K), and fast neutron flux (typically 3 × 10 17 n/m 2 s, E > 1 MeV neutrons). Under this operating environment, the material of the pressure tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr-2.5 wt% Nb pressure tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure tubes in the earlier generation of Indian pressurised heavy water reactors.

  19. Colloids in PWR primary and secondary coolant. Innovative analytical methods

    International Nuclear Information System (INIS)

    Nowotka, Karsten; Guillodo, Michael; Burchardt, Carsten; Geier, Roland; Lehr, Robert; Stellwag, Bernhard

    2014-01-01

    Transport and deposition of corrosion products in the colloid size range between 1 nm and 1 μm are important for heat transfer performance and corrosion in primary and secondary cooling circuits of LWRs. Direct analysis of the properties of small-sized colloids (< 0.45 μm) is difficult due to the pronounced change of the physicochemical properties of coolant samples in sampling lines. An innovative method, based on a filter cascade and developed in the AREVA Technical Center, named 'Colloid Catcher' (CC, patent pending), permits on-line measurements of the properties of corrosion products in the coolant of LWRs. CC measurements are complementary to classic trace analysis addressing the soluble content. Low and high temperature (up to 330°C) test sections are available, depending on our customer's needs. The CC contains differential pressure detectors at each of the three consecutive membrane filters which allow for an in-situ characterization without modification of the corrosion products chemical nature due to temperature changes and subsequent exposure to the atmosphere. With this method, a 'Colloid Fingerprint' of the test solution can be obtained, ideal for an assessment of the transport and deposition of corrosion products in laboratory and on-site studies. The on-line data can of course be complemented by post filtration membrane characterization by digestion and/or optical methods. The high temperature CC serves at the same time as a sampling point for grab samples, with good reproducibility thanks to continuous liquid flow. The CC has been designed to be deployable on laboratory or industrial cooling circuits. The CC test sections have been qualified using AREVA Technical Center's test loops. First test results obtained with the LT CC are presented. Laboratory data can be used to back up existing results and data of on-site measurement campaigns at BWR and PWR plants which were determined with a basic version of the LT CC

  20. CAREM-25: considerations about primary coolant chemistry

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Iglesias, Alberto M.; Raffo Calderon, Maria C.; Villegas, Marina

    2000-01-01

    World operating experience, in conjunction with basic studies has been modifying chemistry specifications for the primary coolant of water cooled nuclear reactors along with the reactor type and structural materials involved in the design. For the reactor CAREM-25, the following sources of information have been used: 1) Experience gained by the Chemistry Department of the National Atomic Energy Commission (CNEA, Argentina); 2) Participation of the Chemistry Department (CNEA) in international cooperation projects; 3) Guidelines given by EPRI, Siemens-KWU, AECL, etc. Given the main objectives: materials integrity, low radiation levels and personnel safety, which are in turn a balance between the lowest corrosion and activity transport achievable and considering that the CAREM-25 is a pressurized vessel integrated reactor, a group of guidelines for the chemistry and additives for the primary coolant have been given in the present work. (author)

  1. Recovery studies for plutonium machining oil coolant

    International Nuclear Information System (INIS)

    Navratil, J.D.; Baldwin, C.E.

    1977-01-01

    Lathe coolant oil, contaminated with plutonium and having a carbon tetrachloride diluent, is generated in plutonium machining areas at Rocky Flats. A research program was initiated to determine the nature of plutonium in this mixture of oil and carbon tetrachloride. Appropriate methods then could be developed to remove the plutonium and to recycle the oil and carbon tetrachloride. Studies showed that the mixtures of spent oil and carbon tetrachloride contained particulate plutonium and plutonium species that are soluble in water or in oil and carbon tetrachloride. The particulate plutonium was removed by filtration; the nonfilterable plutonium was removed by adsorption on various materials. Laboratory-scale tests indicated the lathe-coolant oil mixture could be separated by distilling the carbon tetrachloride to yield recyclable products

  2. Utilization of symmetrical components in a communication-assisted protection scheme for radial MV feeders with variable or reduced short-circuit currents

    DEFF Research Database (Denmark)

    Ciontea, Catalin-Iosif; Bak, Claus Leth; Blaabjerg, Frede

    2017-01-01

    -circuit power and reduced short-circuit currents compared to the OverCurrent protection. In addition, coordination of the relays that use the proposed algorithm does not require a prior knowledge of the load or fault currents. The setting procedure of the relays that use the new protection algorithm......Distribution networks are evolving toward the vision of microgrids, which are defined as power systems that contain both loads and Distributed Energy Resources and operate in a grid-connected or islanded mode. Correct operation of the OverCurrent protection is compromised in a typical microgrid...... because the fault current is significantly lower in the islanded mode compared to the grid-connected mode and consequently a single set of settings for the OC relays is not sufficient. This paper propose a communication-assisted protection scheme that is able to operate correctly in a radial Medium...

  3. Design of automotive engine coolant hoses

    Directory of Open Access Journals (Sweden)

    Hrishikesh D BACHCHHAV

    2018-03-01

    Full Text Available In this paper, we are present the performance of engine coolant hoses (radiator hoses used in passenger cars by checking various physical behaviours such as hose leakage, hose burst, hose collapse or any mechanical damage as studied-thru design guidelines, CFD analysis and product validation testing and also check pressure drop of the hoses when engine will be running. The design term is more likely used for technical part modelling using CAD tool. Later on, we will focus on the transformation of the part design to process design. The process design term is more likely used for "tooling design" for manufacturing of the product using CAD Tool. Then inlet hose carries coolant from engine to radiator inlet tank, then coolant circulated in radiator and passed through radiator outlet tank to water pump of engine with the help of outlet hose. After that …nding any leakage, Burst, damage or collapse of hose and pressure drop of the hose with the help of design checklist, CFD Analysis and product validation testing.

  4. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    Spinks, N.

    1997-01-01

    An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

  5. Developing a Domain Model for Relay Circuits

    DEFF Research Database (Denmark)

    Haxthausen, Anne Elisabeth

    2009-01-01

    for railways madeby Bjørner et.al., however our model is more general: the components can be of any kind and can later be refined to e.g. railway components or circuit components. Then we show how the abstract network model can be refined into an explicit model for relay circuits. The circuit model describes...... the statics as well as the dynamics of relay circuits, i.e. how a relay circuit can be composed legally from electrical components as well as how the components may change state over time. Finally the circuit model is transformed into an executable model, and we show how a concrete circuit can be defined...

  6. Principal working group 3 on primary circuit integrity

    International Nuclear Information System (INIS)

    1992-01-01

    The main themes of this conference (13 papers) are: operating experience on leakages and failures in nuclear power plant piping, coolant circuits and steam generator tubes, probabilistic estimation and risk assessment, system failure analysis, leakage events and frequency, leak rate models and crack propagation mechanics, damage mechanisms and rupture probability

  7. Enhancement of Linear Circuit Program

    DEFF Research Database (Denmark)

    Gaunholt, Hans; Dabu, Mihaela; Beldiman, Octavian

    1996-01-01

    In this report a preliminary user friendly interface has been added to the LCP2 program making it possible to describe an electronic circuit by actually drawing the circuit on the screen. Component values and other options and parameters can easily be set by the aid of the interface. The interfac...

  8. Radiation-sensitive switching circuits

    Energy Technology Data Exchange (ETDEWEB)

    Moore, J.H.; Cockshott, C.P.

    1976-03-16

    A radiation-sensitive switching circuit has a light emitting diode which supplies light to a photo-transistor, the light being interrupted from time to time. When the photo-transistor is illuminated, current builds up and when this current reaches a predetermined value, a trigger circuit changes state. The peak output of the photo-transistor is measured and the trigger circuit is arranged to change state when the output of the device is a set proportion of the peak output, so as to allow for aging of the components. The circuit is designed to control the ignition system in an automobile engine.

  9. Development of a steady-state calculation model for the KALIMER PDRC(Passive Decay Heat Removal Circuit)

    International Nuclear Information System (INIS)

    Chang, Won Pyo; Ha, Kwi Seok; Jeong, Hae Yong; Kwon, Young Min; Eoh, Jae Hyuk; Lee, Yong Bum

    2003-06-01

    A sodium circuit has usually featured for a Liquid Metal Reactor(LMR) using sodium as coolant to remove the decay heat ultimately under accidental conditions because of its high reliability. Most of the system codes used for a Light Water Reactor(LWR) analysis is capable of calculating natural circulation within such circuit, but the code currently used for the LMR analysis does not feature stand alone capability to simulate the natural circulation flow inside the circuit due to its application limitation. To this end, the present study has been carried out because the natural circulation analysis for such the circuit is realistically raised for the design with a new concept. The steady state modeling is presented in this paper, development of a transient model is also followed to close the study. The incompressibility assumption of sodium which allow the circuit to be modeled with a single flow, makes the model greatly simplified. Models such as a heat exchanger developed in the study can be effectively applied to other system analysis codes which require such component models

  10. Reduction of adhesive stain defect in flexible printed circuit board on hot pressing process: A case study of electronic component factory

    Directory of Open Access Journals (Sweden)

    Sakulkaew Srisang

    2015-03-01

    Full Text Available The objective of this research is a reduction of an adhesive stain defect in flexible printed circuit board in hot pressing process, the electronic factory. The manufacturing have been processing by sheet type of products with ninety-six pieces of flexible printed circuit boards. Causes of the problem include the before and internal hot pressing process. In process beginning times, the most right row of products between the cooling plate and the hot pressing machine has temperature 71.2◦C that is higher than glass transition temperature (Tg 60◦C. Those products’ temperature lead to evaporate a polyimide adhesive before hot pressing process beginning. The internal hot pressing process include the preheat times and the pressure time. In the preheat time the problem is a gap between lower and upper plate, was under specification (Under 1 mm and leaded to adhesive polyimide stain. In the actuality this time requires temperature and low pressure that mean a gap within 1 – 2 mm (between lower and upper plate. In pressure times the hot pressing plate surface is not flat and products are pressed by insufficient force that it lead to generate an adhesive stain on flexible printed circuit boards. That force is measured by the pre-scale paper and a result, RGB color, is provided. And then color density (From standard color sample and RGB color (From pre-scale paper is found out the relation by Photoshop program and multiple regression theory using. The formula is applied to compare with defect so as to find out the suitable color density (Defects reducing. The solving solutions is provided including the gap reduced adjustment between cooling plate and hot pressing machine before hot pressing process, the plate adjustment within specification in the preheat time and the pressing plate polishing in the pressure time. Results of study and solving are provide defect reduction from 24.4 percentage to 7.2 percentage of total study product.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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

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

    International Nuclear Information System (INIS)

    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

  14. The binary cycle for dry-coolant electric power plants

    International Nuclear Information System (INIS)

    Jaumotte, A.L.; Gaivao, A.

    1983-01-01

    Steadily growing water requirements have led the designers of large nuclear power plants to consider the use of a dry coolant seriously in the future (since the problem of a water shortage has already emerged). Two fluids are selected, ammonia and freon 22. In fact, this represents a binary cycle, either with ammonia or with freon (phase change heat transfer fluid). Freon 22 offers the major advantage over ammonia of great chemical stability. It is also nonflammable and nontoxic. Moreover, the use of a phase change fluid offers the possibility of reducing the difference between the air temperature and the steam condensation temperature, since exchange in the condenser is isothermal. The feasibility of the mechanical components, and the successful progress of experiments under way, tend to show that the binary cycle may be a highly attractive answer to the problem of water availability [fr

  15. Reactor water chemistry relevant to coolant-cladding interaction

    International Nuclear Information System (INIS)

    1987-09-01

    The report is a summary of the work performed in a frame of a Coordinated Research Program organized by the IAEA and carried out from 1981 till 1986. It consists of a survey on our knowledge on coolant-cladding interaction: the basic phenomena, the relevant parameters, their control and the modelling techniques implemented for their assessment. Based upon the results of this Coordinated Research Program, the following topics are reviewed on the report: role of water chemistry in reliable operation of nuclear power plants; water chemistry specifications and their control; behaviour of fuel cladding materials; corrosion product behaviour and crud build-up in reactor circuits; modelling of corrosion product behaviour. This report should be of interest to water chemistry supervisors at the power plants, to experts in utility engineering departments, to fuel designers, to R and D institutes active in the field and to the consultants of these organizations. A separate abstract was prepared for each of the 3 papers included in the Annex of this document. Refs, figs, tabs

  16. Fast reactors bulk sodium coolant disposal NOAH process application

    International Nuclear Information System (INIS)

    Magny, E. de; Berte, M.

    1997-01-01

    Within the frame of the fast reactors decommissioning, the becoming of contaminated sodium coolant from primary, secondary and auxiliary circuits is an important aspect. The 'NOAH' sodium disposal process, developed by the French Atomic Energy Commission (CEA), is presented as the only process, for destroying large quantities of contaminated sodium, that has attained industrial status. The principles and technical options of the process are described and main advantages such as safety , operating simplicity and compactness of the plant are put forward. The process has been industrially validated in 1993/1994 by successfully reacting the 37 metric tons of primary contaminated sodium from the French Rapsodie experimental reactor. The main outstanding aspects and experience gained from this so called 'DESORA' operation (DEstruction of SOdium from RApsodie) are recalled. Another industrial application concerns the current project for destroying more than 1500 metric tons of contaminated sodium from the British PFR (Prototype Fast Reactor) in Scotland. Although the design is in the continuity of DESORA, it has taken into account the specific requirements of PFR application and the experience feed back from Rapsodie. The main technical options and performances of the PFR sodium reaction unit are presented while mentioning the design evolution. (author)

  17. Primer printed circuit boards

    CERN Document Server

    Argyle, Andrew

    2009-01-01

    Step-by-step instructions for making your own PCBs at home. Making your own printed circuit board (PCB) might seem a daunting task, but once you master the steps, it's easy to attain professional-looking results. Printed circuit boards, which connect chips and other components, are what make almost all modern electronic devices possible. PCBs are made from sheets of fiberglass clad with copper, usually in multiplelayers. Cut a computer motherboard in two, for instance, and you'll often see five or more differently patterned layers. Making boards at home is relatively easy

  18. ANALYSIS OF THE IMPACT PROPERTIES OF THE COOLANT RECOVERY SYSTEM HEAT LOSSES OF COMBINED COMPRESSOR-POWER PLANT ON ITS CHARACTERISTICS

    OpenAIRE

    Yusha V.L.; Chernov G.; Raykovsky N.

    2012-01-01

    The paper presents results of theoretical analysis of the effectiveness of an ideal thermodynamic cycle internal combustion engine combined with an external utilization of exhaust heat. The influence of the properties of the coolant circuit of utilization on its operational parameters and characteristics of the power plant.

  19. ANALYSIS OF THE IMPACT PROPERTIES OF THE COOLANT RECOVERY SYSTEM HEAT LOSSES OF COMBINED COMPRESSOR-POWER PLANT ON ITS CHARACTERISTICS

    Directory of Open Access Journals (Sweden)

    Yusha V.L.

    2012-12-01

    Full Text Available The paper presents results of theoretical analysis of the effectiveness of an ideal thermodynamic cycle internal combustion engine combined with an external utilization of exhaust heat. The influence of the properties of the coolant circuit of utilization on its operational parameters and characteristics of the power plant.

  20. Nuclear power plant component protection

    International Nuclear Information System (INIS)

    Michel, E.; Ruf, R.; Dorner, H.

    1976-01-01

    Described is a nuclear power plant installation which includes a concrete biological shield forming a pit in which a reactor pressure vessel is positioned. A steam generator on the outside of the shield is connected with the pressure vessel via coolant pipe lines which extend through the shield, the coolant circulation being provided by a coolant pump which is also on the outside of the shield. To protect these components on the outside of the shield and which are of mainly or substantially cylindrical shape, semicylindrical concrete segments are interfitted around them to form complete outer cylinders which are retained against outward separation radially from the components, by rings of high tensile steel which may be interspaced so closely that they provide, in effect, an outer steel cylinder. The invention is particularly applicable to pressurized-water coolant reactor installations

  1. Coolant Void Reactivity Analysis of CANDU Lattice

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    Models of CANDU-6 and ACR-700 fuel lattices were constructed for a single bundle and 2 by 2 checkerboard to understand the physics related to CVR. Also, a familiar four factor formula was used to predict the specific contributions to reactivity change in order to achieve an understanding of the physics issues related to the CVR. At the same time, because the situation of coolant voiding should bring about a change of neutron behavior, the spectral changes and neutron current were also analyzed. The models of the CANDU- 6 and ACR-700 fuel lattices were constructed using the Monte Carlo code MCNP6 using the ENDF/B-VII.0 continuous energy cross section library based on the specification from AECL. The CANDU fuel lattice was searched through sensitivity studies of each design parameter such as fuel enrichment, fuel pitch, and types of burnable absorber for obtaining better behavior in terms of CVR. Unlike the single channel coolant voiding, the ACR-700 bundle has a positive reactivity change upon 2x2 checkerboard coolant voiding. Because of the new path for neutron moderation, the neutrons from the voided channel move to the no-void channel where they lose energy and come back to the voided channel as thermal neutrons. This phenomenon causes the positive CVR when checkerboard voiding occurs. The sensitivity study revealed the effects of the moderator to fuel volume ratio, fuel enrichment, and burnable absorber on the CVR. A fuel bundle with low moderator to fuel volume ratio and high fuel enrichment can help achieve negative CVR.

  2. The Maplin electronic circuits handbook

    CERN Document Server

    Tooley, Michael

    1990-01-01

    The Maplin Electronic Circuits Handbook provides pertinent data, formula, explanation, practical guidance, theory and practical guidance in the design, testing, and construction of electronic circuits. This book discusses the developments in electronics technology techniques.Organized into 11 chapters, this book begins with an overview of the common types of passive component. This text then provides the reader with sufficient information to make a correct selection of passive components for use in the circuits. Other chapters consider the various types of the most commonly used semiconductor

  3. Multirods burst tests under loss-of-coolant conditions

    International Nuclear Information System (INIS)

    Kawasaki, S.; Uetsuka, H.; Furuta, T.

    1983-01-01

    In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

  4. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  5. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  6. Reduction of adhesive stain defect in flexible printed circuit board on hot pressing process: a case study of electronic component factory

    Directory of Open Access Journals (Sweden)

    Sakulkaew Srisang

    2014-09-01

    Full Text Available The objective of this research is a reduction of an adhesive stain defect in flexible printed circuit board in hot pressing process, the electronic factory. The manufacturing have been processing by sheet type of products with ninety-six pieces of flexible printed circuit boards. Causes of the problem include the before and internal hot pressing process. In process beginning times, the most right row of products between the cooling plate and the hot pressing machine has temperature 71.2◦C that is higher than glass transition temperature (Tg 60◦C. Those products’ temperature lead to evaporate a polyimide adhesive before hot pressing process beginning. The internal hot pressing process include the preheat times and the pressure time. In the preheat time the problem is a gap between lower and upper plate, was under specification(Under 1 mm and leaded to adhesive polyimide stain. In the actuality this time requires temperature and low pressure that mean a gap within 1 – 2 mm (between lower and upper plate. In pressure times the hot pressing plate surface is not flat and products are pressed by insufficient force that it lead to generate an adhesive stain on flexible printed circuit boards. That force is measured by the pre-scale paper and a result, RGB color, is provided. And then color density (From standard color sample and RGB color (From pre-scale paper is found out the relation by Photoshop program and multiple regression theory using. The formula is applied to compare with defect so as to find out the suitable color density (Defects reducing. The solving solutions is provided including the gap reduced adjustment between cooling plate and hot pressing machine before hot pressing process, the plate adjustment within specification in the preheat time and the pressing plate polishing in the pressure time. Results of study and solving are provide defect reduction from 24.4 percentage to 7.2 percentage of total study product.

  7. An Equivalent Circuit for Landau Damping

    DEFF Research Database (Denmark)

    Pécseli, Hans

    1976-01-01

    An equivalent circuit simulating the effect of Landau damping in a stable plasma‐loaded parallel‐plate capacitor is presented. The circuit contains a double infinity of LC components. The transition from stable to unstable plasmas is simulated by the introduction of active elements into the circuit....

  8. Experimental research and development of main circulation pump bearings in reactor plants using heavy liquid-metal coolants

    International Nuclear Information System (INIS)

    Zudin, A.; Beznosov, A.; Chernysh, A.; Prikazchikov, G.

    2015-01-01

    At the present time, specialists in Russia are engaged in designing the BREST-OD-300 fast neutron lead-coolant reactor plant. There is currently no experience in designing and operating axial pumps of lead-coolant reactor plants, including one of their major units – bearing unit. Selection and substantiation of operating and structural parameters of plain friction bearings used in main circulation pumps of reactor plants running on heavy liquid-metal coolants are important tasks that are solved at the NNSTU. Development of a feasible procedure for designing bearings and its components operating within the structure of the main circulation pump of a reactor plant running on a heavy liquid-metal coolant as well as guidelines for an optimized structural scheme of such bearings set a goal of performing a range of theoretically-calculated and experimental works. The report contains testing data of a hydrostatic bearing with reciprocal fricative choking tested on the NNSTU FT-4 bench running on a lead coolant within the range of 420-500degC. There have been presented a scheme of a bench for testing a contact friction bearing on a high-temperature coolant and the results of investigation tests of bearings of such type at T = 450 ÷ 500degC. Material of the bearing sleeve is steel 08X18H10T, and a possibility is provided with regard to installation of the bearing sleeves and shaft made of non-metal materials (ceramic materials, silicified graphite, etc.). The presented testing data of plain friction bearings operating in a high-temperature heavy liquid-metal coolant will serve as a ground for making an alternative choice of a plain friction bearing for the main circulation pump of a reactor plant running on a heavy liquid-metal coolant. (author)

  9. Electronic components

    CERN Document Server

    Colwell, Morris A

    1976-01-01

    Electronic Components provides a basic grounding in the practical aspects of using and selecting electronics components. The book describes the basic requirements needed to start practical work on electronic equipment, resistors and potentiometers, capacitance, and inductors and transformers. The text discusses semiconductor devices such as diodes, thyristors and triacs, transistors and heat sinks, logic and linear integrated circuits (I.C.s) and electromechanical devices. Common abbreviations applied to components are provided. Constructors and electronics engineers will find the book useful

  10. Computer simulation of the pneumatic separator in the pneumatic-electrostatic separation system for recycling waste printed circuit boards with electronic components.

    Science.gov (United States)

    Xue, Mianqiang; Xu, Zhenming

    2013-05-07

    Technologies could be integrated in different ways into automatic recycling lines for a certain kind of electronic waste according to practical requirements. In this study, a new kind of pneumatic separator with openings at the dust hooper was applied combing with electrostatic separation for recycling waste printed circuit boards. However, the flow pattern and the particles' movement behavior could not be obtained by experimental methods. To better control the separation quantity and the material size distribution, computational fluid dynamics was used to model the new pneumatic separator giving a detailed understanding of the mechanisms. Simulated results showed that the tangential velocity direction reversed with a relatively small value. Axial velocity exhibited two sharp decreases at the x axis. It is indicated that the bottom openings at the dust hopper resulted in an enormous change in the velocity profile. A new phenomenon that was named dusting was observed, which would mitigate the effect of particles with small diameter on the following electrostatic separation and avoid materials plugging caused by the waste printed circuit boards special properties effectively. The trapped materials were divided into seven grades. Experimental results showed that the mass fraction of grade 5, grade 6, and grade 7 materials were 27.54%, 15.23%, and 17.38%, respectively. Grade 1 particles' mass fraction was reduced by 80.30% compared with a traditional separator. Furthermore, the monocrystalline silicon content in silicon element in particles with a diameter of -0.091 mm was 18.9%, higher than that in the mixed materials. This study could serve as guidance for the future material flow control, automation control, waste recycling, and semiconductor storage medium destruction.

  11. The major components of particles emitted during recycling of waste printed circuit boards in a typical e-waste workshop of South China

    Science.gov (United States)

    Bi, Xinhui; Simoneit, Bernd R. T.; Wang, ZhenZhen; Wang, Xinming; Sheng, Guoying; Fu, Jiamo

    2010-11-01

    Electronic waste from across the world is dismantled and disposed of in China. The low-tech recycling methods have caused severe air pollution. Air particle samples from a typical workshop of South China engaged in recycling waste printed circuit boards have been analyzed with respect to chemical constituents. This is the first report on the chemical composition of particulate matter (PM) emitted in an e-waste recycling workshop of South China. The results show that the composition of PM from this recycling process was totally different from other emission sources. Organic matter comprised 46.7-51.6% of the PM. The major organic constituents were organophosphates consisting mainly of triphenyl phosphate (TPP) and its methyl substituted compounds, methyl esters of hexadecanoic and octadecanoic acids, levoglucosan and bisphenol A. TPP and bisphenol A were present at 1-5 orders of magnitude higher than in other indoor and outdoor environments throughout the world, which implies that they might be used as potential markers for e-waste recycling. The elemental carbon, inorganic elements and ions had a minor contribution to the PM (<5% each). The inorganic elements were dominated by phosphorus and followed by crustal elements and metal elements Pb, Zn, Sn, and lesser Cu, Sb, Mn, Ni, Ba and Cd. The recycling of printed circuit boards was demonstrated as an important contributor of heavy metal contamination, particularly Cd, Pb and Ni, to the local environment. These findings suggest that this recycling method represents a strong source of PM associated with pollutants to the ambient atmosphere of an e-waste recycling locale.

  12. Modeling the transport of nitrogen in an NPP-2006 reactor circuit

    Science.gov (United States)

    Stepanov, O. E.; Galkin, I. Yu.; Sledkov, R. M.; Melekh, S. S.; Strebnev, N. A.

    2016-07-01

    Efficient radiation protection of the public and personnel requires detecting an accident-initiating event quickly. Specifically, if a heat-exchange tube in a steam generator is ruptured, the 16N radioactive nitrogen isotope, which contributes to a sharp increase in the steam activity before the turbine, may serve as the signaling component. This isotope is produced in the core coolant and is transported along the circulation circuit. The aim of the present study was to model the transport of 16N in the primary and the secondary circuits of a VVER-1000 reactor facility (RF) under nominal operation conditions. KORSAR/GP and RELAP5/Mod.3.2 codes were used to perform the calculations. Computational models incorporating the major components of the primary and the secondary circuits of an NPP-2006 RF were constructed. These computational models were subjected to cross-verification, and the calculation results were compared to the experimental data on the distribution of the void fraction over the steam generator height. The models were proven to be valid. It was found that the time of nitrogen transport from the core to the heat-exchange tube leak was no longer than 1 s under RF operation at a power level of 100% N nom with all primary circuit pumps activated. The time of nitrogen transport from the leak to the γ-radiation detection unit under the same operating conditions was no longer than 9 s, and the nitrogen concentration in steam was no less than 1.4% (by mass) of its concentration at the reactor outlet. These values were obtained using conservative approaches to estimating the leak flow and the transport time, but the radioactive decay of nitrogen was not taken into account. Further research concerned with the calculation of thermohydraulic processes should be focused on modeling the transport of nitrogen under RF operation with some primary circuit pumps deactivated.

  13. Circuit Training.

    Science.gov (United States)

    Nelson, Jane B.

    1998-01-01

    Describes a research-based activity for high school physics students in which they build an LC circuit and find its resonant frequency of oscillation using an oscilloscope. Includes a diagram of the apparatus and an explanation of the procedures. (DDR)

  14. Affective Circuits

    DEFF Research Database (Denmark)

    to the intersecting streams of goods, people, ideas, and money as they circulate between African migrants and their kin who remain back home. They also show the complex ways that emotions become entangled in these exchanges. Examining how these circuits operate in domains of social life ranging from child fosterage...

  15. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  16. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  17. Coolant make-up device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

  18. Upper internals of PWR with coolant flow separator

    International Nuclear Information System (INIS)

    Chevereau, G.; Heuze, A.

    1989-01-01

    The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

  19. Implementation of Chua's circuit using simulated inductance

    Science.gov (United States)

    Gopakumar, K.; Premlet, B.; Gopchandran, K. G.

    2011-05-01

    In this study we describe how to build an inductorless version of the classic Chua's circuit. A suitable inductor for Chua's circuit is often hard to procure. The required inductor for the circuit is designed using simple circuit elements such as resistors, capacitors and operational amplifiers. The complete circuit can be implemented by using off-the-shelf components, and it can readily be integrated on a single chip. This design of Chua's circuit allows the original dynamics to be slowed down to just a few hertz, enabling implementation of sophisticated control schemes without severe time restrictions. Another novel feature of the circuit is that losses associated with capacitors due to leakages can easily be compensated by providing negative resistance using the same setup. The chaotic behaviour of the circuit is verified by PSpice and Multisim simulation and also by experimental study on a circuit breadboard. The results give excellent agreement with each other and with the results of previous investigators.

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

    Science.gov (United States)

    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.

  1. Detection of primary coolant leaks in NPP

    International Nuclear Information System (INIS)

    Slavov, S.; Bakalov, I.; Vassilev, H.

    2001-01-01

    The thermo-hydraulic analyses of the SG box behaviour of Kozloduy NPP units 3 and 4 in case of small primary circuit leaks and during normal operation of the existing ventilation systems in order to determine the detectable leakages from the primary circuit by analysing different parameters used for the purposes of 'Leak before break' concept, performed by ENPRO Consult Ltd. are presented. The following methods for leak detection: measurement of relative air humidity in SG box (can be used for detection of leaks with flow rate 3.78 l/min within one hour at ambient parameters - temperature 40 0 - 60 0 C and relative humidity form 30% to 60%); measurement of water level in SG box sumps (can not be used for reliable detection of small primary circuit leakages with flow rate about 3.78 l/min); measurement of gaseous radioactivity in SG box( can be used as a general global indication for detection of small leakages from the primary circuit); measurement of condensate flow after the air coolers of P-1 venting system (can be used for primary circuit leak detection) are considered. For determination of the confinement behaviour, a model used with computer code MELCOR has been developed by ENPRO Consult Ltd. A brief summary based on the capabilities of the different methods of leak detection, from the point of view of the applicability of a particular method is given. For both Units 3 and 4 of Kozloduy NPP a qualified complex system for small leak detection is planned to be constructed. Such a system has to unite the following systems: acoustic system for leak detection 'ALUS'; system for control of the tightness of the main primary circuit pipelines by monitoring the local humidity; system for primary circuit leakage detection by measuring condensate run-off in collecting tank after ventilation system P-1 air coolers

  2. CLAPTRAP-II, a code for calculating pressure transients in the interconnected compartments of a total containment enclosing a breached water reactor circuit

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1977-03-01

    This report describes a code, CLAPTRAP II, which allows the pressure distribution in a multicompartmented total containment to be analysed when the primary circuit suffers a loss-of-coolant accident. (author)

  3. Electron beam welding and laser welding of FRAGEMA fuel assembly components

    International Nuclear Information System (INIS)

    Couturier, J.M.; Etellin, B.; Duthoo, B.; Wache, C.; Taillandier, T.

    1988-01-01

    Neutron balance and activity of the primary coolant circuit are improved in PWR if inconel 718 is replaced by a zirconium alloy for fuel element grids. This paper examines laser welding and EB welding of these zirconium alloy grids [fr

  4. TIMING CIRCUIT

    Science.gov (United States)

    Heyd, J.W.

    1959-07-14

    An electronic circuit is described for precisely controlling the power delivered to a load from an a-c source, and is particularly useful as a welder timer. The power is delivered in uniform pulses, produced by a thyratron, the number of pulses being controlled by a one-shot multivibrator. The starting pulse is synchronized with the a-c line frequency so that each multivlbrator cycle begins at about the same point in the a-c cycle.

  5. Statistical techniques for the identification of reactor component structural vibrations

    International Nuclear Information System (INIS)

    Kemeny, L.G.

    1975-01-01

    The identification, on-line and in near real-time, of the vibration frequencies, modes and amplitudes of selected key reactor structural components and the visual monitoring of these phenomena by nuclear power plant operating staff will serve to further the safety and control philosophy of nuclear systems and lead to design optimisation. The School of Nuclear Engineering has developed a data acquisition system for vibration detection and identification. The system is interfaced with the HIFAR research reactor of the Australian Atomic Energy Commission. The reactor serves to simulate noise and vibrational phenomena which might be pertinent in power reactor situations. The data acquisition system consists of a small computer interfaced with a digital correlator and a Fourier transform unit. An incremental tape recorder is utilised as a backing store and as a means of communication with other computers. A small analogue computer and an analogue statistical analyzer can be used in the pre and post computational analysis of signals which are received from neutron and gamma detectors, thermocouples, accelerometers, hydrophones and strain gauges. Investigations carried out to date include a study of the role of local and global pressure fields due to turbulence in coolant flow and pump impeller induced perturbations on (a) control absorbers, (B) fuel element and (c) coolant external circuit and core tank structure component vibrations. (Auth.)

  6. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  7. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  8. Slow coolant phaseout could worsen warming

    Science.gov (United States)

    Reese, April

    2018-03-01

    In the summer of 2016, temperatures in Phalodi, an old caravan town on a dry plain in northwestern India, reached a blistering 51°C—a record high during a heat wave that claimed more than 1600 lives across the country. Wider access to air conditioning (AC) could have prevented many deaths—but only 8% of India's 249 million households have AC. As the nation's economy booms, that figure could rise to 50% by 2050. And that presents a dilemma: As India expands access to a life-saving technology, it must comply with international mandates—the most recent imposed just last fall—to eliminate coolants that harm stratospheric ozone or warm the atmosphere.

  9. Power module assemblies with staggered coolant channels

    Science.gov (United States)

    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.

  10. Coolant sampling device for a reactor

    International Nuclear Information System (INIS)

    Yanagisawa, Ko.

    1982-01-01

    Purpose: To enable accurate measurement of fine quantity of metal impurity in primary coolant by removing unneccessary impurity adhered to a collecting tube when the metal impurity is detected. Constitution: A sampling device is mounted at a recirculating system pipe in a BWR type reactor, a collecting tube is connected to the pipe, a sampled water collecting nozzle is also mounted, and a sluice valve for flowing and shutting off the sampled water is disposed in the vicinity of the nozzle. Acid cleaning liquid which is bubbled by inert gas from a tube provided at the downstream side of the valve cleans the inner surface of the collecting tube over the entire length, thereby effectively removing impurity adhered onto the inner surface of the tube. (Kamimura, M.)

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

    Science.gov (United States)

    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.

  12. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  13. Method of charging instruments into liquid metal coolant

    International Nuclear Information System (INIS)

    Yamazaki, Hiroshi

    1980-01-01

    Purpose: To alleviate the thermal shock of a reactor charging machine when charging the machine into liquid metal coolant after the machine is preheated in cover gas. Method: When a reactor fueling machine reaches at the lowermost portion the position immediately above liquid metal coolant surface level, the machine is stopped moving down. The reactor fueling machine is heated at the lowermost portion by thermal radiation from the surface of the liquid metal coolant. After the machine is thus preheated in cover gas, it is again steadily moved down by a winch and charged into the liquid metal coolant. Therefore, the thermal shock of the machine becomes low when charging the machine into the liquid metal coolant to eliminate the damage and deformation at the machine. (Yoshihara, H.)

  14. Determination of iron content in the nitrogen tetroxide coolant

    International Nuclear Information System (INIS)

    Kovaleva, E.P.; Bordak, L.V.

    1979-01-01

    When utilizing dissociating nitrogen tetroxide as a coolant and working fluid of NPP, it is necessary to work out the methods of control of the coolant quality characteristic. The important characteristic of the coolant quality on impurities is the content of corrosion products in it and iron in particular, as the basic macrocomponent of structural materials. For determination of iron content in the nitrogen tetroxide-based coolant, sulphosalicylic method in ammoniacal variant is proposed. A reflected the experimental working outs of the o-phenanthroline method and the method with sulphosalicylic acid. The method of the sample preparation of the N 2 O 4 -based coolant for the subsequent calorimetric analysis is cited

  15. Numerical modeling of the waves evolution generated by the depressurization of the vessels containing a supercritical parameters coolant

    Science.gov (United States)

    Alekseev, Maksim V.; Vozhakov, Ivan S.; Lezhnin, Sergey I.; Pribaturin, Nikolay A.

    2017-10-01

    The development of power plants focuses on increasing the parameters of water coolants up to a supercritical level. Depressurization of the unit circuits with such a coolant leads to emergency situations. Their scenarios can change significantly with the variation of initial pressure and temperature before the start of depressurization. When the pressure drops from the supercritical single-phase region of the initial thermodynamic parameters of the coolant, either the liquid boils up, or the vapor is condensed. Because of the rapid pressure decrease, the phase transition can be non-equilibrium that must be taken into account in the simulation. In the present study, an axisymmetric problem of the outflow of a water coolant from the pipe butt-end is considered. The equations of continuity, momentum and energy for a two-phase homogeneous mixture are solved numerically. The vapor and liquid properties are calculated using the TTSE software package (The Tabular Taylor Series Expansion Method). On the basis of the computer complex LCPFCT (The Flux-Corrected Transport Algorithm) the program code was developed for solving numerous problems on the depressurization of vessels or pipelines, containing superheated water or gas under high pressure. Different variants of outflow in the external model atmosphere and generation of waves are analyzed. The calculated data on the interaction of pressure waves with a barrier are calculated. To describe phase transitions, an asymptotic relaxation model of nonequilibrium evaporation and condensation has been created and tested.

  16. Molten Fuel-Coolant Interactions induced by coolant injection into molten fuel

    International Nuclear Information System (INIS)

    Park, H.S.; Yamano, Norihiko; Maruyama, Yu; Moriyama, Kiyofumi; Yang, Y.; Sugimoto, Jun

    1999-01-01

    To investigate Molten Fuel-Coolant Interactions (MFCIs) in various contact geometries, an experimental program, called MUSE (MUlti-configurations in Steam Explosions), has been initiated under the ALPHA program at JAERI in Japan. The first series of MUSE test has been focused on the coolant injection (CI) and stratified modes of FCIs using water as coolant and molten thermite as molten fuel. The effects of water jet subcooling, jet dynamics, jet shape and system constraint on FCIs energetic in these modes were experimentally investigated by precisely measuring their mechanical energy release in the MUSE facility. It was observed that measured mechanical energy increased with increasing of jet subcooling in a weakly constraint system but decreased in a strongly constraint system. FCI energetic also increased with increasing of water jet velocity. These results suggested that the penetration and dispersion phenomena of a water jet inside a melt determined the mixing conditions of FCIs in these contact modes and consequently played important roles on FCI energetics. To understand fundamental physics of these phenomena and possible mixing conditions in the MUSE tests, a set of visualization tests with several pairs of jet-pool liquids in non-boiling and isothermal conditions were carried out. Numerical simulations of a water jet penetrating into a water pool at non-boiling conditions showed similar behaviors to those observed in the visualization tests. (author)

  17. Electronic logic circuits

    CERN Document Server

    Gibson, J

    2013-01-01

    Most branches of organizing utilize digital electronic systems. This book introduces the design of such systems using basic logic elements as the components. The material is presented in a straightforward manner suitable for students of electronic engineering and computer science. The book is also of use to engineers in related disciplines who require a clear introduction to logic circuits. This third edition has been revised to encompass the most recent advances in technology as well as the latest trends in components and notation. It includes a wide coverage of application specific integrate

  18. Photonic Integrated Circuits

    Science.gov (United States)

    Krainak, Michael; Merritt, Scott

    2016-01-01

    Integrated photonics generally is the integration of multiple lithographically defined photonic and electronic components and devices (e.g. lasers, detectors, waveguides passive structures, modulators, electronic control and optical interconnects) on a single platform with nanometer-scale feature sizes. The development of photonic integrated circuits permits size, weight, power and cost reductions for spacecraft microprocessors, optical communication, processor buses, advanced data processing, and integrated optic science instrument optical systems, subsystems and components. This is particularly critical for small spacecraft platforms. We will give an overview of some NASA applications for integrated photonics.

  19. Water chemistry of the secondary circuit at a nuclear power station with a VVER power reactor

    Science.gov (United States)

    Tyapkov, V. F.; Erpyleva, S. F.

    2017-05-01

    Results of implementation of the secondary circuit organic amine water chemistry at Russian nuclear power plant (NPP) with VVER-1000 reactors are presented. The requirements for improving the reliability, safety, and efficiency of NPPs and for prolonging the service life of main equipment items necessitate the implementation of new technologies, such as new water chemistries. Data are analyzed on the chemical control of power unit coolant for quality after the changeover to operation with the feed of higher amines, such as morpholine and ethanolamine. Power units having equipment containing copper alloy components were converted from the all-volatile water chemistry to the ethanolamine or morpholine water chemistry with no increase in pH of the steam generator feedwater. This enables the iron content in the steam generator feedwater to be decreased from 6-12 to 2.0-2.5 μg/dm3. It is demonstrated that pH of high-temperature water is among the basic factors controlling erosion and corrosion wear of the piping and the ingress of corrosion products into NPP steam generators. For NPP power units having equipment whose construction material does not include copper alloys, the water chemistries with elevated pH of the secondary coolant are adopted. Stable dosing of correction chemicals at these power units maintains pH25 of 9.5 to 9.7 in the steam generator feedwater with a maximum iron content of 2 μg/dm3 in the steam generator feedwater.

  20. Radiation-sensitive switching circuits

    Energy Technology Data Exchange (ETDEWEB)

    Moore, J.H.; Cockshott, C.P.

    1976-03-16

    A radiation-sensitive switching circuit includes a light emitting diode which from time to time illuminates a photo-transistor, the photo-transistor serving when its output reaches a predetermined value to operate a trigger circuit. In order to allow for aging of the components, the current flow through the diode is increased when the output from the transistor falls below a known level. Conveniently, this is achieved by having a transistor in parallel with the diode, and turning the transistor off when the output from the phototransistor becomes too low. The circuit is designed to control the ignition system in an automobile engine.

  1. Quasi-Linear Circuit

    Science.gov (United States)

    Bradley, William; Bird, Ross; Eldred, Dennis; Zook, Jon; Knowles, Gareth

    2013-01-01

    This work involved developing spacequalifiable switch mode DC/DC power supplies that improve performance with fewer components, and result in elimination of digital components and reduction in magnetics. This design is for missions where systems may be operating under extreme conditions, especially at elevated temperature levels from 200 to 300 degC. Prior art for radiation-tolerant DC/DC converters has been accomplished utilizing classical magnetic-based switch mode converter topologies; however, this requires specific shielding and component de-rating to meet the high-reliability specifications. It requires complex measurement and feedback components, and will not enable automatic re-optimization for larger changes in voltage supply or electrical loading condition. The innovation is a switch mode DC/DC power supply that eliminates the need for processors and most magnetics. It can provide a well-regulated voltage supply with a gain of 1:100 step-up to 8:1 step down, tolerating an up to 30% fluctuation of the voltage supply parameters. The circuit incorporates a ceramic core transformer in a manner that enables it to provide a well-regulated voltage output without use of any processor components or magnetic transformers. The circuit adjusts its internal parameters to re-optimize its performance for changes in supply voltage, environmental conditions, or electrical loading at the output

  2. Thermohydraulic status and component behavior in the PWR during the selected meltdown scenario station blackout (SBO); Thermohydraulisches Verhalten und Komponentenverhalten eines DWR bei ausgewaehltem Kernschmelzszenarium infolge Station Blackout (SBO). Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Band, Sebastian; Blaesius, Christoph; Scheuerer, Martina; Steinroetter, Thomas

    2017-09-15

    The report on the thermohydraulic status and component behavior in the PWR during the selected meltdown scenario station blackout (SBO) includes the following issues: status of science and technology on this topic, analysis of a high-pressure meltdown scenario using ATHLET-CD for a German PWR starting from the initiating event station blackout, three-dimensional computational fluid dynamic (CFD) analyses of the pressurizer coolant loop in a generic German PWR, evaluation of the thermohydraulic steam generator behavior and its effect on the involved primary circuit components.

  3. Deeply etched MMI-based components on 4 μm thick SOI for SOA-based optical RAM cell circuits

    Science.gov (United States)

    Cherchi, Matteo; Ylinen, Sami; Harjanne, Mikko; Kapulainen, Markku; Aalto, Timo; Kanellos, George T.; Fitsios, Dimitrios; Pleros, Nikos

    2013-02-01

    We present novel deeply etched functional components, fabricated by multi-step patterning in the frame of our 4 μm thick Silicon on Insulator (SOI) platform based on singlemode rib-waveguides and on the previously developed rib-tostrip converter. These novel components include Multi-Mode Interference (MMI) splitters with any desired splitting ratio, wavelength sensitive 50/50 splitters with pre-filtering capability, multi-stage Mach-Zehnder Interferometer (MZI) filters for suppression of Amplified Spontaneous Emission (ASE), and MMI resonator filters. These novel building blocks enable functionalities otherwise not achievable on our SOI platform, and make it possible to integrate optical RAM cell layouts, by resorting to our technology for hybrid integration of Semiconductor Optical Amplifiers (SOAs). Typical SOA-based RAM cell layouts require generic splitting ratios, which are not readily achievable by a single MMI splitter. We present here a novel solution to this problem, which is very compact and versatile and suits perfectly our technology. Another useful functional element when using SOAs is the pass-band filter to suppress ASE. We pursued two complimentary approaches: a suitable interleaved cascaded MZI filter, based on a novel suitably designed MMI coupler with pre-filtering capabilities, and a completely novel MMI resonator concept, to achieve larger free spectral ranges and narrower pass-band response. Simulation and design principles are presented and compared to preliminary experimental functional results, together with scaling rules and predictions of achievable RAM cell densities. When combined with our newly developed ultra-small light-turning concept, these new components are expected to pave the way for high integration density of RAM cells.

  4. Quality assurance requirements for NSSS-equipment of a BWR-6, and quality levels in the stores of a manufacturer of primary-circuit nuclear components

    International Nuclear Information System (INIS)

    Gomez, E.; Alamo, P.; Perez, A.; Cue, G.

    1982-01-01

    The first part of the paper describes the system used for determining the various levels of quality assurance for nuclear steam supply system (NSSS) equipment of the BWR-6 reactors which General Electric is now building in Spain. It also shows the procedure followed by the General Electric Technical Services Company (GETSCO) for the purchase and fabrication of equipment in Spain, paying due attention to the existing quality assurance requirements. The second part of the paper describes the instrumentation of the various quality assurance levels in the storage system of Equipos Nucleares, S.A. (ENSA), the Spanish manufacturer of the primary components of the NSSS. (author)

  5. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

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

    Science.gov (United States)

    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.

  7. Triboengineering problems of lead coolant in innovative fast reactors

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

    2013-01-01

    Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

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

    International Nuclear Information System (INIS)

    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

  9. Development of core design and analysis technology for integral reactor; development of coolant activity and dose evaluation program

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Sun; Kim, Byeong Soo; Go, Hyun Seok; Lee, Young Wook; Jang, Mee [Seoul National University, Seoul (Korea)

    2002-03-01

    SMART, small- medium-sized integral reactor, is different from the customary electricity-generation PWR in design concepts and structures. The conventional coolant activity evaluation codes used in customary PWRs cannot be applied to SMART. In this study, SAEP(Specific Activity Evaluation Program) is developed that can be applied to both customary PWR and advanced reactor such as SMART. SAEP uses three methods(SAEP Ver.02, Ver.05, Ver.06) to evaluate coolant activity. They solve inhomogeneous linearly-coupled differential equations generated by considering nuclear system as N sub-components. Coolant activities of customary PWR are evaluated by use of SAEP. The results show good agreement with FSAR data. SAEP is used to evaluate coolant activities for SMART and the results are proposed in this study. These results show that SAEP is able to perform coolant activity evaluation for both customary PWR and advanced reactor such as SMART. In addition, with respect to radiation shielding optimization, conventional optimization methods and their characteristics related to radiation shielding are reviewed and analyzed. Strategies for proper usage of conventional methods are proposed to agree with the shielding design cases. 30 refs., 25 figs., 6 tabs. (Author)

  10. Liquid metal cooled nuclear power plant with direct heat transfer from the primary coolant to the working medium

    International Nuclear Information System (INIS)

    Hahn, G.

    1974-01-01

    The cooling systems of the sodium-cooled reactor are entirely inside a containment. The heat transfer from the primary to the secondary coolant - i.e. water - is done in heat exchangers with three-layer tubes. As there is no component cooling heat exchanger, it is advantageous that the layers that are in touch with the primary coolant form part of the wall of the containment. An emergency cooling system inside the containment is also made of three-layer tubes. The tubes of the primary loops have the shape of loops, helices, and spirals surrounding the reactor tank or a biological shield. Between the tubes and the safety wall there are maintenance areas which are accessible from the outside. The three-layer construction prevents a reaction of leaked-out or evaporated sodium with the secondary coolant. (DG) [de

  11. Effect of membrane and through-wall bending stresses on fatigue crack growth behavior and coolant leakage velocity

    International Nuclear Information System (INIS)

    Yoo, Yeon-Sik

    2003-11-01

    This study clarified the effect of a membrane and a through-wall bending stresses on fatigue crack growth behavior and coolant leakage velocity due to irregularity of crack surface. Each stress component relates to fatigue crack growth behavior directly in general and thus the wild-used K I solutions are anticipated to give good evaluation results on it. Meanwhile, it is necessary to notify that surface irregularity for coolant leakage assessment is made by stress history in nature. Surface irregularity is known to be largely classified into the following two aspects: surface roughness due to continuous crack opening and closure behavior and surface turnover due to cyclic bending stress dominance. Therefore, the deterministic parameters on resistance of coolant leakage by surface irregularity are considered to be not only stress history but crack opening behavior. (author)

  12. Determination of mean molecular weights in organic reactor coolants. III. Differential cryoscopy with thermoelectric thermometer

    International Nuclear Information System (INIS)

    Becerro, E.; Carreira, M.

    1968-01-01

    The solubility problems raised by some components of the polymeric residue of irradiated polyphenolic coolants, which make it necessary to operate with very small samples, have been solved by means of a differential cryoscopic technique using a thermoelectric thermometer (thermal) as sensitive element. The method is based on the direct measurement of the difference between the freezing points of the investigated solution and of a reference solution whose concentration may be changed at will. The change of Δ V (mV) versus c(molal) is linear, the equivalent point being determined either analytically or graphically depending on the required accuracy. The method has been tested by measurements on pure polyphenyls, using diphenyl ether as solvent. It has been also applied to the main prospective coolants for the DON reactor. Working with 10 2 molal solutions the accuracy is better than ± 2 per cent. (Author) 2 refs

  13. Assessment of the effect of level change in the pressurizer in the event of main coolant pump failure. (Draft)

    International Nuclear Information System (INIS)

    Sommer, J.

    1999-09-01

    A series of analyses have been performed to assess the effect of the Dukovany NPP instrumentation and control system upgrade on plant safety. The analyses, however, failed to include change in the pressurizer level control. Change in pressurizer level control serves to reduce the amount of coolant which the normal makeup system must supply to the primary circuit after reactor shutdown. This is achieved so that in the nominal regime the pressurizer is controlled to a higher water level while the required level of the shut-down reactor remains as before. The present report was aimed at testing on a case of main coolant pump failure, how a change in the pressurizer level control affects the results of the safety analyses, or in other words, how reliable the results of analyses performed using the initial level control model are. (author)

  14. Process and kinetics of the fundamental radiation-electrochemical reactions in the primary coolant loop of nuclear reactors

    International Nuclear Information System (INIS)

    Kozomara-Maic, S.

    1987-06-01

    In spite of the rather broad title of this report, its major part is devoted to the corrosion problems at the RA reactor, i.e. causes and consequences of the reactor shutdown in 1979 and 1982. Some problems of reactor chemistry are pointed out because they are significant for future reactor operation. The final conclusion of this report is that corrosion processes in the primary coolant circuit of the nuclear reactor are specific and that radiation effects cannot be excluded when processes and reaction kinetics are investigated. Knowledge about the kinetics of all the chemical reactions occurring in the primary coolant loop are of crucial significance for safe and economical reactor operation [sr

  15. Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

    International Nuclear Information System (INIS)

    Soares, Alexandre de Souza

    2014-01-01

    The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm 2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

  16. Collective of mechatronics circuit

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1987-02-15

    This book is composed of three parts, which deals with mechatronics system about sensor, circuit and motor. The contents of the first part are photo sensor of collector for output, locating detection circuit with photo interrupts, photo sensor circuit with CdS cell and lamp, interface circuit with logic and LED and temperature sensor circuit. The second part deals with oscillation circuit with crystal, C-R oscillation circuit, F-V converter, timer circuit, stability power circuit, DC amp and DC-DC converter. The last part is comprised of bridge server circuit, deformation bridge server, controlling circuit of DC motor, controlling circuit with IC for PLL and driver circuit of stepping motor and driver circuit of Brushless.

  17. Collective of mechatronics circuit

    International Nuclear Information System (INIS)

    1987-02-01

    This book is composed of three parts, which deals with mechatronics system about sensor, circuit and motor. The contents of the first part are photo sensor of collector for output, locating detection circuit with photo interrupts, photo sensor circuit with CdS cell and lamp, interface circuit with logic and LED and temperature sensor circuit. The second part deals with oscillation circuit with crystal, C-R oscillation circuit, F-V converter, timer circuit, stability power circuit, DC amp and DC-DC converter. The last part is comprised of bridge server circuit, deformation bridge server, controlling circuit of DC motor, controlling circuit with IC for PLL and driver circuit of stepping motor and driver circuit of Brushless.

  18. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1984-01-01

    A review of the French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all actual leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by compliance with the criteria defined in the operating technical specifications. (orig.)

  19. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1986-01-01

    A review of the French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all actual leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by compliance with the criteria defined in the operating technical specifications

  20. Power supplyer for reactor coolant recycling pump

    International Nuclear Information System (INIS)

    Nara, Hiroshi; Okinaka, Yo.

    1991-01-01

    The present invention concerns a variable voltage/variable frequency static power source (static power source) used as a power source for a coolants recycling pump motor of a nuclear power plant. That is, during lower power operation such as start up or shutdown in which stoppage of the power source gives less effect to a reactor core, power is supplied from a power system, a main power generator connected thereto or a high voltage bus in the plant or a common high voltage bus to the static power source. However, during rated power operation, power is supplied from the output of an axially power generator connected with a main power generator having an extremely great inertia moment to the static power device. With such a constitution, the static power device is not stopped by the lowering of the voltage due to a thunderbolt falling accident or the like to a power-distribution line suddenly occurred in the power system. Accordingly, reactor core flowrate is free from rapid decrease caused by the reduction of rotation speed of the recycling pump. Accordingly, disadvantgages upon operation control in the reactor core is not caused. (I.S.)

  1. Numerical model simulation of atmospheric coolant plumes

    International Nuclear Information System (INIS)

    Gaillard, P.

    1980-01-01

    The effect of humid atmospheric coolants on the atmosphere is simulated by means of a three-dimensional numerical model. The atmosphere is defined by its natural vertical profiles of horizontal velocity, temperature, pressure and relative humidity. Effluent discharge is characterised by its vertical velocity and the temperature of air satured with water vapour. The subject of investigation is the area in the vicinity of the point of discharge, with due allowance for the wake effect of the tower and buildings and, where application, wind veer with altitude. The model equations express the conservation relationships for mometum, energy, total mass and water mass, for an incompressible fluid behaving in accordance with the Boussinesq assumptions. Condensation is represented by a simple thermodynamic model, and turbulent fluxes are simulated by introduction of turbulent viscosity and diffusivity data based on in-situ and experimental water model measurements. The three-dimensional problem expressed in terms of the primitive variables (u, v, w, p) is governed by an elliptic equation system which is solved numerically by application of an explicit time-marching algorithm in order to predict the steady-flow velocity distribution, temperature, water vapour concentration and the liquid-water concentration defining the visible plume. Windstill conditions are simulated by a program processing the elliptic equations in an axisymmetrical revolution coordinate system. The calculated visible plumes are compared with plumes observed on site with a view to validate the models [fr

  2. Alternative protections for loss of coolant accidents

    International Nuclear Information System (INIS)

    Estevez, E.A.

    1997-01-01

    One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

  3. Corrosion problems with aqueous coolants, final report

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. A mechanistic model for predicting flow-assisted and general corrosion of carbon steel in reactor primary coolants

    International Nuclear Information System (INIS)

    Lister, D.

    2002-01-01

    Flow-assisted corrosion (FAC) of carbon steel in high-temperature lithiated water can be described with a model that invokes dissolution of the protective oxide film and erosion of oxide particles that are loosened as a result. General corrosion under coolant conditions where oxide is not dissolved is described as well. In the model, the electrochemistry of magnetite dissolution and precipitation and the effect of particle size on solubility move the dependence on film thickness of the diffusion processes (and therefore the corrosion rate) away from reciprocal. Particle erosion under dissolving conditions is treated stochastically and depends upon the fluid shear stress at the surface. The corrosion rate dependence on coolant flow under FAC conditions then becomes somewhat less than that arising purely from fluid shear (proportional to the velocity squared). Under non-dissolving conditions, particle erosion occurs infrequently and general corrosion is almost unaffected by flow For application to a CANDU primary circuit and its feeders, the model was bench-marked against the outlet feeder S08 removed from the Point Lepreau reactor, which furnished one value of film thickness and one of corrosion rate for a computed average coolant velocity. Several constants and parameters in the model had to be assumed or were optimised, since values for them were not available. These uncertainties are no doubt responsible for the rather high values of potential that evolved as steps in the computation. The model predicts film thickness development and corrosion rate for the whole range of coolant velocities in outlet feeders very well. In particular, the detailed modelling of FAC in the complex geometry of one outlet feeder (F11) is in good agreement with measurements. When the particle erosion computations are inserted in the balance equations for the circuit, realistic values of crud level are obtained. The model also predicts low corrosion rates and thick oxide films for inlet

  5. Definition of loss-of-coolant accident radiation source. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist.

  6. Hot gas path component

    Science.gov (United States)

    Lacy, Benjamin Paul; Kottilingam, Srikanth Chandrudu; Porter, Christopher Donald; Schick, David Edward

    2017-09-12

    Various embodiments of the disclosure include a turbomachine component. and methods of forming such a component. Some embodiments include a turbomachine component including: a first portion including at least one of a stainless steel or an alloy steel; and a second portion joined with the first portion, the second portion including a nickel alloy including an arced cooling feature extending therethrough, the second portion having a thermal expansion coefficient substantially similar to a thermal expansion coefficient of the first portion, wherein the arced cooling feature is located within the second portion to direct a portion of a coolant to a leakage area of the turbomachine component.

  7. Nondestructive control and periodic examination of primary reactor circuits

    International Nuclear Information System (INIS)

    Prot, A.C.; Saglio, R.

    A knowledge of failures in the different components of a nuclear power plant is important for ensuring the reliability of the plant. The primary coolant circuit is particularly important from the safety point of view, which is why so many countries are working on regulations concerning periodic inspection of this system. The authors start by describing the parallel development of such regulations and of the nondestructive testing techniques designed to meet their requirements. After stating these requirements, they discuss the progress made in France during the past few years in different areas: ultrasonic nondestructive testing, especially through the stainless steel lining of the reactor vessel; extension of the results to external testing (advantage inherent in the method), to the testing of the mixed welds of safe-ends and of austenitic stainless steel welds, and to the testing of the reactor cover bolts; improvement of an acoustic holography process (which is compared with the ultrasonic method); improvement of testing methods based on Foucault currents (especially for the inspection of steam generators); and improvement of detection and location by acoustic emission during hydraulic trials and continuous operation. In conclusion, the authors suggest the main lines of work that should be followed to achieve better non-destructive testing during the construction, entry into service and operation of nuclear power plants

  8. Corrosion particles in the primary coolant of VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vajda, N.; Molnar, Z., E-mail: vajdanor@gmail.com [RadAnal Ltd., Budapest (Hungary); Macsik, Z.; Szeles, E.; Hargittai, P., E-mail: szeles@iki.kfki.hu [Inst. of Isotopes of the Hungarian Academy of Sciences, Budapest (Hungary); Csordas, A.; Pinter, T., E-mail: csordas@aeki.kfki.hu [Atomic Energy Research Inst. of the Hungarian Academy of Sciences, Budapest (Hungary); Pinter, T., E-mail: pinter@npp.hu [Paks Nuclear Power Plant, Paks (Hungary)

    2010-07-01

    Corrosion and activity build-up processes are of major concern in ageing and life-extension of nuclear power reactors. Researches to study the migration of radioactive corrosion particles have been initiated at Paks Nuclear Power Plant (NPP), Hungary in order to better understand the corrosion of the primary circuit surfaces, the transport and activation of the particles of corrosion origin and their deposition on in-core and out-of-core surfaces. Radioactive corrosion particles were collected from the primary coolant and the steam generator surfaces of the 4 reactor units and subjected to detailed microanalytical and radioanalytical investigations. Scanning electron microscopy and energy dispersive X-ray microanalysis (SEM-EDX) were used to study the morphology and the composition of the matrix elements in the particles and the deposited corrosion layers. Particles identified by SEM-EDX were re-located under optical microscope by means of a coordinate transformation algorithm and were separated with a micromanipulator for further studies. Activities of γ emitting radionuclides were determined by high resolution γ spectrometry, and those of β decaying isotopes were measured by liquid scintillation (LS) spectrometry after radiochemical processing. High sensitivity of the nuclear measuring techniques allowed us to determine the low activity concentrations of the long-lived radionuclides, i.e. {sup 60}Co, {sup 54}Mn, {sup 63}Ni, {sup 55}Fe in the individual particles. Finally, high resolution sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS) was applied to determine the ultralow concentrations of Co, Fe, Ni in the same particles. Specific activities of {sup 60}Co/Co, {sup 54}Mn/Fe, {sup 55}Fe/Fe and {sup 63}Ni/Ni were derived from the measured activity and concentration data. Specific activities of the radioactive corrosion products reveal the history of activity buildup processes in the particle. Typically, Fe-Cr-Ni oxide particles formed as a

  9. Corrosion particles in the primary coolant of VVER-440 reactors

    International Nuclear Information System (INIS)

    Vajda, N.; Molnar, Z.; Macsik, Z.; Szeles, E.; Hargittai, P.; Csordas, A.; Pinter, T.; Pinter, T.

    2010-01-01

    Corrosion and activity build-up processes are of major concern in ageing and life-extension of nuclear power reactors. Researches to study the migration of radioactive corrosion particles have been initiated at Paks Nuclear Power Plant (NPP), Hungary in order to better understand the corrosion of the primary circuit surfaces, the transport and activation of the particles of corrosion origin and their deposition on in-core and out-of-core surfaces. Radioactive corrosion particles were collected from the primary coolant and the steam generator surfaces of the 4 reactor units and subjected to detailed microanalytical and radioanalytical investigations. Scanning electron microscopy and energy dispersive X-ray microanalysis (SEM-EDX) were used to study the morphology and the composition of the matrix elements in the particles and the deposited corrosion layers. Particles identified by SEM-EDX were re-located under optical microscope by means of a coordinate transformation algorithm and were separated with a micromanipulator for further studies. Activities of γ emitting radionuclides were determined by high resolution γ spectrometry, and those of β decaying isotopes were measured by liquid scintillation (LS) spectrometry after radiochemical processing. High sensitivity of the nuclear measuring techniques allowed us to determine the low activity concentrations of the long-lived radionuclides, i.e. 60 Co, 54 Mn, 63 Ni, 55 Fe in the individual particles. Finally, high resolution sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS) was applied to determine the ultralow concentrations of Co, Fe, Ni in the same particles. Specific activities of 60 Co/Co, 54 Mn/Fe, 55 Fe/Fe and 63 Ni/Ni were derived from the measured activity and concentration data. Specific activities of the radioactive corrosion products reveal the history of activity buildup processes in the particle. Typically, Fe-Cr-Ni oxide particles formed as a result of corrosion of the steel

  10. Method for removing cesium from aqueous liquid, method for purifying the reactor coolant in boiling water and pressurized water reactors and a mixed ion exchanged resin bed, useful in said purification

    International Nuclear Information System (INIS)

    Otte, J.N.A.; Liebmann, D.

    1989-01-01

    The invention relates to a method for removing cesium from an aqueous liquid, and to a resin bed containing a mixture of an anion exchange resin and cation exchange resin useful in said purification. In a preferred embodiment, the present invention is a method for purifying the reactor coolant of a presurized water or boiling water reactor. Said method, which is particularly advantageously employed in purifying the reactor coolant in the primary circuit of a pressurized reactor, comprises contacting at least a portion of the reactor coolant with a strong base anion exchange resin and the strong acid cation exchange resin derived from a highly cross-linked, macroporous copolymer of a monovinylidene aromatic and a cross-linking monomer copolymerizable therewith. Although the reactor coolant can sequentially be contacted with one resin type and thereafter with the second resin type, the contact is preferably conducted using a resin bed comprising a mixture of the cation and anion exchange resins. 1 fig., refs

  11. Comparative analysis of coolants for FBR of future nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I. [State Scientific Center of Russian Federation, Institute for Physics and Power Engineering named after Academician A.I. Leipusky, Kaluga Region (Russian Federation)

    2001-07-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR.

  12. Comparative analysis of coolants for FBR of future nuclear power

    International Nuclear Information System (INIS)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I.

    2001-01-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR

  13. Analog circuit design designing dynamic circuit response

    CERN Document Server

    Feucht, Dennis

    2010-01-01

    This second volume, Designing Dynamic Circuit Response builds upon the first volume Designing Amplifier Circuits by extending coverage to include reactances and their time- and frequency-related behavioral consequences.

  14. Application of an Underwater Robot in Reactor Coolant System

    International Nuclear Information System (INIS)

    Choi, Young-Soo; Jeong, Kyung-Min; Lee, Sung-Uk; Cho, Jai-Wan

    2006-01-01

    Nuclear energy is a major source of electric energy consumed in Korea. It has the advantage of other energy sources, nuclear energy is cost effective and little pollution. But the fearfulness of an accident and/or failure has scared us the utilization of nuclear energy extensively. So, the safety and reliability of nuclear power plants become more important. Inspection and maintenance of component should be achieved continuously. The RCS(reactor coolant system) of PWR(pressurized water reactor) has a role to cool down the reactor's temperature. Cooling water is injected through the SI(safety injection) nozzle into the cold leg of the primary loop. Thermal sleeves are attached inside the cylindrical SI nozzle to reduce the thermal shock of the cooling water to the weld zone of the safety injection nozzle. The human workers are susceptible to radiation exposure and manual handling machine is hard to access because of the complexity of the path. So, we developed and applied free running, tele-operated underwater vehicle to inspect SI nozzle close to the place. Tele-operated robot is useful to inspect and maintain the component of nuclear power plants to reduce the radiation exposure of human operators and improve the reliability of the operation in nuclear power plants. Underwater robot is comprised of two parts; one is robot vehicle and the other is remote control module. Underwater robot vehicle has 4 DOF(degree of freedom) of mobility and 1 DOF of camera observation. The task to inspect the internal of RCS in nuclear power plant is achieved successfully. And the reliability for the maintenance is increased by the aid of tele-operated robot

  15. Application of an Underwater Robot in Reactor Coolant System

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young-Soo; Jeong, Kyung-Min; Lee, Sung-Uk; Cho, Jai-Wan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    Nuclear energy is a major source of electric energy consumed in Korea. It has the advantage of other energy sources, nuclear energy is cost effective and little pollution. But the fearfulness of an accident and/or failure has scared us the utilization of nuclear energy extensively. So, the safety and reliability of nuclear power plants become more important. Inspection and maintenance of component should be achieved continuously. The RCS(reactor coolant system) of PWR(pressurized water reactor) has a role to cool down the reactor's temperature. Cooling water is injected through the SI(safety injection) nozzle into the cold leg of the primary loop. Thermal sleeves are attached inside the cylindrical SI nozzle to reduce the thermal shock of the cooling water to the weld zone of the safety injection nozzle. The human workers are susceptible to radiation exposure and manual handling machine is hard to access because of the complexity of the path. So, we developed and applied free running, tele-operated underwater vehicle to inspect SI nozzle close to the place. Tele-operated robot is useful to inspect and maintain the component of nuclear power plants to reduce the radiation exposure of human operators and improve the reliability of the operation in nuclear power plants. Underwater robot is comprised of two parts; one is robot vehicle and the other is remote control module. Underwater robot vehicle has 4 DOF(degree of freedom) of mobility and 1 DOF of camera observation. The task to inspect the internal of RCS in nuclear power plant is achieved successfully. And the reliability for the maintenance is increased by the aid of tele-operated robot.

  16. Estimation of the mechanical behavior of irradiated coolant channels at a nuclear plant for its decomissing

    International Nuclear Information System (INIS)

    Piquin, Ruben; Zanni, Pablo

    2003-01-01

    The widespread replacement of reactor internals generates a substantial volume of active material.It is essential to work with these components at least in a partial way before the next planned stop.Due to the fact that the reactor internals pool and the storage pools for irradiated nuclear fuel have limited capacities, it has been proposed to compact an experimental shift of 50 irradiated coolant channels, that are currently placed in storage pools.Basically the processed waste will be put in baskets at the bottom of the storage pools.The alternative choice proposes to divide an irradiation coolant channel tube into different parts: stainless steel section, zircaloy-4 section and stainless steel section with hardened zones with cobalt alloys named Estelite-6.Having planned the constructive and operative solutions, the mechanical characterization of the different parts of the channel tube remains to be done.In the present paper, the necessary compacted strength of the irradiation coolant channel tube will be estimated for the stainless steel section and for the zircaloy-4 section, starting from experiment with unirradiated material and considering effects of radiation damage and hydrides on the ductility.These results will be used to design the necessary compacted tools for the semi-industrial installation

  17. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems

    Energy Technology Data Exchange (ETDEWEB)

    Titov, Gene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-04-05

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  18. refining device for helium coolants capable of continuous tritium recovery for use with HTGR type reactors

    International Nuclear Information System (INIS)

    Tone, Hirohito.

    1981-01-01

    Purpose: To separate and recover hydrogen and tritium from helium coolants by a device comprising a tritium permeable metal film, a gas compressor and a recovery store container. Constitution: Impurity gas-containing helium sent from a primary helium coolant circuit exit of an HTGR type reactor to a helium refining system is introduced by way of a high temperature heat exchanger, a gas recycling device and a medium temperature heat exchanger to an absorber bed filled with absorber, where H 2 O and CO 2 in the impurity gas-containing helium are removed. Then, Kr, Xe, Ar, O 2 , N 2 , CO, CH 3 are removed and the remaining impurities, tritium and hydrogen, are introduced via each of heat exchangers into a tritium absorber bed. Then, they are permeated through a tritium permeable metal film, transferred to the low pressure side of a tritium permeation device and then compressed and stored in a tritium recovery store container by a gas compressor connected to the low pressure side. (Seki, T.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Roussel, G. [AIB Vincotte Nuclear, Brussels (Belgium)

    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.

  20. Analog circuit design designing waveform processing circuits

    CERN Document Server

    Feucht, Dennis

    2010-01-01

    The fourth volume in the set Designing Waveform-Processing Circuits builds on the previous 3 volumes and presents a variety of analog non-amplifier circuits, including voltage references, current sources, filters, hysteresis switches and oscilloscope trigger and sweep circuitry, function generation, absolute-value circuits, and peak detectors.

  1. High-pressure coolant effect on the surface integrity of machining titanium alloy Ti-6Al-4V: a review

    Science.gov (United States)

    Liu, Wentao; Liu, Zhanqiang

    2018-03-01

    Machinability improvement of titanium alloy Ti-6Al-4V is a challenging work in academic and industrial applications owing to its low thermal conductivity, low elasticity modulus and high chemical affinity at high temperatures. Surface integrity of titanium alloys Ti-6Al-4V is prominent in estimating the quality of machined components. The surface topography (surface defects and surface roughness) and the residual stress induced by machining Ti-6Al-4V occupy pivotal roles for the sustainability of Ti-6Al-4V components. High-pressure coolant (HPC) is a potential choice in meeting the requirements for the manufacture and application of Ti-6Al-4V. This paper reviews the progress towards the improvements of Ti-6Al4V surface integrity under HPC. Various researches of surface integrity characteristics have been reported. In particularly, surface roughness, surface defects, residual stress as well as work hardening are investigated in order to evaluate the machined surface qualities. Several coolant parameters (including coolant type, coolant pressure and the injection position) deserve investigating to provide the guidance for a satisfied machined surface. The review also provides a clear roadmap for applications of HPC in machining Ti-6Al4V. Experimental studies and analysis are reviewed to better understand the surface integrity under HPC machining process. A distinct discussion has been presented regarding the limitations and highlights of the prospective for machining Ti-6Al4V under HPC.

  2. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  3. Unique rod lens/video system designed to observe flow conditions in emergency core coolant loops of pressurized water reactors

    International Nuclear Information System (INIS)

    Carter, G.W.

    1979-01-01

    Techniques and equipment are described which are used for video recordings of the single- and two-phase fluid flow tests conducted with the PKL Spool Piece Measurement System designed by Lawrence Livermore Laboratory and EG and G Inc. The instrumented spool piece provides valuable information on what would happen in pressurized water reactor emergency coolant loops should an accident or rupture result in loss of fluid. The complete closed-circuit television video system, including rod lens, light supply, and associated spool mounting fixtures, is discussed in detail. Photographic examples of test flows taken during actual spool piece system operation are shown

  4. Analytical Study on Thermal and Mechanical Design of Printed Circuit Heat Exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Su-Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kim, Eung-Soo [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-09-01

    The analytical methodologies for the thermal design, mechanical design and cost estimation of printed circuit heat exchanger are presented in this study. In this study, three flow arrangements of parallel flow, countercurrent flow and crossflow are taken into account. For each flow arrangement, the analytical solution of temperature profile of heat exchanger is introduced. The size and cost of printed circuit heat exchangers for advanced small modular reactors, which employ various coolants such as sodium, molten salts, helium, and water, are also presented.

  5. Experimental research to investigate the performance of bio coolant when turning of mild carbon steel

    Science.gov (United States)

    Agus Susanto, Tri; Nur, Rusdi

    2017-04-01

    Some literatures have been reported that the using bio coolant show better lubricating and cooling performances and reduce the occupational health risks associated with petroleum-oil-based coolant since they have lower toxicity. This paper investigates the effect the cutting conditions on the surface roughness through turning of mild carbon steel using dry, coolant and bio coolant. Measurement of surface roughness was conducted and then compared with the change of the cutting conditions. The relationship between surface roughness and cutting conditions was created in a curve for different of the cutting speed and coolant. The results indicate that the surface roughness was reduced when the speed of cutting is set to the highest level for all of coolant conditions (dry, coolant, and bio coolant) and constant of DOC and feed. The surface roughness had better performance using bio coolant than coolant conventional (mineral fluid).

  6. Open Component Portability Infrastructure (OPENCPI)

    Science.gov (United States)

    2013-03-01

    Integrated Circuit ( ASIC ) processor components. As used here, the HDL was applied to the Verilog, the Very High Speed Integrated Circuit (VHSIC...and a number of existing tests were migrated and ported into it. The motivation for selecting a unit test framework was to improve and enforce the...Interface ASIC Application-Specific Integrated Circuit BSV Bluespec System Verilog CBD Component-Based Development CDK Component Developer’s Kit

  7. Corrosion and deposition of metal oxides in the primary coolant system of PWR's

    International Nuclear Information System (INIS)

    Morrison, Jonathan; Hewett, John; Cooper, Chris; Ponton, Clive; Connolly, Brian; Banks, Andrew

    2014-01-01

    Regardless of how well processed and preconditioned reactor construction materials have been prior to operation, they will inevitably corrode. The products of corrosion, metal oxides with low thermodynamic solubility levels, can be found deposited throughout the reactor coolant circuit if they have not been removed by the plant's water treatment systems. Deposition of these corrosion products in the primary coolant circuit of a PWR represent a concern in terms of operational safety, efficiency, financial cost and longevity of the plant. The on-going work at the University of Birmingham seeks to study three main areas; fundamental corrosion behaviour of stainless steel, with specific focus on in the effect of surface pre-treatment, the solubility of the main metal oxides produced by stainless steels corroding, and the deposition of metal oxides within flow restrictions at high flow velocities. Several pieces of bespoke test equipment have been designed and built to address each area of interest individually. These include a rig for studying the corrosion of a particular material as a function of time, which consists of a series of cells into which samples are mounted and removed after a period of time and another rig for studying the solubility of a given metal oxide. Both of these pieces of equipment operate at up to 300°C and can maintain a pressure of up to 120 bar. A water loop for the study of deposition of corrosion products within flow restrictions has also been built and is in operation; this loop is capable of producing flow rates in excess of 12 ms -1 through a test section 6 mm in diameter, while maintaining temperatures up to 300°C and pressures of up to 120 bar. (author)

  8. CREEP in tubes: theoretical notes and application to PEC primary coolant circuit

    International Nuclear Information System (INIS)

    Cesari, F.; Calcedonio Cappello, C.

    1975-01-01

    Creep and stress relaxation in the hot leg of PEC reactor are analitically examined, considering also the effects of varying loads and thermal transients. The expression, used to describe creep phenomena, are of the ''time-hardening'' type, so that the strain rate is a function only of the actual stress and the current time. A qualitative approach is attempted to describe the history of a part, when subjected to real cycles of loads/temperatures. Although in cases of rapidly varying or abrupt cyclic stresses the use of a time-hardening expression may lead to nearly absurd results, discussion on the better agreement with experiments of time or stress hardening laws is not presented. A brief illustration of physical phenomena bases and a conclusive chapter with a certain number of analytical appendices to analyse creep on simple structures due to many loads are also included

  9. Experimental interaction of magma and “dirty” coolants

    Science.gov (United States)

    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 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 that even modest concentrations of sediment in water will significantly limit heat transfer during non-explosive magma-water interactions. At high concentrations, the dramatic reduction in cooling efficiency

  10. Delta connected resonant snubber circuit

    Science.gov (United States)

    Lai, J.S.; Peng, F.Z.; Young, R.W. Sr.; Ott, G.W. Jr.

    1998-01-20

    A delta connected, resonant snubber-based, soft switching, inverter circuit achieves lossless switching during dc-to-ac power conversion and power conditioning with minimum component count and size. Current is supplied to the resonant snubber branches solely by the dc supply voltage through the main inverter switches and the auxiliary switches. Component count and size are reduced by use of a single semiconductor switch in the resonant snubber branches. Component count is also reduced by maximizing the use of stray capacitances of the main switches as parallel resonant capacitors. Resonance charging and discharging of the parallel capacitances allows lossless, zero voltage switching. In one embodiment, circuit component size and count are minimized while achieving lossless, zero voltage switching within a three-phase inverter. 36 figs.

  11. Delta connected resonant snubber circuit

    Science.gov (United States)

    Lai, Jih-Sheng; Peng, Fang Zheng; Young, Sr., Robert W.; Ott, Jr., George W.

    1998-01-01

    A delta connected, resonant snubber-based, soft switching, inverter circuit achieves lossless switching during dc-to-ac power conversion and power conditioning with minimum component count and size. Current is supplied to the resonant snubber branches solely by the dc supply voltage through the main inverter switches and the auxiliary switches. Component count and size are reduced by use of a single semiconductor switch in the resonant snubber branches. Component count is also reduced by maximizing the use of stray capacitances of the main switches as parallel resonant capacitors. Resonance charging and discharging of the parallel capacitances allows lossless, zero voltage switching. In one embodiment, circuit component size and count are minimized while achieving lossless, zero voltage switching within a three-phase inverter.

  12. Corrosion and radioactivity in the primary circuit of a sodium cooled reactor. The computer code ''Corona''

    International Nuclear Information System (INIS)

    Costa, L.; Fremont, R. de; Mougniot, J.C.; Msika, D.

    1976-01-01

    In this paper the proble ms of corrosion and activity in the primary coolant circuit of a sodium cooled fast reactor are treated. The main features of Corona, a computer code which has been used to perform the calculations, are outlined

  13. Noise in biological circuits.

    Science.gov (United States)

    Simpson, Michael L; Cox, Chris D; Allen, Michael S; McCollum, James M; Dar, Roy D; Karig, David K; Cooke, John F

    2009-01-01

    Noise biology focuses on the sources, processing, and biological consequences of the inherent stochastic fluctuations in molecular transitions or interactions that control cellular behavior. These fluctuations are especially pronounced in small systems where the magnitudes of the fluctuations approach or exceed the mean value of the molecular population. Noise biology is an essential component of nanomedicine where the communication of information is across a boundary that separates small synthetic and biological systems that are bound by their size to reside in environments of large fluctuations. Here we review the fundamentals of the computational, analytical, and experimental approaches to noise biology. We review results that show that the competition between the benefits of low noise and those of low population has resulted in the evolution of genetic system architectures that produce an uneven distribution of stochasticity across the molecular components of cells and, in some cases, use noise to drive biological function. We review the exact and approximate approaches to gene circuit noise analysis and simulation, and review many of the key experimental results obtained using flow cytometry and time-lapse fluorescent microscopy. In addition, we consider the probative value of noise with a discussion of using measured noise properties to elucidate the structure and function of the underlying gene circuit. We conclude with a discussion of the frontiers of and significant future challenges for noise biology. (c) 2009 John Wiley & Sons, Inc.

  14. Management of the Post-Shuttle Extravehicular Mobility Unit (EMU) Water Circuits

    Science.gov (United States)

    Steele, John W.; Etter, David; Rector, Tony; Hill, Terry; Wells, Kevin

    2011-01-01

    The EMU incorporates two separate water circuits for the rejection of metabolic heat from the astronaut and the cooling of electrical components. The first (the Transport Water Loop) circulates in a semi-closed-loop manner and absorbs heat into a Liquid Coolant and Ventilation Garment (LCVG) warn by the astronaut. The second (the Feed Water Loop) provides water to a cooling device (Sublimator) with a porous plate, and that water subsequently sublimates to space vacuum. The cooling effect from the sublimation of this water translates to a cooling of the LCVG water that circulates through the Sublimator. Efforts are underway to streamline the use of a water processing kit (ALCLR) that is being used to periodically clean and disinfect the Transport Loop Water. Those efforts include a fine tuning of the duty cycle based on a review of prior performance data as well as an assessment of a fixed installation of this kit into the EMU backpack or within on-orbit EMU interface hardware. Furthermore, testing is being conducted to ensure compatibility between the International Space Station (ISS) Water Processor Assembly (WPA) effluent and the EMU Sublimator as a prelude to using the WPA effluent as influent to the EMU Feed Water loop. This work is undertaken to reduce the crew-time and logistics burdens for the EMU, while ensuring the long-term health of the EMU water circuits for a post-Shuttle 6-year service life.

  15. Nuclear code case development of printed-circuit heat exchangers with thermal and mechanical performance testing

    Energy Technology Data Exchange (ETDEWEB)

    Aakre, Shaun R. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Mechanical Engineering; Jentz, Ian W. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Mechanical Engineering; Anderson, Mark H. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Mechanical Engineering

    2018-03-27

    The U.S. Department of Energy has agreed to fund a three-year integrated research project to close technical gaps involved with compact heat exchangers to be used in nuclear applications. This paper introduces the goals of the project, the research institutions, and industrial partners working in collaboration to develop a draft Boiler and Pressure Vessel Code Case for this technology. Heat exchanger testing, as well as non-destructive and destructive evaluation, will be performed by researchers across the country to understand the performance of compact heat exchangers. Testing will be performed using coolants and conditions proposed for Gen IV Reactor designs. Preliminary observations of the mechanical failure mechanisms of the heat exchangers using destructive and non-destructive methods is presented. Unit-cell finite element models assembled to help predict the mechanical behavior of these high-temperature components are discussed as well. Performance testing methodology is laid out in this paper along with preliminary modeling results, an introduction to x-ray and neutron inspection techniques, and results from a recent pressurization test of a printed-circuit heat exchanger. The operational and quality assurance knowledge gained from these models and validation tests will be useful to developers of supercritical CO2 systems, which commonly employ printed-circuit heat exchangers.

  16. Radiation-hardened CMOS integrated circuits

    International Nuclear Information System (INIS)

    Derbenwick, G.F.; Hughes, R.C.

    1977-01-01

    Electronic circuits that operate properly after exposure to ionizing radiation are necessary for nuclear weapon systems, satellites, and apparatus designed for use in radiation environments. The program to develop and theoretically model radiation-tolerant integrated circuit components has resulted in devices that show an improvement in hardness up to a factor of ten thousand over earlier devices. An inverter circuit produced functions properly after an exposure of 10 6 Gy (Si) which, as far as is known, is the record for an integrated circuit

  17. Seismic response analysis of a reactor coolant pump

    International Nuclear Information System (INIS)

    Villasor, A.P. Jr.

    1975-01-01

    The Westinghouse Type 93A reactor coolant pump (RCP) is analyzed for seismic response. Simply described the RCP is a vertical, single-stage, centrifugal pump designed to move 90,000gpm (568m 3 /sec) of water and driven by a 6000hp motor for use in the PWR primary system. The RCP assembly is generally axisymmetric and is modeled using 3-dimensional elements to represent its parts. These finite elements are of the types normally found in general purpose computer programs such as ANSYS or NASTRAN. The structural frame and the rotating shaft are the principal branches of the model. Each consists of a series of pipe elements complemented by mass elements. Orthogonal sets of linear spring elements connect the branches at the bearings and possibly at each labyrinth. Fluid elements are added to include the interaction between the shaft and the pump case through the intervening water mass. To complete the model, stiffness matrix elements representing the support structure and the neighboring loop piping are attached. It is impractical to idealize faithfully each geometric irregularity. Several adjacent sections are combined into one suitable element with total stiffness and mass equivalence. The number of elements in the model is thus minimized. Shear deflection of the pipe elements is considered; mass and mass inertia are lumped at nodal points, as needed, to compensate for the actual material distribution. The in-depth analysis in this work, believed to have no published precedent, has found the RCP adequate to withstand the imposed seismic loading. All component stresses were within the applicable faulted criteria and the relative movements between closely-mated parts fell inside their nominal clearance limits

  18. Reduction of radioactivity in the reactor coolant system

    International Nuclear Information System (INIS)

    Takashima, Yoshie; Oosumi, Katsumi; Uchida, Shunsuke; Izumiya, Masakiyo

    1980-01-01

    In evaluating the radioactivity in a reactor coolant system so far, the fission products leaking from fuel rods and the corrosion products activated in a core have been the objects to be considered. However, the target of dose rate reduction is being concentrated only to the suppression of the generation of activated corrosion products, as the former problem was solved by the improvement of operating techniques and fuel manufacturing workmanship. In examining BWR cooling systems, the sources of generating corrosion products are mainly the component materials of feed water and condensate systems and the internal structures of a reactor. Of these, the contribution to dose rate by the former is much larger. Accordingly the relationship between the concentration of corrosion products in feed water and the piping surface dose rate in recirculation system has been analyzed for Shimane and Tsuruga nuclear power stations, investigating the operation results. This analysis has clarified that the insoluble iron cruds and cobalt ions carried into the reactors from the feed water systems were playing an important role for raising the dose rate. The investigation into the suppression of these matters has been done as the main countermeasure for the reduction of radioactivity. As a result, it is expected that the dose rate can be suppressed to 100 mR/h or less in almost all plants if the water quality control is practiced so as to maintain the iron cruds in feed water to the level of 1 ppb or so. The oxygen injection to feed water systems has also large effect on the improvement of iron crud rejection. (Wakatsuki, Y.)

  19. The effect of constraint on fuel-coolant interactions in a confined geometry

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.; Corradini, M.L. [Univ. of Wisconsin, Madison, WI (United States)

    1995-09-01

    A Fuel-Coolant Interaction (FCI or vapor explosion) is the phenomena in which a hot liquid rapidly transfers its internal energy into a surrounding colder and more volatile liquid. The energetics of such a complex multi-phase and multi-component phenomenon is partially determined by the surrounding boundary conditions. As one of the boundary conditions, we studied the effect of constraint on FCIs. The WFCI-D series of experiments were performed specifically to observe this effect. The results from these and our previous WFCI tests as well as those of other investigators are compared.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    Peyrouty, P.

    1996-01-01

    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

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

    International Nuclear Information System (INIS)

    Peyrouty, P.

    1997-01-01

    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 enables 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 represented by deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible. (author)

  3. Solid-state circuits

    CERN Document Server

    Pridham, G J

    2013-01-01

    Solid-State Circuits provides an introduction to the theory and practice underlying solid-state circuits, laying particular emphasis on field effect transistors and integrated circuits. Topics range from construction and characteristics of semiconductor devices to rectification and power supplies, low-frequency amplifiers, sine- and square-wave oscillators, and high-frequency effects and circuits. Black-box equivalent circuits of bipolar transistors, physical equivalent circuits of bipolar transistors, and equivalent circuits of field effect transistors are also covered. This volume is divided

  4. Circuit analysis for dummies

    CERN Document Server

    Santiago, John

    2013-01-01

    Circuits overloaded from electric circuit analysis? Many universities require that students pursuing a degree in electrical or computer engineering take an Electric Circuit Analysis course to determine who will ""make the cut"" and continue in the degree program. Circuit Analysis For Dummies will help these students to better understand electric circuit analysis by presenting the information in an effective and straightforward manner. Circuit Analysis For Dummies gives you clear-cut information about the topics covered in an electric circuit analysis courses to help

  5. Current limiter circuit system

    Energy Technology Data Exchange (ETDEWEB)

    Witcher, Joseph Brandon; Bredemann, Michael V.

    2017-09-05

    An apparatus comprising a steady state sensing circuit, a switching circuit, and a detection circuit. The steady state sensing circuit is connected to a first, a second and a third node. The first node is connected to a first device, the second node is connected to a second device, and the steady state sensing circuit causes a scaled current to flow at the third node. The scaled current is proportional to a voltage difference between the first and second node. The switching circuit limits an amount of current that flows between the first and second device. The detection circuit is connected to the third node and the switching circuit. The detection circuit monitors the scaled current at the third node and controls the switching circuit to limit the amount of the current that flows between the first and second device when the scaled current is greater than a desired level.

  6. Embedded systems circuits and programming

    CERN Document Server

    Sanchez, Julio

    2012-01-01

    During the development of an engineered product, developers often need to create an embedded system--a prototype--that demonstrates the operation/function of the device and proves its viability. Offering practical tools for the development and prototyping phases, Embedded Systems Circuits and Programming provides a tutorial on microcontroller programming and the basics of embedded design. The book focuses on several development tools and resources: Standard and off-the-shelf components, such as input/output devices, integrated circuits, motors, and programmable microcontrollers The implementat

  7. Effect of stratified coolant voiding on the reactivity and distribution of neutron flux of a CANDU-6 cell; Effet d'une vidange stratifiee du caloporteur sur la reactivite et la distribution du flux neutronique d'une cellule CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Massicotte, M. [Ecole Polytechnique de Montreal, Dept. de genie physique, Montreal, Quebec (Canada)

    2008-07-01

    The fuel bundles of CANDU-6 reactor are inserted into horizontal tubes where heavy water coolant circulates. In case of a breach in the coolant circuit, coolant in the tube empties from the top to the bottom. This results in a rapid rise in the fuel temperature that can exceed the safety limits. Furthermore, the presence of a vacuum in the tube will affect the safety of the reactor. The computer code DRAGON developed at the Ecole Polytechnique de Montreal allows the analysis and simulations of such a scenario.

  8. Development of larval motor circuits in Drosophila.

    Science.gov (United States)

    Kohsaka, Hiroshi; Okusawa, Satoko; Itakura, Yuki; Fushiki, Akira; Nose, Akinao

    2012-04-01

    How are functional neural circuits formed during development? Despite recent advances in our understanding of the development of individual neurons, little is known about how complex circuits are assembled to generate specific behaviors. Here, we describe the ways in which Drosophila motor circuits serve as an excellent model system to tackle this problem. We first summarize what has been learned during the past decades on the connectivity and development of component neurons, in particular motor neurons and sensory feedback neurons. We then review recent progress in our understanding of the development of the circuits as well as studies that apply optogenetics and other innovative techniques to dissect the circuit diagram. New approaches using Drosophila as a model system are now making it possible to search for developmental rules that regulate the construction of neural circuits. © 2012 The Authors Development, Growth & Differentiation © 2012 Japanese Society of Developmental Biologists.

  9. Influence of coolant motion on structure of hardened steel element

    Directory of Open Access Journals (Sweden)

    A. Kulawik

    2008-08-01

    Full Text Available Presented paper is focused on volumetric hardening process using liquid low melting point metal as a coolant. Effect of convective motion of the coolant on material structure after hardening is investigated. Comparison with results obtained for model neglecting motion of liquid is executed. Mathematical and numerical model based on Finite Element Metod is described. Characteristic Based Split (CBS method is used to uncouple velocities and pressure and finally to solve Navier-Stokes equation. Petrov-Galerkin formulation is employed to stabilize convective term in heat transport equation. Phase transformations model is created on the basis of Johnson-Mehl and Avrami laws. Continuous cooling diagram (CTPc for C45 steel is exploited in presented model of phase transformations. Temporary temperatures, phases participation, thermal and structural strains in hardening element and coolant velocities are shown and discussed.

  10. Estimates of limits to fuel/coolant mixing

    International Nuclear Information System (INIS)

    Park, G.C.; Corradini, M.L.

    1991-06-01

    The vapor explosion process can involve the mixing of fuel with coolant prior to the explosion. A number of analysts have suggested limits to the amount of fuel/coolant mixing that could occur within the reactor vessel following a core melt accident. Past models are reviewed and a simplified approach is suggested to estimate a limit on the amount of fuel/coolant mixing possible. The approach uses concepts first advanced by Fauske in a different way. The results indicate that fuel temperature, ambient pressure and in particular the mixing length scale D mix are important parameters. For large values of D mix the fuel mass mixed in-vessel is limited to the range of 1-12 metric tons (1-10% of the core mass). (orig.) [de

  11. LOFT advanced densitometer for nuclear loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Johnson, L.O.; Lassahn, G.D.; Wood, D.B.

    1979-01-01

    A ''nuclear hardened'' gamma densitometer, a device which uses radiation attenuation to measure fluid density in the presence of a background radiation field, is described. Data from the nuclear hardened gamma densitometer are acquired by time sampling the coolant fluid piping and fluid attenuated source energy spectrum. The data are used to calculate transient coolant fluid cross sectional average density to analyze transient mass flow and other thermal-hydraulic characteristics during the Loss-of-Fluid Test (LOFT) loss-of-coolant experiments. The nuclear hardened gamma densitometer uses a pulse height analysis or energy discrimination, pulse counting technique which makes separation of the gamma radiation source signal from the reactor generated gamma radiation background noise signal possible by processing discrete pulses which retain their pulse amplitude information

  12. A study on the characteristics of alternative coolants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Ji Young; Kim, B. H.; Kim, T. J.; Jeong, K. C.; Choi, Y. D.; Choi, J. H.; Hwang, S. T

    2000-12-01

    The role of the coolant in liquid metal fast breeder reactor is very important for reasons of system safety. Recently, it has revealed that lead and lead-bismuth alloy show good safety characteristics as a fast reactor coolant compared to the sodium, such as low chemical activity, high boiling temperature and more negative void coefficient. So many countries take interest in these metals. The objectives of this project are to study the characteristics of heavy liquid metals(lead, lead-bismuth alloy) and to provide valuble information useful for the estimate the possibilities of its as the alternative coolant materials. An intensive research was performed into the global development status, basic properties, safety assurance methods, and direction of research in the futures and so on.

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

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    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.

  14. Final report on the in situ testing of electrical components and devices at TMI-2

    International Nuclear Information System (INIS)

    Soberano, F.T.

    1984-06-01

    A total of 88 electrical components and devices were in situ tested. Of these, 11 totally failed and 21 suffered degradation that varied from mild to severe. The equipment that failed or incurred severe degradation was located in areas of known environmental extremes. Several motor operated valves in the Reactor Building basement failed because of submersion in water. Others severely degraded from contamination tracking, resulting in the alteration of their circuit electrical characteristics - a circumstance that could compromise their designed function. One backup oil lift pump motor for a reactor coolant pump motor, although located well above the Reactor Building basement high water mark, failed because of a break in its armature and field circuits; this failure was surmised to be a result of corrosion. The limit switch of a Class 1E solenoid valve likewise failed due to moisture intrusion. Components that noticeably degraded exhibited anomalies, likely due to the incursion of moisture, that varied from high capacitance to increased circuit resistance. The effect of the other degenerating conditions that existed during the accident, such as high temperature, high radiation levels, and the hydrogen burn, could not be evaluated individually or synergistically

  15. Coolant mixing in pressurized water reactors. Pt. 2: Experimental equipment and simulation of the mixing. Final report

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Richter, K.H.; Weiss, F.P.; Kliem, S.; Rohde, U.

    2003-02-01

    The project aimed at the quantification of distributions of coolant parameters at the core entrance during transient processes in which coolant streams of different temperatures and/or boron concentrations arrive at the reactor inlet nozzles. The safety relevance is given by the consequences of boron dilution and cold water transients. With the help of experiments at a reactor model as well as by validating and applying three-dimensional fluid dynamics codes (CFX-4) and the development of a simplified mixing model (SAPR) tools were developed for the description of the conditions at the core inlet, which are necessary for the determination of the response of the core. By modelling the space and time dependent boron concentration or temperature distribution these tools allow the coupled neutron kinetic / thermal hydraulic simulation of boron dilution and cold water transients and the safety assessment of such scenarios. The project was limited to processes without steam generation. Furthermore, the coolant mixing inside the core was excluded from the investigation. In order adequately to describe multiple circulation of temperature or boron concentration perturbations in the primary circuit, mixing in the upper plenum was studied, too. For the solution of the task, the flexible test facility ROCOM (Rossendorf Coolant Mixing Model) was erected. ROCOM is a 1:5 model of the German KONVOI PWR. Due to the full representation of all four loops, the flexible control system of the test facility and the dedicated novel technique of mesh sensors for concentration measurements with high resolution in space and time, ROCOM now represents a leading facility in the world. In different experiment series, transient concentration distributions were recorded for all scenarios defined in the work plan. The results were used for the validation of CFX-4. Furthermore, the novel simplified mixing model SAPR was successfully validated. It serves as an efficient interface for the coupling of

  16. Integrated Fuel-Coolant Interaction (IFCI 6.0) code

    International Nuclear Information System (INIS)

    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

  17. Estimation of fuel temperature increase during coolant boiloff accident

    International Nuclear Information System (INIS)

    Abe, Kiyoharu

    1981-10-01

    Fuel rod temperature increase during coolant boiloff accident due to unavailable ECCS was analyzed using a simple time dependent model. A standard case was first selected and its results clarified how is the fuel temperature behavior during the boiloff accident. Then the sensitivity studies for various parameters were performed to know what parameters have important roles. As a result of analyses, it was shown that the coolant mixture level in the core has a dominant effect on the core heatup and that fuel rod claddings will probably slump or melt before its full oxidation. (author)

  18. Probabilistic analyses of failure in reactor coolant piping

    International Nuclear Information System (INIS)

    Holman, G.S.

    1984-01-01

    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)

  19. Sensitivity calculation of the coolant temperature regarding the thermohydraulic parameters

    International Nuclear Information System (INIS)

    Andrade Lima, F.R. de; Silva, F.C. da; Thome Filho, Z.D.; Alvim, A.C.M.; Oliveira Barroso, A.C. de.

    1985-01-01

    It's studied the application of the Generalized Perturbation Theory (GPT) in the sensitivity calculation of thermalhydraulic problems, aiming at verifying the viability of the extension of the method. For this, the axial distribution, transient, of the coolant temperature in a PWR channel are considered. Perturbation expressions are developed using the GPT formalism, and a computer code (Tempera) is written, to calculate the channel temperature distribution and the associated importance function, as well as the effect of the thermalhydraulic parameters variations in the coolant temperature (sensitivity calculation). The results are compared with those from the direct calculation. (E.G.) [pt

  20. Control rod drive mechanism stator loss of coolant test

    International Nuclear Information System (INIS)

    Besel, L.; Ibatuan, R.

    1977-04-01

    This report documents the stator loss of coolant test conducted at HEDL on the lead unit Control Rod Drive Mechanism (CRDM) in February, 1977. The purpose of the test was to demonstrate scram capability of the CRDM with an uncooled stator and to obtain a time versus temperature curve of an uncooled stator under power. Brief descriptions of the test, hardware used, and results obtained are presented in the report. The test demonstrated that the CRDM could be successfully scrammed with no anomalies in both the two-phase and three-phase stator winding hold conditions after the respective equilibrium stator temperatures had been obtained with no stator coolant

  1. Resistor Extends Life Of Battery In Clocked CMOS Circuit

    Science.gov (United States)

    Wells, George H., Jr.

    1991-01-01

    Addition of fixed resistor between battery and clocked complementary metal oxide/semiconductor (CMOS) circuit reduces current drawn from battery. Basic idea to minimize current drawn from battery by operating CMOS circuit at lowest possible current consistent with use of simple, fixed off-the-shelf components. Prolongs lives of batteries in such low-power CMOS circuits as watches and calculators.

  2. RF DEMO ceramic helium cooled blanket, coolant and energy transformation systems

    International Nuclear Information System (INIS)

    Kovalenko, V.; Leshukov, A.; Poliksha, V.; Popov, A.; Strebkov, Yu.; Borisov, A.; Shatalov, G.; Demidov, V.; Kapyshev, V.

    2004-01-01

    RF DEMO-S reactor is a prototype of commercial fusion reactors for further generation. A blanket is the main element unit of the reactor design. The segment structure is the basis of the ceramic blanket. The segments mounting/dismounting operations are carried out through the vacuum vessel vertical port. The inboard/outboard blanket segment is the modules welded design, which are welded by back plate. The module contains the back plate, the first wall, lateral walls and breeding zone. The 9CrMoVNb steel is used as structural material. The module internal space formed by the first wall, lateral walls and back plate is used for breeding zone arrangement. The breeding zone design based upon the poloidal BIT (Breeder Inside Tube) concept. The beryllium is used as multiplier material and the lithium orthosilicate is used as breeder material. The helium at 0.1 MPa is used as purge gas. The cooling is provided by helium at 10 MPa. The coolant supply/return to the blanket modules are carrying out on the two independent circuits. The performed investigations of possible transformation schemes of DEMO-S blanket heat power into the electricity allowed to make a conclusion about the preferable using of traditional steam-turbine facility in the secondary circuit. (author)

  3. Study of the iodine transport mechanisms in a nuclear accidental medium by high temperature mass spectrometry in the framework of the 'CHIP' (Iodine Chemistry in the reactor coolant system) program. I - Thermodynamics of CsOH vaporization; Etude des mecanismes de transport de l iode en milieu nucleaire accidentel par spectrometrie de masse haute temperature dans le cadre du programme CHIP (CHimie de l Iode dans le circuit Primaire). I - Thermodynamique de la vaporisation de CsOH

    Energy Technology Data Exchange (ETDEWEB)

    Roki, F.Z.; Ohnet, M.N.; Jacquemain, D. [IRSN, DPAM, SEREA, Laboratoire d Essais Analytiques, Centre de Cadarache, BP 3, 13115 St Paul-Lez- Durance Cedex, (France); Chatillon, C. [Laboratoire de Thermodynamique et Physico-Chimie Metallurgiques (associe au CNRS, UMR 5614), BP 75 - 38402 Saint Martin d Heres, (France)

    2006-07-01

    Iodine is one of the fission products which vaporizes in a complex chemical medium and can be transported under different gaseous forms in the PWR primary coolant system during a severe nuclear accident. The 'CHIP' analytical program has been implemented to study in a first time, the quaternary complex systems of the type Cs-I-O-H. In this system, the CsOH(l or s) vaporization is an important element. A study by high temperature mass spectrometry shows that the vapor species are the CsOH(g), its dimers Cs{sub 2}O{sub 2}H{sub 2}(g) and its trimers Cs{sub 3}O{sub 3}H{sub 3}(g). For this last specie, no thermodynamic determination has been done until today. The quantitative exploitation of the vapor pressures leads to the determination of the dimerization enthalpy reaction: (CsOH){sub 2}(g)=2 CsOH(g) {delta}H(average)(298.14 K)136.5 {+-} 3.85 kJ.mol{sup -1} (calculated with the third thermodynamic law). (O.M.)

  4. Using noise to probe and characterize gene circuits.

    Science.gov (United States)

    Cox, Chris D; McCollum, James M; Allen, Michael S; Dar, Roy D; Simpson, Michael L

    2008-08-05

    Stochastic fluctuations (or "noise") in the single-cell populations of molecular species are shaped by the structure and biokinetic rates of the underlying gene circuit. The structure of the noise is summarized by its autocorrelation function. In this article, we introduce the noise regulatory vector as a generalized framework for making inferences concerning the structure and biokinetic rates of a gene circuit from its noise autocorrelation function. Although most previous studies have focused primarily on the magnitude component of the noise (given by the zero-lag autocorrelation function), our approach also considers the correlation component, which encodes additional information concerning the circuit. Theoretical analyses and simulations of various gene circuits show that the noise regulatory vector is characteristic of the composition of the circuit. Although a particular noise regulatory vector does not map uniquely to a single underlying circuit, it does suggest possible candidate circuits, while excluding others, thereby demonstrating the probative value of noise in gene circuit analysis.

  5. Electronic Control Circuit

    International Nuclear Information System (INIS)

    Kim, Sang Jin; Kim, Je Hwong; Cha, In Su

    2001-08-01

    This book consists of nine chapters, which are basis of thyristor about its use and classify, structure of thyristor like outside, inside, manufacturing and structure of thyristor sorts of thyristor family and sub thyristor, how to use thyristor such as standard chart, choice of thyristor and way of on and off, electric heat control circuit like control temperature of heating apparatus and cooker, lighting control circuit for light bulb, neon lamp, traffic signal, lamp regulator and strobe, motor control circuit including an inverter circuit transistor and speed control of direct motor by transistor, electric power source circuit and a spark-plug, applied circuit for protection of fire.

  6. Electronic devices and circuits

    CERN Document Server

    Pridham, Gordon John

    1972-01-01

    Electronic Devices and Circuits, Volume 3 provides a comprehensive account on electronic devices and circuits and includes introductory network theory and physics. The physics of semiconductor devices is described, along with field effect transistors, small-signal equivalent circuits of bipolar transistors, and integrated circuits. Linear and non-linear circuits as well as logic circuits are also considered. This volume is comprised of 12 chapters and begins with an analysis of the use of Laplace transforms for analysis of filter networks, followed by a discussion on the physical properties of

  7. Intuitive analog circuit design

    CERN Document Server

    Thompson, Marc

    2013-01-01

    Intuitive Analog Circuit Design outlines ways of thinking about analog circuits and systems that let you develop a feel for what a good, working analog circuit design should be. This book reflects author Marc Thompson's 30 years of experience designing analog and power electronics circuits and teaching graduate-level analog circuit design, and is the ideal reference for anyone who needs a straightforward introduction to the subject. In this book, Dr. Thompson describes intuitive and ""back-of-the-envelope"" techniques for designing and analyzing analog circuits, including transistor amplifi

  8. Compensation of equipment housing elements of reactor units with heavy liquid metal coolant vessel temperature deformations

    International Nuclear Information System (INIS)

    Lebedevich, V.; Ahmetshin, M.; Mendes, D.; Kaveshnikov, S.; Vinogradov, A.

    2015-01-01

    In Russia a lot of different versions of fast reactors (FRs) are investigated and one of these is FR cooled by liquid lead and liquid lead-bismuth alloy. In this poster we are interested by FR with concrete vessel; its components are placed in cavities inside the vessel, and connected by a channel system. During the installation the equipment components are placed in several equipment housings. Between these housings there are cavities with coolant. The alignment of the housings should be provided. It can be broken by irregular concrete vessel heating during FR starting or other transition regimes. Our goal is to suggest a list of designing steps to compensate temperature deformations of equipment housing elements. A simplified model of equipment housing was suggested. It consists of two cylinders - tunnels in the concrete vessel, separated by a cavity filled by coolant and inert gas. The bottom part was considered as heated to 420 C. degrees while in the top part temperature decreased to 45 C. degrees (on the concrete surface). According to this data, results show that temperature gradient leads to a concrete layer dislocation of about 12.5 mm, which can lead to damage and breaking alignment. We propose the following solution to compensate for temperature deformation: -) to chisel out part of the upper top of the insulating concrete; -) to install an adequate misalignment of equipment housing elements preliminary; and -) to use a torsion system like a piston-type device for providing additional strength in order to compensate deformation and vibrations

  9. Design of coolant distribution system (CDS) for ITER PF AC/DC converter

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Bin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Zhiquan, E-mail: zhquansong@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Fu, Peng; Xu, Xuesong; Li, Chuan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wang, Min; Dong, Lin [China International Nuclear Fusion Energy Program Execution Center, Beijing 100862 (China)

    2016-10-15

    Highlights: • System process and arrangement has been proposed to meet the multiple requirements from the converter system. • Thermal hydraulic analysis model has been developed to size and predict the system operation behavior. • Prototype test has been performed to validate the proposed design methodology. - Abstract: The Poloidal Field (PF) converter unit, playing an essential role in the plasma shape and position control in vertical and horizontal direction, which is an important part of ITER power supply system. As an important subsystem of the converter unit, the coolant distribution system has the function to distribute the cooling water from ITER component cooling water system (CCWS) to its main components at the required flow rate, pressure and temperature. This paper presents the thermal hydraulic design of coolant distribution system for the ITER PF converter unit. Different operational requirements of the PF converter unit regarding flow rate, temperature and pressure have been analyzed to design the system process and arrangement. A thermal-hydraulic analysis model has been built to size the system and predict the flow rate and temperature distribution of the system under the normal operation. Based on the system thermal-hydraulic analysis results, the system pressure profile has been plotted to evaluate the pressure behavior along each client flow path. A CDS prototype for the ITER PF converter has been constructed and some experiments have been performed on it. A good agreement of the flow distribution and temperature behavior between the simulated and test results validate the proposed design methodology.

  10. Numerical experimentation on convective coolant flow in Ghana ...

    African Journals Online (AJOL)

    Numerical experiments on one dimensional convective coolant flow during steady state operation of the Ghana Research Reactor-1 (GHARR-I) were performed to determine the thermal hydraulic parameters of temperature, density and flow rate. The computational domain was the reactor vessel, including the reactor core.

  11. Reactor coolant and associated systems in nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide outlines the design requirements for the reactor coolant and associated systems (RCAS) and the features required in order to achieve their safety functions. It covers design considerations for various reactor types and encompasses the safety aspects of the functions of the RCAS both during normal operation and following postulated initiating events, and to some extent also for decommissioning

  12. Stress analysis of a LWR main coolant pump

    International Nuclear Information System (INIS)

    Novobilsky, R.

    1985-01-01

    Because of its importance for the safe operation of a nuclear power plant the main coolant pump of a LWR has to be designed according to the same rules as the pressure vessel or the main coolant piping itself. As prescribed in the design specification, the stress and fatigue analysis of the main coolant pump has to follow the rules of the ASME Code as well as the German KTA-Regel. The main coolant pump is designed as an axial-flow pump. The housing and the housing cover are made of forged steel. At the inner surface, the housing has a 6 mm thick austenitic cladding. The pump operation details are given. The finite element model of the pump was constructed using the computer code PDA/PATRAN-G. Because of the symmetry, only one half of the housing has to be modelled. The stress analysis of the pump housing was performed using MSC/NASTRAN. Details are given. The results are discussed. (U.K.)

  13. En-mass coolant channel replacement in Indian PHWR

    International Nuclear Information System (INIS)

    Varghese, Joy P.; Verma, V.S.; Rajput, C.D.; Ramamurthy, K.

    2010-01-01

    This paper describes in brief the need for coolant channel replacement, challenges faced, methodology adopted to successfully overcome the shortcomings, improvements made in various areas that set a new standard and lifted the Indian image in the global nuclear industry. These changes have streamlined the EMCCR programme, which was followed on during the subsequent campaigns

  14. Transient performance of flow in PWR reactor circuits

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.; Carajilescov, P.

    1988-12-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  15. Transient performance of flow in circuits of PWR type reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  16. Coolant voiding analysis following SGTR for an HLMC reactor

    International Nuclear Information System (INIS)

    Farmer, M.T.; Spencer, B.W.; Sienicki, J.J.

    2000-01-01

    Concepts are under development at Argonne National Laboratory for a small, modular, proliferation-resistant nuclear power steam supply system. Of primary interest here is the simplified system design, featuring steam generators that are directly immersed in the lead-bismuth eutectic (LBE) coolant of the primary system. To support the safety case for this design approach, model development and analysis of transient coolant voiding during a postulated guillotine-type steam generator tube rupture event has been carried out. For the current design, the blowdown will occur from the steam generator shell into the ruptured 12.7-mm-inside-diameter tube through which the LBE coolant passes. The steam will expand biaxially in the tube, with a portion of the flow vented upward to eventually expand into the cover-gas region, while the balance of the flow is vented downward as a jet into the surrounding downward-flowing LBE. Coolant freezing is not an issue in this case because of high feedwater temperature in relation to the freezing point of the LBE. The specific objectives of the current work are to (a) determine the penetration behavior of the steam jet into the lower cold-leg region, (b) characterize the resultant void behavior in terms of coherent bubble versus breakup into a size distribution of small bubbles, and (c) characterize the motion of the bubbles with regard to rise to the cover-gas region (via the liner-to-coolant vessel gap) versus downward transport with the flowing LBE and subsequent upflow through the core to the cover-gas region

  17. Electrical Circuits and Water Analogies

    Science.gov (United States)

    Smith, Frederick A.; Wilson, Jerry D.

    1974-01-01

    Briefly describes water analogies for electrical circuits and presents plans for the construction of apparatus to demonstrate these analogies. Demonstrations include series circuits, parallel circuits, and capacitors. (GS)

  18. Electric circuits essentials

    CERN Document Server

    REA, Editors of

    2012-01-01

    REA's Essentials provide quick and easy access to critical information in a variety of different fields, ranging from the most basic to the most advanced. As its name implies, these concise, comprehensive study guides summarize the essentials of the field covered. Essentials are helpful when preparing for exams, doing homework and will remain a lasting reference source for students, teachers, and professionals. Electric Circuits I includes units, notation, resistive circuits, experimental laws, transient circuits, network theorems, techniques of circuit analysis, sinusoidal analysis, polyph

  19. Performance investigation of an automotive car radiator operated with nanofluid-based coolants (nanofluid as a coolant in a radiator)

    International Nuclear Information System (INIS)

    Leong, K.Y.; Saidur, R.; Kazi, S.N.; Mamun, A.H.

    2010-01-01

    Water and ethylene glycol as conventional coolants have been widely used in an automotive car radiator for many years. These heat transfer fluids offer low thermal conductivity. With the advancement of nanotechnology, the new generation of heat transfer fluids called, 'nanofluids' have been developed and researchers found that these fluids offer higher thermal conductivity compared to that of conventional coolants. This study focused on the application of ethylene glycol based copper nanofluids in an automotive cooling system. Relevant input data, nanofluid properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nanofluid-based coolants. It was observed that, overall heat transfer coefficient and heat transfer rate in engine cooling system increased with the usage of nanofluids (with ethylene glycol the basefluid) compared to ethylene glycol (i.e. basefluid) alone. It is observed that, about 3.8% of heat transfer enhancement could be achieved with the addition of 2% copper particles in a basefluid at the Reynolds number of 6000 and 5000 for air and coolant respectively. In addition, the reduction of air frontal area was estimated.

  20. Examination of a failed reactor coolant pump rotating assembly from Crystal River Unit 3

    International Nuclear Information System (INIS)

    Hayner, G.O.; Lubnow, T.; Clary, M.

    1990-01-01

    On January 18, 1989, the A reactor coolant pump rotating assembly at the Crystal River Unit 3 Nuclear Power Plant failed during operation. A rotating assembly from this pump had previously failed in 1986. The reactor coolant pump was fabricated by Byron Jackson Pump Division of Borg-Warner Ind. Products, Inc. from UNS S66286 superalloy (Alloy A286). A root cause failure analysis examination was performed on the pump shaft and other components. The failure analysis included shaft vibrational mode and stress analyses, pump clearance and alignment analyses, and detailed destructive examination of the shaft and hydrostatic bearing assemblies. Based on the detailed physical examination of the shaft it was concluded that cracks initiated in the pump shaft at two sites approximately 180 0 apart in a band of shallow, thermally induced fatigue cracks. The cracks initiated at the bottom edge of the motor end shrink fit pad under the shrink fit sleeve supporting the hydrostatic bearing journal. The band of thermally induced fatigue cracks was apparently caused by mixing of cold seal injection water and hot reactor coolant in gaps between the pump shaft and sleeve. The motor end shrink fit was apparently not effective in preventing introduction of the seal injection water to this area. Initial crack propagation occurred by fatigue due to lateral vibration; however, the majority of crack propagation occurred by abnormal torsional fatigue loading induced by contact and sticking between the rotating and stationary portions of the hydrostatic bearing. Final fracture of the shaft occurred by torsional overload. Metallurgical characteristics and mechanical properties of the shaft were within design specification and probably did not significantly influence the cracking process

  1. Optimization of Coolant Technique Conditions for Machining A319 Aluminium Alloy Using Response Surface Method (RSM)

    Science.gov (United States)

    Zainal Ariffin, S.; Razlan, A.; Ali, M. Mohd; Efendee, A. M.; Rahman, M. M.

    2018-03-01

    Background/Objectives: The paper discusses about the optimum cutting parameters with coolant techniques condition (1.0 mm nozzle orifice, wet and dry) to optimize surface roughness, temperature and tool wear in the machining process based on the selected setting parameters. The selected cutting parameters for this study were the cutting speed, feed rate, depth of cut and coolant techniques condition. Methods/Statistical Analysis Experiments were conducted and investigated based on Design of Experiment (DOE) with Response Surface Method. The research of the aggressive machining process on aluminum alloy (A319) for automotive applications is an effort to understand the machining concept, which widely used in a variety of manufacturing industries especially in the automotive industry. Findings: The results show that the dominant failure mode is the surface roughness, temperature and tool wear when using 1.0 mm nozzle orifice, increases during machining and also can be alternative minimize built up edge of the A319. The exploration for surface roughness, productivity and the optimization of cutting speed in the technical and commercial aspects of the manufacturing processes of A319 are discussed in automotive components industries for further work Applications/Improvements: The research result also beneficial in minimizing the costs incurred and improving productivity of manufacturing firms. According to the mathematical model and equations, generated by CCD based RSM, experiments were performed and cutting coolant condition technique using size nozzle can reduces tool wear, surface roughness and temperature was obtained. Results have been analyzed and optimization has been carried out for selecting cutting parameters, shows that the effectiveness and efficiency of the system can be identified and helps to solve potential problems.

  2. Operating experience of desalination unit coupled to primary coolant system of CIRUS

    International Nuclear Information System (INIS)

    Ranjan, Rakesh; Kumar, Surinder; Ramesh, N.; Mathur, D.; Sharma, R.C.

    2010-01-01

    Cirus is a 40 MWth, heavy water moderated, light water cooled, natural uranium fuelled research reactor located at Trombay, Mumbai. The reactor had been in operation since 1960 with an average availability factor of about 70% for about three decades. Reactor was shut down in October 1997 for refurbishment and was started back in 2003. Towards improving quality of life of public, since seventies, BARC has been engaged in R and D activities related to desalination. These efforts have led to the successful development of desalination technologies based on Multistage Flash (MSF) evaporation, Reverse Osmosis (RO) and Low Temperature Evaporation (LTE). Using low temperature vacuum evaporation technology, the centre had been studying the possibility of using the waste heat of nuclear reactors for seawater desalination. Based on this know-how, a 30 cubic meter per day (30 ton /day) pilot plant had been designed. With the aim of demonstrating the utilization of waste heat from research reactor, it was decided to integrate the unit to CIRUS during refurbishing outage of the reactor. The work involved design, installation and commissioning of set up to transfer heat from primary coolant of CIRUS to desalination unit through an intermediate demineralized water circuit. The unit was installed and commissioned during refurbishment outage of the reactor. The unit has been operated at its rated capacity and its performance has been satisfactory. The product water has been utilized to augment the demineralized water inventory of primary coolant system after processing it through ion exchanger resin. This paper highlights the experience gained during installation, commissioning and operation of the desalination unit which will be useful for application of the technology of utilization of waste heat. (author)

  3. Classical circuit theory

    CERN Document Server

    Wing, Omar

    2008-01-01

    Starting with the basic principles of circuits, this book derives their analytic properties in both the time and frequency domains. It develops an algorithmic method to design common and uncommon types of circuits, such as prototype filters, lumped delay lines, constant phase difference circuits, and delay equalizers.

  4. Piezoelectric drive circuit

    Science.gov (United States)

    Treu, C.A. Jr.

    1999-08-31

    A piezoelectric motor drive circuit is provided which utilizes the piezoelectric elements as oscillators and a Meacham half-bridge approach to develop feedback from the motor ground circuit to produce a signal to drive amplifiers to power the motor. The circuit automatically compensates for shifts in harmonic frequency of the piezoelectric elements due to pressure and temperature changes. 7 figs.

  5. Signal sampling circuit

    NARCIS (Netherlands)

    Louwsma, S.M.; Vertregt, Maarten

    2011-01-01

    A sampling circuit for sampling a signal is disclosed. The sampling circuit comprises a plurality of sampling channels adapted to sample the signal in time-multiplexed fashion, each sampling channel comprising a respective track-and-hold circuit connected to a respective analogue to digital

  6. Signal sampling circuit

    NARCIS (Netherlands)

    Louwsma, S.M.; Vertregt, Maarten

    2010-01-01

    A sampling circuit for sampling a signal is disclosed. The sampling circuit comprises a plurality of sampling channels adapted to sample the signal in time-multiplexed fashion, each sampling channel comprising a respective track-and-hold circuit connected to a respective analogue to digital

  7. Designing Novel Quaternary Quantum Reversible Subtractor Circuits

    Science.gov (United States)

    Haghparast, Majid; Monfared, Asma Taheri

    2018-01-01

    Reversible logic synthesis is an important area of current research because of its ability to reduce energy dissipation. In recent years, multiple valued logic has received great attention due to its ability to reduce the width of the reversible circuit which is a main requirement in quantum technology. Subtractor circuits are between major components used in quantum computers. In this paper, we will discuss the design of a quaternary quantum reversible half subtractor circuit using quaternary 1-qudit, 2-qudit Muthukrishnan-Stroud and 3-qudit controlled gates and a 2-qudit Generalized quaternary gate. Then a design of a quaternary quantum reversible full subtractor circuit based on the quaternary half subtractor will be presenting. The designs shall then be evaluated in terms of quantum cost, constant input, garbage output, and hardware complexity. The proposed quaternary quantum reversible circuits are the first attempt in the designing of the aforementioned subtractor.

  8. Radionuclide identification and dose rate projection for en-masse coolant channel replacement at Rajasthan Atomic Power Station, Unit-2

    International Nuclear Information System (INIS)

    Shuka, A.K.; Jain, S.K.; Nischal, Rajan; Agarwal, Sharad; Gandhimathinathan, S.; Sharma, R.M.

    1999-01-01

    The en-masse coolant channel replacement (EMCCR) in Rajasthan Atomic Power Station (RAPS) Unit-2 was carried out during 1996-97. This job which involved high intensity of nonuniformi radiation field, was done for the first time in India in any PHWR. Identification and estimation of gamma emitting radionuclides present in the highly active components played an important role in planning this job at various stages. Dose rates projected on the basis of samples collected from 0-11 coolant tube (RAPS-2) in 1992 were of great help in formulating the methodology and designing the facility at Solid Waste Management Plant for safe disposal of highly radioactive reactor components removed during the campaign. Based on dose rate projection, the initial dose rate was expected to reduce to a value somewhere between 2-15% after a cooling time of about 600 days. The projected values were later found in good agreement with the actual values obtained on analysis of chip samples collected from coolant tubes of channel K-8 and G-15. These data were of vital importance in job planning and exposure control during EMCCR campaign. (author)

  9. Cross-comparison of fast reactor concepts with various coolants

    International Nuclear Information System (INIS)

    Hejzlar, Pavel; Todreas, Neil E.; Shwageraus, Eugene; Nikiforova, Anna; Petroski, Robert; Driscoll, Michael J.

    2009-01-01

    Four fast reactor concepts using lead (LFR), liquid salt, NaCl-KCl-MgCl 2 (LSFR), sodium (SFR), and supercritical CO 2 (GFR) coolants are compared. Since economy of scale and power conversion system compactness are the same by virtue of the consistent 2400 MWt rating and use of the S-CO 2 power conversion system, the achievable plant thermal efficiency, core power density and core specific powers become the dominant factors. The potential to achieve the highest efficiency among the four reactor concepts can be ranked from highest to lowest as follows: (1) GFR, (2) LFR and LSFR, and (3) SFR. Both the lead- and salt-cooled designs achieve about 30% higher power density than the gas-cooled reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor. Fuel cycle costs are favored for the sodium reactor by virtue of its high specific power of 65 kW/kgHM compared to the lead, salt and gas reactor values of 45, 35, and 21 kW/kgHM, respectively. In terms of safety, all concepts can be designed to accommodate the unprotected limiting accidents through passive means in a self-controllable manner. However, it does not seem to be a preferable option for the GFR where the active or semi-passive approach will likely result in a more economic and reliable plant. Lead coolant with its superior neutronic characteristics and the smallest coolant temperature reactivity coefficient is easiest to design for self-controllability, while the LSFR requires special reactivity devices to overcome its large positive coolant temperature coefficient. The GFR required a special core design using BeO diluent and a supercritical CO 2 reflector to achieve negative coolant void worth-one of the conditions necessary for inherent shutdown following large LOCA. Protected accidents need to be given special attention in the LSFR and LFR due to the small margin to freezing of their coolants, and to a lesser extent in the SFR.

  10. Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no-load conditions

    Directory of Open Access Journals (Sweden)

    Ge Shao

    2015-06-01

    Full Text Available The advanced passive pressurized water reactor (PWR is being constructed in China and the passive residual heat removal (PRHR system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB event and inadvertent opening of a steam generator (SG safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

  11. Robotics in the nuclear environment-inspection and repairs inside the primary coolant system

    International Nuclear Information System (INIS)

    Guillet, J.; Marcel Tortolano

    2005-01-01

    The increase in the lifetime of the power plants and the ageing of materials require the intervention inside the components to carry out controls and possibly repairs in the event of discovered defects. Within this framework, EDF is investigating the feasibility of robotized repairs of the components and pipes of the main primary coolant system of a nuclear power plant. For several years, EDF R and D has engaged projects whose subject of study is the possibility of repairing components such as the main vessel; the pressurizer or the primary coolant pipes with the help of robots and dedicated tools. INTERVENTIONS INSIDE PRIMARY COOLANT PIPES: Studies undertaken by EDF highlighted that certain zones, particularly in pipe connections, can be affected by thermal fatigue which causes crackling defects or crackings. In anticipation of this phenomenon which would affect primary pipes and to avoid their replacements, EDF R and D has been studying the feasibility of examining and repairing these zones using robots. Robotized repair consists in introducing into the pipe while passing by the vessel, a 6 degrees of freedom manipulator mounted on a mobile carrier. This robot implements and carries out the trajectories of the different processes of repair: - Precise localization of the defects, - Elimination (possibly sampling) of the defects by machining, - Control that the defects were eliminated, - Weld metal buildup if the repair cavity is too deep, - Grinding followed by a new control of the surface. These studies and tests were conducted in the laboratory of EDF R and D in Chatou. The sequence of operations included machining by grinding and milling, profilometric control, dye penetrant testing, TIG welding and ultrasonic examinations. The results of the tests, executed on full scale models of components, are satisfactory and show the advantages of robotics compared with classical methods. ROBOTIZED INTERVENTIONS IN THE REACTOR VESSEL: Another difficult issue is the

  12. Feedback in analog circuits

    CERN Document Server

    Ochoa, Agustin

    2016-01-01

    This book describes a consistent and direct methodology to the analysis and design of analog circuits with particular application to circuits containing feedback. The analysis and design of circuits containing feedback is generally presented by either following a series of examples where each circuit is simplified through the use of insight or experience (someone else’s), or a complete nodal-matrix analysis generating lots of algebra. Neither of these approaches leads to gaining insight into the design process easily. The author develops a systematic approach to circuit analysis, the Driving Point Impedance and Signal Flow Graphs (DPI/SFG) method that does not require a-priori insight to the circuit being considered and results in factored analysis supporting the design function. This approach enables designers to account fully for loading and the bi-directional nature of elements both in the feedback path and in the amplifier itself, properties many times assumed negligible and ignored. Feedback circuits a...

  13. Liquid metal reactor development -Studies on safety measure of LMR coolant

    International Nuclear Information System (INIS)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author)

  14. Loss of coolant acident analyses on Osiris research reactor using the RELAP5 code

    International Nuclear Information System (INIS)

    Soares, Humberto Vitor; Costa, Antonella Lombardi; Lima, Claubia Pereira Bezerra; Veloso, Maria Auxiliadora Fortini

    2011-01-01

    RELAP5/MOD 3.3 code is widely used for thermal hydraulic studies of commercial nuclear power plants. However, several current investigations have shown that RELAP5 code can also be applied for thermal hydraulic analysis of nuclear research systems with good predictions. In this paper, a nodalization of the core and the most important components of the primary cooling system of the OSIRIS reactor developed for RELAP5 thermal hydraulic code are presented as well as results of steady state and transient simulations. OSIRIS has thermal power of 70 MW and it is an open pool type research reactor moderated and cooled by water. The OSIRIS reactor characteristics have been used as a base for the development of a model for the Multipurpose Brazilian Reactor (RMB). The aim of the present work is to investigate the behavior of the core during a loss of coolant accident and the possible damage of the fuel elements due an inadequate heat removal. Although the core coolant reached the saturation point due the large break, the fuel element conditions were out of the damage zone. (author)

  15. Analysis of events related to cracks and leaks in the reactor coolant pressure boundary

    Energy Technology Data Exchange (ETDEWEB)

    Ballesteros, Antonio, E-mail: Antonio.Ballesteros-Avila@ec.europa.eu [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Sanda, Radian; Peinador, Miguel; Zerger, Benoit [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Negri, Patrice [IRSN: Institut de Radioprotection et de Sûreté Nucléaire (France); Wenke, Rainer [GRS: Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH (Germany)

    2014-08-15

    Highlights: • The important role of Operating Experience Feedback is emphasised. • Events relating to cracks and leaks in the reactor coolant pressure boundary are analysed. • A methodology for event investigation is described. • Some illustrative results of the analysis of events for specific components are presented. - Abstract: The presence of cracks and leaks in the reactor coolant pressure boundary may jeopardise the safe operation of nuclear power plants. Analysis of cracks and leaks related events is an important task for the prevention of their recurrence, which should be performed in the context of activities on Operating Experience Feedback. In response to this concern, the EU Clearinghouse operated by the JRC-IET supports and develops technical and scientific work to disseminate the lessons learned from past operating experience. In particular, concerning cracks and leaks, the studies carried out in collaboration with IRSN and GRS have allowed to identify the most sensitive areas to degradation in the plant primary system and to elaborate recommendations for upgrading the maintenance, ageing management and inspection programmes. An overview of the methodology used in the analysis of cracks and leaks related events is presented in this paper, together with the relevant results obtained in the study.

  16. A guide to printed circuit board design

    CERN Document Server

    Hamilton, Charles

    1984-01-01

    A Guide to Printed Circuit Board Design discusses the basic design principles of printed circuit board (PCB). The book consists of nine chapters; each chapter provides both text discussion and illustration relevant to the topic being discussed. Chapter 1 talks about understanding the circuit diagram, and Chapter 2 covers how to compile component information file. Chapter 3 deals with the design layout, while Chapter 4 talks about preparing the master artworks. The book also covers generating computer aided design (CAD) master patterns, and then discusses how to prepare the production drawing a

  17. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  18. Nuclear reactor coolant pump impeller/shaft assembly

    International Nuclear Information System (INIS)

    Jenkins, L.S.

    1987-01-01

    A pump is described comprising: (a) a casing having an inlet and an outlet in fluid communication for circulating fluid coolant through the pump; (b) a shaft positioned in the casing; (c) an impeller nut connected to the shaft; (d) a lockbolt fixedly connecting the impeller nut relative to the shaft; (e) passageway means with first and second ends for directing fluid from the lockbolt to the shaft; (f) first conduit means formed in the impeller nut in fluid communication with the first end of the passageway means and the inlet of the casing; and (g) second conduit means formed in the impeller nut in fluid communication with the second end of the passageway means and the outlet of the casing. A portion of the fluid coolant circulating through the inlet of the casing is pumped through the first conduit means, through the passageway means, out the second conduit means and into the outlet of the casing

  19. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    1985-11-01

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  20. Peculiarities of evolution of shock waves generated by boiling coolant

    Science.gov (United States)

    Alekseev, M. V.; Vozhakov, I. S.; Lezhnin, S. I.; Pribaturin, N. A.

    2016-11-01

    Simulation of compression wave generation and evolution at the disk target was performed for the case of explosive-type boiling of coolant; the boiling is initiated by endwall rupture of a high-pressure pipeline. The calculations were performed for shock wave amplitude at different times and modes of pipe rupture. The simulated pressure of a target-reflected shock wave is different from the theoretical value for ideal gas; this discrepancy between simulation and theory becomes lower at higher distances of flow from the nozzle exit. Comparative simulation study was performed for flow of two-phase coolant with account for slip flow effect and for different sizes of droplets. Simulation gave the limiting droplet size when the single-velocity homogeneous flow model is valid, i.e., the slip flow effect is insignificant.

  1. Trends and experiences in reactor coolant pump motors

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    A review of the requirements and features of these motors is given as background along with a discussion of trends and experiences. Included are a discussion of thrust bearings and a review of safety related requirements and design features. Primary coolant pump motors are vertical induction motors for pumps that circulate huge quantities of water through the reactor core to carry the heat generated there to steam generator heat exchangers. 4 refs

  2. Method of injecting iron ion into reactor coolant

    International Nuclear Information System (INIS)

    Ito, Kazuyuki; Sawa, Toshio; Nishino, Yoshitaka; Adachi, Tetsuro; Osumi, Katsumi.

    1988-01-01

    Purpose: To form iron ions stably and inject them into nuclear reactor coolants with no substantial degradation of the severe water quality conditions for reactor coolants. Method: Iron ions are formed by spontaneous corrosion of iron type materials and electroconductivity is increased with the iron ions. Then, the liquids are introduced into an electrolysis vessel using iron type material as electrodes and, thereafter, incorporation of newly added ions other than the iron ions are prevented by supplying electric current. Further, by retaining the iron type material in the packing vessel by the magnetic force therein, only the iron ions are flow out substantially from the packing vessel while preventing the discharge of iron type materials per se or solid corrosion products and then introduced into the electrolysis vessel. Powdery or granular pure iron or carbon steel is used as the iron type material. Thus, iron ions and hydroxides thereof can be injected into coolants by using reactor water at low electroconductivity and incapable of electrolysis. (Kamimura, M.)

  3. Optimum parameters for treating coolant wastewater using PVDF-membrane

    Directory of Open Access Journals (Sweden)

    Yuliwati Erna

    2018-01-01

    Full Text Available Chemical oxygen demand (COD removal on ultrafiltration membrane were studied from treating of coolant wastewater. The membranes were fabricated via the dry-jet wet spinning method. The experiment was conducted using synthetic coolant wastewater as effluent, and an experimental set-up consist of membrane reservoir was used throughout the investigation. Deposition and accumulation of suspended solids on the membrane surface during filtration were prohibited with continuous aeration. Membrane performance was measured as well as flux and COD removal efficiency. The RSM was performed to investigate the optimised process conditions for treating coolant wastewater. Results using RSM-approximation have demonstrated the improvement in COD removal of 99.63%. This improvement is achieved by optimised process conditions of mixed liquor suspended solids (MLSS concentration, air bubble flow rate (ABFR, hydraulic retention time (HRT and pH at 6.00 g/L, 3.05 ml/min, 280.50 min, and 6.50 respectively.

  4. On-line real time gamma analysis of primary coolant

    International Nuclear Information System (INIS)

    Kalechstein, W.; Kupca, S.; Lipsett, J.J.

    1985-10-01

    The evolution of failed fuel monitoring at CANDU power stations is briefly summarized and the design of the latest system for failed fuel detection at a multi-unit power station is described. At each reactor, the system employs a germanium spectrometer combined with a novel spectrum analyzer that simultaneously accumulates the gamma-ray spectrum of the coolant and provides the control room with the concentration of radioisotope activity in the coolant for the gaseous fission products Xe-133, Xe-135, Kr-88 and I-131 in real time and with statistical precision independent of count rate. A gross gamma monitor is included to provide independent information on the level of radioactivity in the coolant and extend the measurement range at very high count rates. A central computer system archives spectra received from all four spectrum analyzers and provides both the activity concentrations and the release rates of specified isotopes. Compared with previous systems the current design offers improvements in that the activity concentrations are updated much more frequently, improved tools are provided for long term surveillance of the heat transport system and the monitor is more reliable and less costly

  5. Coolant Design System for Liquid Propellant Aerospike Engines

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    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.

  7. Analysis of actual status of works on technology of heavy liquid metal coolants

    International Nuclear Information System (INIS)

    Martynov, P.N.; Askhadullin, R.Sh.; Orlov, Yu.I.; Storozhenko, A.N.

    2014-01-01

    Principle duties in heavy liquid metal coolant technology (HLMC) are provision of the purity of coolant and surfaces of circulation loop for maintenance of design thermohydraulic characteristics, prevention of structural materials corrosion and erosion during long service life and present-day safety precautions on different stages of reactor facility operation. For this reason, current HLMC (Pb-Bi, Pb) technology must include coolant pre-operation and charging; monitoring and regulating of coolant oxygen potential; hydrogen purification of coolant and surfaces of circulation loop from lead oxides-based slags; coolant filtration; reactor cover gas purification from coolant aerosols. The current topical problem is personnel training on the questions of HLMC technology [ru

  8. A comparison of core perturbation by coolant loss between sodium and lead-bismuth cooled reactor

    International Nuclear Information System (INIS)

    Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok

    2003-01-01

    This study performs a comparative analysis of the core perturbation caused by coolant loss between sodium and lead-bismuth eutectic. Considering the Zr-based and the U-based fuel in a 1,000MWth class reactor for TRU incineration, we investigate which coolant shows better performance for negative coolant loss reactivity in each case of fuel type. The calculation results show that in the case of U-based fuel, sodium gives rise to more positive coolant loss reactivity than lead-bismuth. However, when the Zr-based (U-free) fuel is considered, sodium offers negative coolant loss reactivity, whereas lead-bismuth makes the coolant loss reactivity positive. It is recommended to employ sodium coolant for the fertile-free fueled core and lead-bismuth for the core with fertile nuclides

  9. Monitoring Sodium Circuits and ACSR cables using Fiber Optic Sensors

    International Nuclear Information System (INIS)

    Kasinathan, M.; Sosamma, S.; Babu-Rao, C.; Kumar, Anish; Purna-Chandra-Rao, B.; Murali, N; Jayakumar, T.

    2013-06-01

    Raman Distributed Temperature Sensors (RDTS) are attractive for the monitoring of coolant loop systems in nuclear power plants and monitoring of overhead power transmission lines. This paper discusses deployment of RDTS on double walled pipelines of primary sodium circuits in Fast Breeder Reactors (FBR). It is demonstrated as a proof-of-concept on a test loop with water as the leaking medium. Path delay multiplexing is adopted to improve the spatial resolution from 1.02 m to 0.5 m. A second application focuses on the influence of environmental factors on the detectability of defects in the ACSR cables using RDTS. (authors)

  10. BWR loss-of-coolant accident tests at ROSA-III with high temperature emergency core coolant injection

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Kukita, Yutaka; Tasaka, Kanji

    1988-01-01

    The effects of emergency core coolant (ECC) temperature on the performance of the emergency core cooling system (ECCS) during a loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) were investigated experimentally using the Rig of Safety Assessment (ROSA)-3 integral test facility. The ECC temperature had no direct influence on the ECCS core cooling performance, since ECC became nearly saturated before reaching the core irrespective of the initial temperature, however, had indirect effects by changing the vessel pressure response. The ECCS injection timings and flow rates, and the core inlet flooding behavior were affected. The measured peak cladding temperature (PCT) was not affected by the ECC temperature for both large (200%) and small (5%) break tests. (author)

  11. Digital signal processing in power electronics control circuits

    CERN Document Server

    Sozanski, Krzysztof

    2013-01-01

    Many digital control circuits in current literature are described using analog transmittance. This may not always be acceptable, especially if the sampling frequency and power transistor switching frequencies are close to the band of interest. Therefore, a digital circuit is considered as a digital controller rather than an analog circuit. This helps to avoid errors and instability in high frequency components. Digital Signal Processing in Power Electronics Control Circuits covers problems concerning the design and realization of digital control algorithms for power electronics circuits using

  12. Membrane technology for treating of waste nanofluids coolant: A review

    Science.gov (United States)

    Mohruni, Amrifan Saladin; Yuliwati, Erna; Sharif, Safian; Ismail, Ahmad Fauzi

    2017-09-01

    The treatment of cutting fluids wastes concerns a big number of industries, especially from the machining operations to foster environmental sustainability. Discharging cutting fluids, waste through separation technique could protect the environment and also human health in general. Several methods for the separation emulsified oils or oily wastewater have been proposed as three common methods, namely chemical, physicochemical and mechanical and membrane technology application. Membranes are used into separate and concentrate the pollutants in oily wastewater through its perm-selectivity. Meanwhile, the desire to compensate for the shortcomings of the cutting fluid media in a metal cutting operation led to introduce the using of nanofluids (NFs) in the minimum quantity lubricant (MQL) technique. NFs are prepared based on nanofluids technology by dispersing nanoparticles (NPs) in liquids. These fluids have potentially played to enhance the performance of traditional heat transfer fluids. Few researchers have studied investigation of the physical-chemical, thermo-physical and heat transfer characteristics of NFs for heat transfer applications. The use of minimum quantity lubrication (MQL) technique by NFs application is developed in many metal cutting operations. MQL did not only serve as a better alternative to flood cooling during machining operation and also increases better-finished surface, reduces impact loads on the environment and fosters environmental sustainability. Waste coolant filtration from cutting tools using membrane was treated by the pretreated process, coagulation technique and membrane filtration. Nanomaterials are also applied to modify the membrane structure and morphology. Polyvinylidene fluoride (PVDF) is the better choice in coolant wastewater treatment due to its hydrophobicity. Using of polyamide nanofiltration membranes BM-20D and UF-PS-100-100, 000, it resulted in the increase of permeability of waste coolant filtration. Titanium dioxide

  13. Electric circuits and signals

    CERN Document Server

    Sabah, Nassir H

    2007-01-01

    Circuit Variables and Elements Overview Learning Objectives Electric Current Voltage Electric Power and Energy Assigned Positive Directions Active and Passive Circuit Elements Voltage and Current Sources The Resistor The Capacitor The Inductor Concluding Remarks Summary of Main Concepts and Results Learning Outcomes Supplementary Topics on CD Problems and Exercises Basic Circuit Connections and Laws Overview Learning Objectives Circuit Terminology Kirchhoff's Laws Voltage Division and Series Connection of Resistors Current Division and Parallel Connection of Resistors D-Y Transformation Source Equivalence and Transformation Reduced-Voltage Supply Summary of Main Concepts and Results Learning Outcomes Supplementary Topics and Examples on CD Problems and Exercises Basic Analysis of Resistive Circuits Overview Learning Objectives Number of Independent Circuit Equations Node-Voltage Analysis Special Considerations in Node-Voltage Analysis Mesh-Current Analysis Special Conside...

  14. Analog circuit design

    CERN Document Server

    Dobkin, Bob

    2012-01-01

    Analog circuit and system design today is more essential than ever before. With the growth of digital systems, wireless communications, complex industrial and automotive systems, designers are being challenged to develop sophisticated analog solutions. This comprehensive source book of circuit design solutions aids engineers with elegant and practical design techniques that focus on common analog challenges. The book's in-depth application examples provide insight into circuit design and application solutions that you can apply in today's demanding designs. <

  15. Regenerative feedback resonant circuit

    Science.gov (United States)

    Jones, A. Mark; Kelly, James F.; McCloy, John S.; McMakin, Douglas L.

    2014-09-02

    A regenerative feedback resonant circuit for measuring a transient response in a loop is disclosed. The circuit includes an amplifier for generating a signal in the loop. The circuit further includes a resonator having a resonant cavity and a material located within the cavity. The signal sent into the resonator produces a resonant frequency. A variation of the resonant frequency due to perturbations in electromagnetic properties of the material is measured.

  16. Advanced Microwave Circuits and Systems

    DEFF Research Database (Denmark)

    such as voltage-controlled oscillators and electron devices for millimeter wave and submillimeter wave applications. This part also covers studies of integrated buffer circuits. Passive components are indispensable elements of any electronic system. The increasing demands to miniaturization and cost effectiveness...... push currently available technologies to the limits. Some considerations to meet the growing requirements are provided in the fifth part of this book. The following part deals with circuits based on LTCC and MEMS technologies. The book concludes with chapters considering application of microwaves...... in measurement and sensing systems. This includes topics related to six-port reflectometers, remote network analysis, inverse scattering for microwave imaging systems, spectroscopy for medical applications and interaction with transponders in medical sensors....

  17. Timergenerator circuits manual

    CERN Document Server

    Marston, R M

    2013-01-01

    Timer/Generator Circuits Manual is an 11-chapter text that deals mainly with waveform generator techniques and circuits. Each chapter starts with an explanation of the basic principles of its subject followed by a wide range of practical circuit designs. This work presents a total of over 300 practical circuits, diagrams, and tables.Chapter 1 outlines the basic principles and the different types of generator. Chapters 2 to 9 deal with a specific type of waveform generator, including sine, square, triangular, sawtooth, and special waveform generators pulse. These chapters also include pulse gen

  18. Circuits and filters handbook

    CERN Document Server

    Chen, Wai-Kai

    2003-01-01

    A bestseller in its first edition, The Circuits and Filters Handbook has been thoroughly updated to provide the most current, most comprehensive information available in both the classical and emerging fields of circuits and filters, both analog and digital. This edition contains 29 new chapters, with significant additions in the areas of computer-aided design, circuit simulation, VLSI circuits, design automation, and active and digital filters. It will undoubtedly take its place as the engineer's first choice in looking for solutions to problems encountered in the design, analysis, and behavi

  19. CMOS circuits manual

    CERN Document Server

    Marston, R M

    1995-01-01

    CMOS Circuits Manual is a user's guide for CMOS. The book emphasizes the practical aspects of CMOS and provides circuits, tables, and graphs to further relate the fundamentals with the applications. The text first discusses the basic principles and characteristics of the CMOS devices. The succeeding chapters detail the types of CMOS IC, including simple inverter, gate and logic ICs and circuits, and complex counters and decoders. The last chapter presents a miscellaneous collection of two dozen useful CMOS circuits. The book will be useful to researchers and professionals who employ CMOS circu

  20. MOS integrated circuit design

    CERN Document Server

    Wolfendale, E

    2013-01-01

    MOS Integral Circuit Design aims to help in the design of integrated circuits, especially large-scale ones, using MOS Technology through teaching of techniques, practical applications, and examples. The book covers topics such as design equation and process parameters; MOS static and dynamic circuits; logic design techniques, system partitioning, and layout techniques. Also featured are computer aids such as logic simulation and mask layout, as well as examples on simple MOS design. The text is recommended for electrical engineers who would like to know how to use MOS for integral circuit desi

  1. Security electronics circuits manual

    CERN Document Server

    MARSTON, R M

    1998-01-01

    Security Electronics Circuits Manual is an invaluable guide for engineers and technicians in the security industry. It will also prove to be a useful guide for students and experimenters, as well as providing experienced amateurs and DIY enthusiasts with numerous ideas to protect their homes, businesses and properties.As with all Ray Marston's Circuits Manuals, the style is easy-to-read and non-mathematical, with the emphasis firmly on practical applications, circuits and design ideas. The ICs and other devices used in the practical circuits are modestly priced and readily available ty

  2. MATLAB/Simulink Framework for Modeling Complex Coolant Flow Configurations of Advanced Automotive Thermal Management Systems: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Titov, Eugene; Lustbader, Jason; Leighton, Daniel; Kiss, Tibor

    2016-03-22

    The National Renewable Energy Laboratory's (NREL's) CoolSim MATLAB/Simulink modeling framework was extended by including a newly developed coolant loop solution method aimed at reducing the simulation effort for arbitrarily complex thermal management systems. The new approach does not require the user to identify specific coolant loops and their flow. The user only needs to connect the fluid network elements in a manner consistent with the desired schematic. Using the new solution method, a model of NREL's advanced combined coolant loop system for electric vehicles was created that reflected the test system architecture. This system was built using components provided by the MAHLE Group and included both air conditioning and heat pump modes. Validation with test bench data and verification with the previous solution method were performed for 10 operating points spanning a range of ambient temperatures between -2 degrees C and 43 degrees C. The largest root mean square difference between pressure, temperature, energy and mass flow rate data and simulation results was less than 7%.

  3. Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

    Directory of Open Access Journals (Sweden)

    Lefèvre Grégory

    2012-01-01

    Full Text Available In the primary circuit of pressurized water reactors (PWR, the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspected to be the cause. As better understanding of the adhesion mechanism is the key factor in the prevention of fouling and particle removal, an experimental study was carried out using a laboratory set-up. With model materials, hematite and sintered alumina, the adhesion rate and surface potentials of the interacting solids were measured under different chemical conditions (solution pH and composition in analogy with the PWR ones. The obtained results were in good agreement with the DLVO (Derjaguin-Landau-Verwey- Overbeek theory and used as such to interpret this industrial phenomenon.

  4. Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Kanzleiter, T.F.

    1976-01-01

    For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods. An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m 3 and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1 : 64 compared to the 1200 MW PWR plant Biblis A. (Auth.)

  5. Instrument for continuous supervision of the radioactivity of CO2 coolant in piles - DCCA -CO2 (1960)

    International Nuclear Information System (INIS)

    Fitoussi, L.

    1960-01-01

    This paper describes an apparatus for continuous measurement of CO 2 activity, which can be used on piles cooled by circulation of gas. The first part is devoted mainly to describing the apparatus used and the character of the radioactivity and thermodynamic measurements carried out, and giving the general characteristics of the gas circuit required if the instrument is to be suitably gas-tight. In the second part theoretical calculations are given, particularly on the determination of the ionisation current in an ionisation chamber with circulating gas. Several parameters enter into this determination, such as the mean path of β particles in the ionisation chamber, the linear number of ion pairs formed in the gas by these β particles as a function of their energy, the temperature and pressure of the gas in the ionisation chamber. This part also evaluates the sensitivity areas of the apparatus for measuring the concentrations of radioactive gases such as argon-41 and fission gases from uranium-235 in the CO 2 coolant. In the last part are described the results of measurements performed with such an apparatus on the pile EL2, the special investigations carried out on the CO 2 coolant of this pile, and the information gained during normal operation and during accidents. The DCCA - CO 2 which has just been put in operation at G2 is briefly presented. In the conclusion the possibilities offered by this apparatus are underlined. (author) [fr

  6. Qualification of coolants and cooling pipes for future high-energy-particle detectors

    Science.gov (United States)

    Ilie, Sorin; Tavlet, Marc

    2001-12-01

    In the next generation of high-energy-particle detectors to be installed at the Large Hadron Collider (LHC) at CERN, materials and components will be exposed to a significant level of ionising radiation. Silicon detectors and related electronics will have to be cooled down to -20 °C and therefore appropriate cooling fluids and cooling pipes have to be selected. Analytical methods such as UV-visible and FT-IR spectrometries, electronic microscopy and gas chromatography were used to characterise the radiation-induced effects on some organic coolants irradiated with both gamma and neutron fields. Some impurities were identified as a major source for radio-induced polymerisation and also for hydrofluoric acid (HF) evolution. Mechanical tests were performed to assess the operability of the rubber hoses and plastic pipes. Possible synergistic effects between the pipe material and the environment had to be considered.

  7. Preliminary evaluation of the stress analysis reports for Angra I reactor coolant loop - part 1

    International Nuclear Information System (INIS)

    Ribeiro, S.V.G.; Andrade, J.E.L. de

    1980-03-01

    A methodology that will allow CNEN to approve the stress analysis reports of the components of the Brazilian nuclear power plants, was developed. The reactor coolant loop (RCL)of Angra I was checkd. This is the first part of the complete report and consists of the approval of the design documents, the approval of the equipment support models and the aproval of the steam generator dynamic model. The second part of this work is under way now and should contain the approval of the RCL stress and fatigue analysis according to ASME code section III. As shown in section 7 it appears necessary additional information from Westinghouse about the design of the RCL. (Author) [pt

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

    International Nuclear Information System (INIS)

    Hunsbedt, A.N.

    1988-01-01

    This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

  9. 'Closed Circuit' Anaesthesia

    African Journals Online (AJOL)

    The advantages of using rebreathing circuits in anaes- thesia are discussed and the principles for their correct employment are outlined. Practical methods are described. By using closed circuit equipment in the manner described, the initial cost of the apparatus could be recouped within one year, because of the saving in ...

  10. Offset cancelling circuit

    NARCIS (Netherlands)

    Wiegerink, Remco J.; Seevinck, Evert; de Jager, Wim

    1989-01-01

    A monolithic offset cancelling circuit to reduce the offset voltage at an integrated audio-amplifier output is described. This offset voltage is detected using a low-pass filter with a very large time constant for which only one small on-chip capacitor is needed. The circuit was realized with a

  11. Non-destructive testing and periodic inspection of the primary circuits of reactors

    International Nuclear Information System (INIS)

    Prot, A.C.; Saglio, R.

    1975-01-01

    A knowledge of failures in the different components of a nuclear power plant is important for ensuring the reliability of the plant. The primary coolant circuit is particularly important from the safety point of view, which is why so many countries are working on regulations concerning periodic inspection of this system. The authors start by describing the parallel development of such regulations and of the non-destructive testing techniques designed to meet their requirements. After stating these requirements, they discuss the progress made in France during the past few years in different areas: ultrasonic non-destructive testing, especially through the stainless steel lining of the reactor vessel; extension of the results to external testing (advantage inherent in the method), to the testing of the mixed welds of safe-ends and of austenitic stainless steel welds, and to the testing of the reactor cover bolts; improvement of an acoustic holography process (which is compared with the ultrasonic method); improvement of testing methods based on Foucault currents (especially for the inspection of steam generators); and improvement of detection and location by acoustic emission during hydraulic trials and continuous operation. In conclusion, the authors suggest the main lines of work that should be followed to achieve better non-destructive testing during the construction, entry into service and operation of nuclear power plants

  12. Electronic components and systems

    CERN Document Server

    Dennis, W H

    2013-01-01

    Electronic Components and Systems focuses on the principles and processes in the field of electronics and the integrated circuit. Covered in the book are basic aspects and physical fundamentals; different types of materials involved in the field; and passive and active electronic components such as capacitors, inductors, diodes, and transistors. Also covered in the book are topics such as the fabrication of semiconductors and integrated circuits; analog circuitry; digital logic technology; and microprocessors. The monograph is recommended for beginning electrical engineers who would like to kn

  13. CMOS analog circuit design

    CERN Document Server

    Allen, Phillip E

    1987-01-01

    This text presents the principles and techniques for designing analog circuits to be implemented in a CMOS technology. The level is appropriate for seniors and graduate students familiar with basic electronics, including biasing, modeling, circuit analysis, and some familiarity with frequency response. Students learn the methodology of analog integrated circuit design through a hierarchically-oriented approach to the subject that provides thorough background and practical guidance for designing CMOS analog circuits, including modeling, simulation, and testing. The authors' vast industrial experience and knowledge is reflected in the circuits, techniques, and principles presented. They even identify the many common pitfalls that lie in the path of the beginning designer--expert advice from veteran designers. The text mixes the academic and practical viewpoints in a treatment that is neither superficial nor overly detailed, providing the perfect balance.

  14. PUMP: analog-hybrid reactor coolant hydraulic transient model

    International Nuclear Information System (INIS)

    Grandia, M.R.

    1976-03-01

    The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  16. Use of flow models to analyse loss of coolant accidents

    International Nuclear Information System (INIS)

    Pinet, Bernard

    1978-01-01

    This article summarises current work on developing the use of flow models to analyse loss-of-coolant accident in pressurized-water plants. This work is being done jointly, in the context of the LOCA Technical Committee, by the CEA, EDF and FRAMATOME. The construction of the flow model is very closely based on some theoretical studies of the two-fluid model. The laws of transfer at the interface and at the wall are tested experimentally. The representativity of the model then has to be checked in experiments involving several elementary physical phenomena [fr

  17. COPDIRC - calculation of particle deposition in reactor coolants

    International Nuclear Information System (INIS)

    Reeks, M.W.

    1982-06-01

    A description is given of a computer code COPDIRC intended for the calculation of the deposition of particulate onto smooth perfectly sticky surfaces in a gas cooled reactor coolant. The deposition is assumed to be limited by transport in the boundary layer adjacent to the depositing surface. This implies that the deposition velocity normalised with respect to the local friction velocity, is an almost universal function of the normalised particle relaxation time. Deposition is assumed similar to deposition in an equivalent smooth perfectly absorbing pipe. The deposition is calculated using 2 models. (author)

  18. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    Ramirez G, R.

    1975-01-01

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  19. DETERMINATION OF THE 129I IN PRIMARY COOLANT OF PWR

    Directory of Open Access Journals (Sweden)

    KE CHON CHOI

    2013-02-01

    In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of 129I was examined, as was the effect of 3H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous 3H presence was found with activity concentrations of 3H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 μg/mL.

  20. Signal-to-noise ratio determination circuit

    Science.gov (United States)

    Deerkoski, L. F. (Inventor)

    1973-01-01

    A signal-to-noise ratio (SNR) determination of an input is described, having signal components within a given frequency range and noise components, without actual measurement of the noise components. Bandpass limiter having a constant signal plus noise output level is connected to the output of the first filter, the signal-to-noise ratio of the input to the bandpass limiter being linearly related to the dbm level of signal components at the output. Calibration is connected to the bandpass limiter and is responsive to the signal components at the output to derive the SNR of the input to the determination circuit. The SNR determination circuit is disclosed for use in a diversity receiver having a plurality of input channels.

  1. Experimental simulation of asymmetric heat up of coolant channel under small break LOCA condition for PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, Ashwini K., E-mail: ashwinikumaryadav@gmail.com [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kumar, Ravi, E-mail: ravikfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Chatterjee, B., E-mail: barun@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Gupta, Akhilesh, E-mail: akhilfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667 (India); Mukhopadhyay, D., E-mail: dmukho@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2013-02-15

    Highlights: ► Circumferential temperature gradient of PT for asymmetric heat-up was 440 °C. ► At 2 MPa ballooning initiated at 450 °C and with strain rate of 0.0277%/s. ► At 4 MPa ballooning initiated at 390 °C and with strain rate of 0.0305%/s. ► At 4 MPa, PT ruptured under uneven strain and steep temperature gradient. ► Integrity of PT depends on internal pressure and magnitude of decay power. -- Abstract: During postulated small break loss of coolant accident (SBLOCA) for Pressurised Heavy Water Reactors (PHWRs) as well as for postulated SBLOCA coincident with loss of ECCS, a stratified flow condition can arise in the coolant channels as the gravitational force dominates over the low inertial flow arising from small break flow. A Station Blackout condition without operator intervention can also lead to stratified flow condition during a slow channel boil-off condition. For all these conditions the pressure remains high and under stratified flow condition, the horizontal fuel bundles experience different heat transfer environments with respect to the stratified flow level. This causes the bundle upper portion to get heated up higher as compared to the submerged portion. This kind of asymmetrical heating of the bundle is having a direct bearing on the circumferential temperature gradient of pressure tube (PT) component of the coolant channel. The integrity of the PT is important under normal conditions as well as at different accident loading conditions as this component houses the fuel bundles and serves as a coolant pressure boundary of the reactors. An assessment of PT is required with respect to different accident loading conditions. The present investigation aims to study thermo-mechanical behaviour of PT (Zr, 2.5 wt% Nb) under a stratified flow condition under different internal pressures. The component is subjected to an asymmetrical heat-up conditions as expected during the said situation under different pressure conditions which varies from 2

  2. Analog Circuit Design Automation Tool Using Genetic Algorithms

    Directory of Open Access Journals (Sweden)

    Y. M. A. Khalifa

    2012-07-01

    Full Text Available An investigation into the application of Genetic Algorithms (GA for the design of electronic analog circuits is presented in this paper. In this paper an investigation of the use of genetic algorithms into the problem of analog circuits design is presented. In a single design stage, circuits are produced that satisfy specific frequency response specifications using circuit structures that are unrestricted and with component values that are chosen from a set of preferred values. The extra degrees of freedom resulting from unbounded circuit structures create a huge search space. It is shown in this paper that Genetic Algorithms can be successfully used to search this space. The application chosen is a LC all pass ladder filter circuit design.Key Words: Computer-Aided Design, Analog Circuits, Artificial Intelligence.

  3. Design technology development of the main coolant pump for an integral reactor

    International Nuclear Information System (INIS)

    Park, J. S.; Lee, J. S.; Kim, M. H.; Kim, D. W.; Kim, J. I.

    2004-01-01

    All of the reactor coolant pump currently used in commercial nuclear power plant were imported from foreign country. Now, the developing program of design technology for the reactor coolant pump will be started in a few future by domestic researchers. At this stage, the design technology of the main coolant pump for an integral reactor is developed based on the regulation of domestic nuclear power plant facilities. The main coolant pump is a canned motor axial pump, which accommodates all constraints required from the integral reactor system. The main coolant pump does not have mechanical seal device because the rotor of motor and the shaft of impeller are the same one. There is no flywheel on the rotating shaft of main coolant pump so that the coastdown duration time is short when the electricity supply is cut off

  4. Experimental Investigation of Coolant Boiling in a Half-Heated Circular Tube - Final CRADA Report

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Wenhua [Argonne National Lab. (ANL), Argonne, IL (United States); Singh, Dileep [Argonne National Lab. (ANL), Argonne, IL (United States); France, David M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-11-01

    Coolant subcooled boiling in the cylinder head regions of heavy-duty vehicle engines is unavoidable at high thermal loads due to high metal temperatures. However, theoretical, numerical, and experimental studies of coolant subcooled flow boiling under these specific application conditions are generally lacking in the engineering literature. The objective of this project was to provide such much-needed information, including the coolant subcooled flow boiling characteristics and the corresponding heat transfer coefficients, through experimental investigations.

  5. Preliminary design of reactor coolant pump canned motor for AC600

    International Nuclear Information System (INIS)

    Deng Shaowen

    1998-01-01

    The reactor coolant pump canned motor of AC600 PWR is the kind of shielded motors with high moment of inertia, high reliability, high efficiency and nice starting performance. The author briefly presents the main feature, design criterion and technical requirements, preliminary design, computation results and analysis of performance of AC600 reactor coolant pump canned motor, and proposes some problems to be solved for study and design of AC600 reactor coolant pump canned motor

  6. Pengaruh Penambahan Zat Aditif (Calcium Hypochlorite) terhadap Coolant Properties (Viscosity) Minyak Sawit

    OpenAIRE

    Skp, Buma Gempa; Masnur, Dedy; HS, Irdoni

    2014-01-01

    Cutting fluid (coolant) are required in the machining processes. Most of cutting fluid are from mineral oil, and they are as unrenewable resource, therefore look for alternative resources ofmineral oil, which derived from vegetables oil. Palm oil have potency based of cutting fluid, however of coolant properties increased additive. This research want investigate effect ofadditive on phiysical properties of coolant. The method on mixture combining palm oil and additive (calcium hypochlorite) a...

  7. Approximate circuits for increased reliability

    Science.gov (United States)

    Hamlet, Jason R.; Mayo, Jackson R.

    2015-08-18

    Embodiments of the invention describe a Boolean circuit having a voter circuit and a plurality of approximate circuits each based, at least in part, on a reference circuit. The approximate circuits are each to generate one or more output signals based on values of received input signals. The voter circuit is to receive the one or more output signals generated by each of the approximate circuits, and is to output one or more signals corresponding to a majority value of the received signals. At least some of the approximate circuits are to generate an output value different than the reference circuit for one or more input signal values; however, for each possible input signal value, the majority values of the one or more output signals generated by the approximate circuits and received by the voter circuit correspond to output signal result values of the reference circuit.

  8. The electronics companion devices and circuits for physicists and engineers

    CERN Document Server

    Fischer-Cripps, Anthony C

    2014-01-01

    Updated and expanded with new topics, The Electronics Companion: Devices and Circuits for Physicists and Engineers, 2nd Edition presents a full course in introductory electronics using a unique and educational presentation technique that is the signature style of the author’s companion books. This concise yet detailed book covers introductory electrical principles (DC and AC circuits), the physics of electronics components, circuits involving diodes and transistors, transistors amplifiers, filtering, operational amplifiers, digital electronics, transformers, instrumentation, and power supplies.

  9. Sensor readout detector circuit

    Science.gov (United States)

    Chu, D.D.; Thelen, D.C. Jr.

    1998-08-11

    A sensor readout detector circuit is disclosed that is capable of detecting sensor signals down to a few nanoamperes or less in a high (microampere) background noise level. The circuit operates at a very low standby power level and is triggerable by a sensor event signal that is above a predetermined threshold level. A plurality of sensor readout detector circuits can be formed on a substrate as an integrated circuit (IC). These circuits can operate to process data from an array of sensors in parallel, with only data from active sensors being processed for digitization and analysis. This allows the IC to operate at a low power level with a high data throughput for the active sensors. The circuit may be used with many different types of sensors, including photodetectors, capacitance sensors, chemically-sensitive sensors or combinations thereof to provide a capability for recording transient events or for recording data for a predetermined period of time following an event trigger. The sensor readout detector circuit has applications for portable or satellite-based sensor systems. 6 figs.

  10. Loss-of-Coolant and Loss-of-Flow Accidents in the SEAFP first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1995-01-01

    This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

  11. Loss of Coolant Accidents (LOCA): Study of CAREM Reactor Response

    International Nuclear Information System (INIS)

    Gonzalez, Jose; Gimenez, Marcelo

    2000-01-01

    We analyzed the neutronic and thermohydraulic response of CAREM25 reactor and the safety systems involved in a Loss Of Coolant Accident (LOCA).This parametric analysis considers several break diameters (1/2inch, 3/4inch, 1inch, 1.1/2inch and 2inches) in the vapor zone of the Reactor Pressure Vessel.For each accidental sequence, the successful operation of the following safety systems is modeled: Second Safety System (SSS), Residual Heat Removal System (RHRS) and Safety Injection System (SIS). Availability of only one module is postulated for each system.On the other hand, the unsuccessful operation of all safety systems is postulated for each accidental sequence.In both cases the First Shutdown System (FSS) actuates, and the loss of Steam Generator secondary flow and Chemical and Control of Volume System (CCVS) unavailability are postulated.Maximum loss of coolant flow, reactor power and time for safety systems operation are analyzed, as well as its set point parameters.We verified that safety systems are dimensioned to satisfy the 48 hours cooling criteria

  12. Station blackout with reactor coolant pump seal leakage

    International Nuclear Information System (INIS)

    Evinay, A.

    1993-01-01

    The U.S. Nuclear Regulatory Commission (NRC) amended its regulations in 10CFR50 with the addition of a new section, 50.63, open-quotes Loss of All Alternating Current Power.close quotes The objective of these requirements is to ensure that all nuclear plants have the capability to withstand a station blackout (SBO) and maintain adequate reactor core cooling and containment integrity for a specified period of time. The NRC also issued Regulatory Guide (RG) 1.155, open-quotes Station Blackout,close quotes to provide guidance for meeting the requirements of 10CFR50.63. Concurrent with RG-1.155, the Nuclear Utility Management and Resources Council (NUMARC) has developed NUMARC 87-00 to address SBO-coping duration and capabilities at light water reactors. Licensees are required to submit a topical report based on NUMARC 87-00 guidelines, to demonstrate compliance with the SBO rule. One of the key compliance criteria is the ability of the plant to maintain adequate reactor coolant system (RCS) inventory to ensure core cooling for the required coping duration, assuming a leak rate of 25 gal/min per reactor coolant pump (RCP) seal in addition to technical specification (TS) leak rate

  13. Coolant clean-up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

    1987-01-01

    Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

  14. The effect of coolant boiling on the molten metal pool heat transfer with local solidification

    International Nuclear Information System (INIS)

    Cho, Jae Seon; Suh, Kune Y.; Chung, Chang Hyun; Park, Rae joon; Kim, Sang Baik

    2000-01-01

    This study is concerned with the experimental test and numerical analysis of the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. In the test, the metal pool is heated from the bottom surface and coolant is injected into the molten metal pool. Experiments were performed by changing the test section bottom surface temperature of the metal layer and the coolant injection rate. The two-phase boiling coolant experimental results are compared against the dry test without coolant or solidification of the molten metal pool, and against the crust formation experiment with subcooled coolant. Also, a numerical analysis is performed to check on the measured data. The numerical program is developed using the enthalpy method, the finite volume method and the SIMPLER algorithm. The experimental results of the heat transfer show general agreement with the calculated values. The present empirical test and numerical results of the heat transfer on the molten metal pool are apparently higher than those without coolant boiling. This is probably because this experiment was performed in concurrence of solidification in the molten metal pool and the rapid boiling of the coolant. The other experiments were performed without coolant boiling and the correlation was developed for the pure molten metal without phase change. (author)

  15. Test results of the reactor inlet coolant temperature control system of HTTR

    International Nuclear Information System (INIS)

    Saito, Kenji; Nakagawa, Shigeaki; Hirato, Yoji

    2004-04-01

    The reactor control system of HTTR is composed of the reactor power control system, the reactor inlet coolant temperature control system, the primary coolant flow rate control system and so on. The reactor control system of HTTR achieves reactor power 30 MW, reactor outlet coolant temperature 850degC, reactor inlet coolant temperature 395degC under the condition that primary coolant flow rate is fixed. In the Rise-to-Power Test, the performance test of the reactor inlet coolant temperature control system was carried out in order to confirm the control capability of this control system. This report shows the test results of performance test. As a result, the control parameters, which can control the reactor inlet coolant temperature stably during the reactor operation, were successfully selected. And it was confirmed that the reactor inlet coolant temperature control system has the capability of controlling the reactor inlet coolant temperature stably against any disturbances on the basis of operational condition of HTTR. (author)

  16. Fuel-Coolant Interactions - some Basic Studies at the UKAEA Culham Laboratory

    International Nuclear Information System (INIS)

    Reynolds, J.A.; Dullforce, T.A.; Peckover, R.S.; Vaughan, G.J.

    1976-01-01

    In a hypothetical fault sequence important effects of fuel-coolant interactions include voiding and dispersion of core debris as well as the pressure damage usually discussed. The development of the fuel-coolant interaction probably depends on any pre-mixing Weber break-up that may occur, and is therefore a function of the way the fuel and coolant come together. Four contact modes are identified: jetting, shock tube, drops and static, and Culham's experiments have been mainly concerned with simulating the falling drop mode by using molten tin in water. It was observed that the fuel-coolant interaction is a short series of violent coolant oscillations centred at a localized position on the drop, generating a spray of submillimeter sized debris. The interaction started spontaneously at a specific time after the drop first contacted the water. There was a definite limited fuel-coolant interaction zone on a plot of initial coolant temperature versus initial fuel temperature outside which interactions never occurred. The. interaction time was a function of the initial temperatures. Theoretical scaling formulae are given which describe the fuel-coolant interaction zone and dwell time. Bounds of fuel and coolant temperature below which fuel-coolant interactions do not occur are explained by freezing. Upper bounds of fuel and coolant temperatures above which there were no fuel-coolant interactions are interpreted in terms of heat transfer through vapour films of various thicknesses. In conclusion: We have considered the effects of fuel-coolant interactions in a hypothetical fault sequence, emphasising that debris and vapour production as well as the pressure pulse can be important factors. The fuel-coolant interaction has been classified into types, according to possible modes of mixing in the fault sequence. Culham has been studying one type, the self-triggering of falling drops, by simulant experiments. It is found that there is a definite zone of interaction on a plot

  17. Method for controlling a coolant liquid surface of cooling system instruments in an atomic power plant

    International Nuclear Information System (INIS)

    Monta, Kazuo.

    1974-01-01

    Object: To prevent coolant inventory within a cooling system loop in an atomic power plant from being varied depending on loads thereby relieving restriction of varied speed of coolant flow rate to lowering of a liquid surface due to short in coolant. Structure: Instruments such as a superheater, an evaporator, and the like, which constitute a cooling system loop in an atomic power plant, have a plurality of free liquid surface of coolant. Portions whose liquid surface is controlled and portions whose liquid surface is varied are adjusted in cross-sectional area so that the sum total of variation in coolant inventory in an instrument such as a superheater provided with an annulus portion in the center thereof and an inner cylindrical portion and a down-comer in the side thereof comes equal to that of variation in coolant inventory in an instrument such as an evaporator similar to the superheater. which is provided with an overflow pipe in its inner cylindrical portion or down-comer, thereby minimizing variation in coolant inventory of the entire coolant due to loads thus minimizing variation in varied speed of the coolant. (Kamimura, M.)

  18. Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

  19. Reactor coolant pump type RUV for Westinghouse Electric Company LLC reactor AP1000 TM

    International Nuclear Information System (INIS)

    Baumgarten, S.; Brecht, B.; Bruhns, U.; Fehring, P.

    2010-01-01

    The RUV is a reactor coolant pump, specially designed for the Westinghouse Electric Company LLC AP1000 TM reactor. It is a hermetically sealed, wet winding motor pump. The RUV is a very compact, vertical pump/motor unit, designed to fit into the compartment next to the reactor pressure vessel. Each of the two steam generators has two pump casings welded to the channel head by the suction nozzle. The pump/motor unit consists of a pump part, where a semi-axial impeller/diffuser combination is mounted in a one-piece pump casing. Computational Fluid Dynamics methods combined with various hydraulic tests in a 1:2 scale hydraulic test assure full compliance with the specific customer requirements. A short and rigid shaft, supported by a radial bearing, connects the impeller with the high inertia flywheel. This flywheel consists of a one-piece forged stainless steel cylinder, with an option for several smaller heavy metal cylinders inside. The flywheel is located inside the thermal barrier, which forms part of the pressure boundary. A specific arrangement of cooling water circuits guarantees a homogeneous temperature distribution in and around the flywheel, minimizes the friction losses of the flywheel and protects the motor from hot coolant. The driving torque is transmitted by the motor shaft, which itself is supported by two radial bearings. A three-phase, high-voltage squirrel-cage induction motor generates the driving torque. Due to the wet winding concept it is possible to achieve positive effects regarding motor lifetime. The cooling water is forced through the stator windings and the gap between rotor and stator by an auxiliary impeller. Furthermore, this wet winding motor concept has higher efficiency as compared to a canned motor since there are no eddy current losses. As part of the design process and in addition to the hydraulic scale model, a complete half scale model pump was built. It was used to verify the calculations performed like coast

  20. Optoelectronics circuits manual

    CERN Document Server

    Marston, R M

    2013-01-01

    Optoelectronics Circuits Manual covers the basic principles and characteristics of the best known types of optoelectronic devices, as well as the practical applications of many of these optoelectronic devices. The book describes LED display circuits and LED dot- and bar-graph circuits and discusses the applications of seven-segment displays, light-sensitive devices, optocouplers, and a variety of brightness control techniques. The text also tackles infrared light-beam alarms and multichannel remote control systems. The book provides practical user information and circuitry and illustrations.

  1. Modern TTL circuits manual

    CERN Document Server

    Marston, R M

    2013-01-01

    Modern TTL Circuits Manual provides an introduction to the basic principles of Transistor-Transistor Logic (TTL). This book outlines the major features of the 74 series of integrated circuits (ICs) and introduces the various sub-groups of the TTL family.Organized into seven chapters, this book begins with an overview of the basics of digital ICs. This text then examines the symbology and mathematics of digital logic. Other chapters consider a variety of topics, including waveform generator circuitry, clocked flip-flop and counter circuits, special counter/dividers, registers, data latches, com

  2. Circuit analysis with Multisim

    CERN Document Server

    Baez-Lopez, David

    2011-01-01

    This book is concerned with circuit simulation using National Instruments Multisim. It focuses on the use and comprehension of the working techniques for electrical and electronic circuit simulation. The first chapters are devoted to basic circuit analysis.It starts by describing in detail how to perform a DC analysis using only resistors and independent and controlled sources. Then, it introduces capacitors and inductors to make a transient analysis. In the case of transient analysis, it is possible to have an initial condition either in the capacitor voltage or in the inductor current, or bo

  3. Gallium Arsenide Domino Circuit

    Science.gov (United States)

    Yang, Long; Long, Stephen I.

    1990-01-01

    Advantages include reduced power and high speed. Experimental gallium arsenide field-effect-transistor (FET) domino circuit replicated in large numbers for use in dynamic-logic systems. Name of circuit denotes mode of operation, which logic signals propagate from each stage to next when successive stages operated at slightly staggered clock cycles, in manner reminiscent of dominoes falling in a row. Building block of domino circuit includes input, inverter, and level-shifting substages. Combinational logic executed in input substage. During low half of clock cycle, result of logic operation transmitted to following stage.

  4. 'Speedy' superconducting circuits

    International Nuclear Information System (INIS)

    Holst, T.

    1994-01-01

    The most promising concept for realizing ultra-fast superconducting digital circuits is the Rapid Single Flux Quantum (RSFQ) logic. The basic physical principle behind RSFQ logic, which include the storage and transfer of individual magnetic flux quanta in Superconducting Quantum Interference Devices (SQUIDs), is explained. A Set-Reset flip-flop is used as an example of the implementation of an RSFQ based circuit. Finally, the outlook for high-temperature superconducting materials in connection with RSFQ circuits is discussed in some details. (au)

  5. Multi-qubit circuit quantum electrodynamics

    Energy Technology Data Exchange (ETDEWEB)

    Viehmann, Oliver

    2013-09-03

    Circuit QED systems are macroscopic, man-made quantum systems in which superconducting artificial atoms, also called Josephson qubits, interact with a quantized electromagnetic field. These systems have been devised to mimic the physics of elementary quantum optical systems with real atoms in a scalable and more flexible framework. This opens up a variety of possible applications of circuit QED systems. For instance, they provide a promising platform for processing quantum information. Recent years have seen rapid experimental progress on these systems, and experiments with multi-component circuit QED architectures are currently starting to come within reach. In this thesis, circuit QED systems with multiple Josephson qubits are studied theoretically. We focus on simple and experimentally realistic extensions of the currently operated circuit QED setups and pursue investigations in two main directions. First, we consider the equilibrium behavior of circuit QED systems containing a large number of mutually noninteracting Josephson charge qubits. The currently accepted standard description of circuit QED predicts the possibility of superradiant phase transitions in such systems. However, a full microscopic treatment shows that a no-go theorem for superradiant phase transitions known from atomic physics applies to circuit QED systems as well. This reveals previously unknown limitations of the applicability of the standard theory of circuit QED to multi-qubit systems. Second, we explore the potential of circuit QED for quantum simulations of interacting quantum many-body systems. We propose and analyze a circuit QED architecture that implements the quantum Ising chain in a time-dependent transverse magnetic field. Our setup can be used to study quench dynamics, the propagation of localized excitations, and other non-equilibrium features in this paradigmatic model in the theory of non-equilibrium thermodynamics and quantumcritical phenomena. The setup is based on a

  6. Multi-qubit circuit quantum electrodynamics

    International Nuclear Information System (INIS)

    Viehmann, Oliver

    2013-01-01

    Circuit QED systems are macroscopic, man-made quantum systems in which superconducting artificial atoms, also called Josephson qubits, interact with a quantized electromagnetic field. These systems have been devised to mimic the physics of elementary quantum optical systems with real atoms in a scalable and more flexible framework. This opens up a variety of possible applications of circuit QED systems. For instance, they provide a promising platform for processing quantum information. Recent years have seen rapid experimental progress on these systems, and experiments with multi-component circuit QED architectures are currently starting to come within reach. In this thesis, circuit QED systems with multiple Josephson qubits are studied theoretically. We focus on simple and experimentally realistic extensions of the currently operated circuit QED setups and pursue investigations in two main directions. First, we consider the equilibrium behavior of circuit QED systems containing a large number of mutually noninteracting Josephson charge qubits. The currently accepted standard description of circuit QED predicts the possibility of superradiant phase transitions in such systems. However, a full microscopic treatment shows that a no-go theorem for superradiant phase transitions known from atomic physics applies to circuit QED systems as well. This reveals previously unknown limitations of the applicability of the standard theory of circuit QED to multi-qubit systems. Second, we explore the potential of circuit QED for quantum simulations of interacting quantum many-body systems. We propose and analyze a circuit QED architecture that implements the quantum Ising chain in a time-dependent transverse magnetic field. Our setup can be used to study quench dynamics, the propagation of localized excitations, and other non-equilibrium features in this paradigmatic model in the theory of non-equilibrium thermodynamics and quantumcritical phenomena. The setup is based on a

  7. Printed circuit for ATLAS

    CERN Multimedia

    Laurent Guiraud

    1999-01-01

    A printed circuit board made by scientists in the ATLAS collaboration for the transition radiaton tracker (TRT). This will read data produced when a high energy particle crosses the boundary between two materials with different electrical properties.

  8. Connector and electronic circuit assembly for improved wet insulation resistance

    Energy Technology Data Exchange (ETDEWEB)

    Reese, Jason A.; Teli, Samar R.; Keenihan, James R.; Langmaid, Joseph A.; Maak, Kevin D.; Mills, Michael E.; Plum, Timothy C.; Ramesh, Narayan

    2016-07-19

    The present invention is premised upon a connector and electronic circuit assembly (130) at least partially encased in a polymeric frame (200). The assembly including at least: a connector housing (230); at least one electrical connector (330); at least one electronic circuit component (430); and at least one barrier element (530).

  9. Some Thermophysical Properties of Blood Components and Coolants for Frozen Blood Shipping Containers

    Science.gov (United States)

    1989-09-01

    SP number by sending a DP reading. Subroutine : AutoControl Automatic control to set temperature. : Autodisplay Get ilL Thermocouple readings and...RETURN 133 AutoControl : Auto control Mode ON TIMER(ReportTime) GOSUB Autodisplay Update Screen in constant interval TIMER ON WHILE Success a 0 Turn off

  10. Power supply conditioning circuit

    Science.gov (United States)

    Primas, Lori E. (Inventor); Loveland, Rohan C. (Inventor)

    1988-01-01

    A conditioning circuit is provided with a constant current diode in series with a zener diode, the former having a high dynamic impedance and the latter a low dynamic impedance. The constant current diode can receive an input voltage with PARD. In conjunction with the zener diode fixed to a ground, a voltage divider is provided which can give an output voltage whose PARD was significantly reduced. The conditioning circuit is effective down to dc.

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

    International Nuclear Information System (INIS)

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

    1998-01-01

    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 extended 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

  12. Project Circuits in a Basic Electric Circuits Course

    Science.gov (United States)

    Becker, James P.; Plumb, Carolyn; Revia, Richard A.

    2014-01-01

    The use of project circuits (a photoplethysmograph circuit and a simple audio amplifier), introduced in a sophomore-level electric circuits course utilizing active learning and inquiry-based methods, is described. The development of the project circuits was initiated to promote enhanced engagement and deeper understanding of course content among…

  13. Effects of smoke on functional circuits

    International Nuclear Information System (INIS)

    Tanaka, T.J.

    1997-10-01

    Nuclear power plants are converting to digital instrumentation and control systems; however, the effects of abnormal environments such as fire and smoke on such systems are not known. There are no standard tests for smoke, but previous smoke exposure tests at Sandia National Laboratories have shown that digital communications can be temporarily interrupted during a smoke exposure. Another concern is the long-term corrosion of metals exposed to the acidic gases produced by a cable fire. This report documents measurements of basic functional circuits during and up to 1 day after exposure to smoke created by burning cable insulation. Printed wiring boards were exposed to the smoke in an enclosed chamber for 1 hour. For high-resistance circuits, the smoke lowered the resistance of the surface of the board and caused the circuits to short during the exposure. These circuits recovered after the smoke was vented. For low-resistance circuits, the smoke caused their resistance to increase slightly. A polyurethane conformal coating substantially reduced the effects of smoke. A high-speed digital circuit was unaffected. A second experiment on different logic chip technologies showed that the critical shunt resistance that would cause failure was dependent on the chip technology and that the components used in the smoke exposures were some of the most smoke tolerant. The smoke densities in these tests were high enough to cause changes in high impedance (resistance) circuits during exposure, but did not affect most of the other circuits. Conformal coatings and the characteristics of chip technologies should be considered when designing circuitry for nuclear power plant safety systems, which must be highly reliable under a variety of operating and accident conditions. 10 refs., 34 figs., 18 tabs

  14. Effects of smoke on functional circuits

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.J.

    1997-10-01

    Nuclear power plants are converting to digital instrumentation and control systems; however, the effects of abnormal environments such as fire and smoke on such systems are not known. There are no standard tests for smoke, but previous smoke exposure tests at Sandia National Laboratories have shown that digital communications can be temporarily interrupted during a smoke exposure. Another concern is the long-term corrosion of metals exposed to the acidic gases produced by a cable fire. This report documents measurements of basic functional circuits during and up to 1 day after exposure to smoke created by burning cable insulation. Printed wiring boards were exposed to the smoke in an enclosed chamber for 1 hour. For high-resistance circuits, the smoke lowered the resistance of the surface of the board and caused the circuits to short during the exposure. These circuits recovered after the smoke was vented. For low-resistance circuits, the smoke caused their resistance to increase slightly. A polyurethane conformal coating substantially reduced the effects of smoke. A high-speed digital circuit was unaffected. A second experiment on different logic chip technologies showed that the critical shunt resistance that would cause failure was dependent on the chip technology and that the components used in the smoke exposures were some of the most smoke tolerant. The smoke densities in these tests were high enough to cause changes in high impedance (resistance) circuits during exposure, but did not affect most of the other circuits. Conformal coatings and the characteristics of chip technologies should be considered when designing circuitry for nuclear power plant safety systems, which must be highly reliable under a variety of operating and accident conditions. 10 refs., 34 figs., 18 tabs.

  15. Simulation of a small loss of coolant accident by using RETINA V1.0D code

    International Nuclear Information System (INIS)

    Hazi, G.; Mayer, G.; Farkas, I.; Makovi, P.; El-Kafas, A.A.

    2001-01-01

    RETINA has been developed for modeling of two-phase flow situations in full-scope simulators of nuclear power plants. A special feature of RETINA is that both RETINA V1.0D (drift-flux - 5 equations) and RETINA V1.0-2V (two-fluid - 6 equations) approach are available in the code and the same constitutive relations are used in both cases. The governing equations are discretized implicitly, and an automatic derivation algorithm determines the Jacobian matrix, which is partitioned taking into account the special structure of nuclear power plants. Partitioned inverse formula is used to solve the global equation system providing the possibility of multi-level parallelization. Heat structures are modeled in two dimensions and are coupled to the flow equations explicitly. Since the code will be used in real-time simulators, we paid special attention to time-effective solution. In this paper, we demonstrate the ability of our code by simulating a small loss of coolant accident Paks Model Circuit (PMK). The simulation results are compared to real measurements obtained by Paks Model Circuit

  16. A chaotic circuit based on Hewlett-Packard memristor

    Science.gov (United States)

    Buscarino, Arturo; Fortuna, Luigi; Frasca, Mattia; Valentina Gambuzza, Lucia

    2012-06-01

    Memristors are gaining increasing attention as next generation electronic devices. They are also becoming commonly used as fundamental blocks for building chaotic circuits, although often arbitrary (typically piece-wise linear or cubic) flux-charge characteristics are assumed. In this paper, a chaotic circuit based on the mathematical realistic model of the HP memristor is introduced. The circuit makes use of two HP memristors in antiparallel. Numerical results showing some of the chaotic attractors generated by this circuit and the behavior with respect to changes in its component values are described.

  17. A numerical analysis of molten metal drop and coolant interaction

    International Nuclear Information System (INIS)

    Cao, X.; Hajima, Ryoichi; Furuta, Kazuo; Kondo, Shunsuke

    2000-01-01

    The interaction of molten metal drop and coolant is numerically analyzed to investigate the mechanism of fragmentation in vapor explosion. The numerical study is carried out by using a developed simulation code based on multi-phase thermal hydraulic model, which includes physical phenomena required for the analysis: heat transfer, mass change, liquid evaporation and treatment of surface. Several computational techniques are also implemented to improve the efficiency and stability of the numerical scheme. The obtained numerical results show that the growth of spikes on the molten metal drop surface is similar to that observed in Ciccarelli's experiment. The numerical study suggests that quick growth of spikes is the essential mechanism of fragmentation, which is caused by Taylor instability. (author)

  18. Environmental radiological consequences of a loss of coolant accident

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1981-01-01

    The elaboration of a calculation model to determine safety areas, named Exclusion Zone and Low Population Zone for nuclear power plants, is dealt with. These areas are determined from a radioactive doses calculation for the population living around the NPP after occurence of a postulated ' Maximum Credible Accident' (MCA). The MCA is defined as an accident with complete loss of primary coolant and consequent fusion of a substantial portion of the reactor core. In the calculations carried out, data from NPP Angra I were used and the assumptions made were conservative, to be compatible with licensing requirements. Under the most pessimistic assumption (no filters) the values of 410m and 1000m were obtained for the Exclusion Zone and Low Population Zone radii, respectivily. (Author) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

    DeMuth, J. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meier, W. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frantoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reyes, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    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.

  20. Reactor coolant pump shaft seal behavior during station blackout

    International Nuclear Information System (INIS)

    Kittmer, C.A.; Wensel, R.G.; Rhodes, D.B.; Metcalfe, R.; Cotnam, B.M.; Gentili, H.; Mings, W.J.

    1985-04-01

    A testing program designed to provide fundamental information pertaining to the behavior of reactor coolant pump (RCP) shaft seals during a postulated nuclear power plant station blackout has been completed. One seal assembly, utilizing both hydrodynamic and hydrostatic types of seals, was modeled and tested. Extrusion tests were conducted to determine if seal materials could withstand predicted temperatures and pressures. A taper-face seal model was tested for seal stability under conditions when leaking water flashes to steam across the seal face. Test information was then used as the basis for a station blackout analysis. Test results indicate a potential problem with an elastomer material used for O-rings by a pump vendor; that vendor is considering a change in material specification. Test results also indicate a need for further research on the generic issue of RCP seal integrity and its possible consideration for designation as an unresolved safety issue

  1. Assessment of TRIGA pulsing reactor safety without loss of coolant

    International Nuclear Information System (INIS)

    Baers, B.; Kaall, L.

    1979-11-01

    The fuel temperature, fuel can pressure and cladding stresses in the TRIGA fuel elements at various transient states has in part been measured and semiempirically estimated in cases where no experiments could be carried out. To the end of obtaining more reliable and up-to-date data and reactor parameters, reactor power and fuel temperatures were measured both at steady and transient states, including pulsing up to 280 MW. A novel and more realistic model for the fuel temperature at pulsing was also presented and used in the assessments. Based on the present assessment with the water coolant present and a maximum excess reactivity of (Δksub(max)/β = ) 4 $, no reactivity induced transients can bring about any undue hazards to the reactor or to the surroundings. (author)

  2. SIMMER-III applications to fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

  3. A new inexpensive electrochemical meter for oxygen in sodium coolant

    International Nuclear Information System (INIS)

    Periaswami, G.; Rajan Babu, S.S.; Mathews, C.K.

    1987-01-01

    This report describes the development of an inexpensive oxygen meter for sodium coolant and gives the results of the test experiments. Calcia stabilized zirconia has been found to have necessary domain boundary characteristics at low temperatures for use as oxygen sensor in liquid sodium system. It is possible to obtain acceptable sensor cell resistance at temperatures as low as 230 C if K, K 2 O or Na, Na 2 O is used as reference electrode. The performance of these cells has been tested in bench top sodium loops over long periods. Their performance in terms of cell-out put variation with change in oxygen concentration in sodium has been found to be satisfactory. They also have sufficiently long life times since the kinetics of sodium attack on the electrolyte is slow at low temperatures. (author). 17 refs., 6 figs

  4. Insight into steam explosion in stratified melt-coolant configuration

    OpenAIRE

    Grishchenko, Dmitry; Konovalenko, Alexander; Karbojian, Aram; Kudinova, Valtyna; Bechta, Sevostian; Kudinov, Pavel

    2013-01-01

    Release of core melt from failed reactor vessel into a pool of water is adopted in several existing designs of light water reactors (LWRs) as an element of severe accident  mitigation  strategy.  When  vessel  breach  is  large  and  water  pool  is shallow,  released  corium  melt  can  reach  containment  floor  in  liquid  form  and spread under water creating a stratified configuration of melt covered by coolant. Steam  explosion  in  such  stratified  configuration  was  long  believed  ...

  5. Condensing heat transfer following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Krotiuk, W.J.; Rubin, M.B.

    1978-01-01

    A new method for calculating the steam mass condensation energy removal rates on cold surfaces in contact with an air-steam mixture has been developed. This method is based on the principles of mass diffusion of steam from an area of high concentration to the condensing surface, which is an area of low steam concentration. This new method of calculating mass condensation has been programmed into the CONTEMPT-LT Mod 26 computer code, which calculates the pressure and temperature transients inside a light water reactor containment following a loss-of-coolant accident. The condensing heat transfer coefficient predicted by the mass diffusion method is compared to existing semi-empirical correlations and to the experimental results of the Carolinas Virginia Tube Reactor Containment natural decay test. Closer agreement with test results is shown in the calculation of containment pressure, temperature, and heat sink surface temperature using the mass diffusion condensation method than when using any existing semi-empirical correlation

  6. Vibration Damping Circuit Card Assembly

    Science.gov (United States)

    Hunt, Ronald Allen (Inventor)

    2016-01-01

    A vibration damping circuit card assembly includes a populated circuit card having a mass M. A closed metal container is coupled to a surface of the populated circuit card at approximately a geometric center of the populated circuit card. Tungsten balls fill approximately 90% of the metal container with a collective mass of the tungsten balls being approximately (0.07) M.

  7. Experimental Investigation of Heat Transfer Characteristics of Automobile Radiator using TiO2-Nanofluid Coolant

    Science.gov (United States)

    Salamon, V.; Senthil kumar, D.; Thirumalini, S.

    2017-08-01

    The use of nanoparticle dispersed coolants in automobile radiators improves the heat transfer rate and facilitates overall reduction in size of the radiators. In this study, the heat transfer characteristics of water/propylene glycol based TiO2 nanofluid was analyzed experimentally and compared with pure water and water/propylene glycol mixture. Two different concentrations of nanofluids were prepared by adding 0.1 vol. % and 0.3 vol. % of TiO2 nanoparticles into water/propylene glycol mixture (70:30). The experiments were conducted by varying the coolant flow rate between 3 to 6 lit/min for various coolant temperatures (50°C, 60°C, 70°C, and 80°C) to understand the effect of coolant flow rate on heat transfer. The results showed that the Nusselt number of the nanofluid coolant increases with increase in flow rate. At low inlet coolant temperature the water/propylene glycol mixture showed higher heat transfer rate when compared with nanofluid coolant. However at higher operating temperature and higher coolant flow rate, 0.3 vol. % of TiO2 nanofluid enhances the heat transfer rate by 8.5% when compared to base fluids.

  8. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Science.gov (United States)

    2010-01-01

    ... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high... operated from the control room. (b) The design of the vents and associated controls, instruments and power... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting...

  9. Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor

    Directory of Open Access Journals (Sweden)

    N. Dharmaraju

    2008-01-01

    Full Text Available The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carries the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be going into operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and modeling studies have been carried out. A good correlation has been achieved between the results of experimental and analytical models. The operating life of a typical coolant channel typically ranges from 10 to 15 full-power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. Through the study of dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. A study has been also carried out to characterize the fuel vibration under different flow condition.

  10. Use of Russian technology of ship reactors with lead-bismuth coolant in nuclear power

    International Nuclear Information System (INIS)

    Zrodnikov, A.V.; Chitaykin, V.I.; Gromov, B.F.; Grigoryv, O.G.; Dedoul, A.V.; Toshinsky, G.I.; Dragunov, Yu.G.; Stepanov, V.S.

    2000-01-01

    The experience of using lead-bismuth coolant in Russian nuclear submarine reactors has been presented. The fundamental statements of the concept of using the reactors cooled by lead-bismuth alloy in nuclear power have been substantiated. The results of developments for using lead bismuth coolant in nuclear power have been presented. (author)

  11. Thermo-Hydraulic Behaviour of Coolant in Nuclear Reactor VVER-440 Under Refuelling Conditions

    Directory of Open Access Journals (Sweden)

    Paulech Juraj

    2017-04-01

    Full Text Available The paper presents the numerical simulation of thermo-hydraulic behaviour of coolant in the VVER- 440 nuclear reactor under standard outage conditions. Heating-up and flow of coolant between the reactor pressure vessel and spent fuel storage pool are discussed.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D.P.; Lukas, M.; Lynch, B.K. [Spectro Incorporated, Littleton, MA (United States)

    1997-12-31

    Rotating disk electrode (RDE) arc emission spectrometers are user` 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 it has nearly been forgotten by most practitioners that when RDE spectrometers were first introduced more than 40 years ago, they were routinely used for aqueous samples as well. This presentation reviews early methods of aqueous sample analysis using RDE technology. This presentation also describes recent work to calibrate an RDE spectrometer for both water samples and for engine coolant samples which are a mixture of approximately 50 % water and 50 % ethylene or propylene glycol. Limits of detection determined for aqueous standards are comparable to limits of detection for oil standards. Repeatability of aqueous samples is comparable to the repeatability achieved for oil samples. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and rotating disk electrode (RDE) spectrometers is presented. Not surprisingly, RDE results are significantly higher for samples containing particles larger than a few micrometers. Although limits of detection for aqueous samples are not as low as can be achieved using the more modern ICP spectrometric method or the more cumbersome atomic absorption (AA) method, this presentation suggests that RDE spectrometers may be appropriate for certain types of aqueous samples in situations where the more sensitive ICP or AA spectrometers and the laboratory environment and skilled personnel needed for them to operate are not conveniently available. (orig.) 4 refs.

  14. Dismantling of large components from the PHENIX reactor

    International Nuclear Information System (INIS)

    Roux, A.

    2013-01-01

    The PHENIX reactor was shut down in 2009. The cleaning and dismantling preparation operations are underway. These operations include dealing with large removable components such as primary coolant pumps, intermediate heat exchanger and heat exchanger blanking device (DOTE). This presentation describes the waste transformation operations performed on a large component, from its extraction from the reactor core until transformation to waste for disposal

  15. Nanofabrication of Plasmonic Circuits Containing Single Photon Sources

    DEFF Research Database (Denmark)

    Siampour, Hamidreza; Kumar, Shailesh; Bozhevolnyi, Sergey I.

    2017-01-01

    Nanofabrication of photonic components based on dielectric loaded surface plasmon polariton waveguides (DLSPPWs) excited by single nitrogen vacancy (NV) centers in nanodiamonds is demonstrated. DLSPPW circuits are built around NV containing nanodiamonds, which are certified to be single...

  16. Low latency asynchronous interface circuits

    Science.gov (United States)

    Sadowski, Greg

    2017-06-20

    In one form, a logic circuit includes an asynchronous logic circuit, a synchronous logic circuit, and an interface circuit coupled between the asynchronous logic circuit and the synchronous logic circuit. The asynchronous logic circuit has a plurality of asynchronous outputs for providing a corresponding plurality of asynchronous signals. The synchronous logic circuit has a plurality of synchronous inputs corresponding to the plurality of asynchronous outputs, a stretch input for receiving a stretch signal, and a clock output for providing a clock signal. The synchronous logic circuit provides the clock signal as a periodic signal but prolongs a predetermined state of the clock signal while the stretch signal is active. The asynchronous interface detects whether metastability could occur when latching any of the plurality of the asynchronous outputs of the asynchronous logic circuit using said clock signal, and activates the stretch signal while the metastability could occur.

  17. Associative Pattern Recognition In Analog VLSI Circuits

    Science.gov (United States)

    Tawel, Raoul

    1995-01-01

    Winner-take-all circuit selects best-match stored pattern. Prototype cascadable very-large-scale integrated (VLSI) circuit chips built and tested to demonstrate concept of electronic associative pattern recognition. Based on low-power, sub-threshold analog complementary oxide/semiconductor (CMOS) VLSI circuitry, each chip can store 128 sets (vectors) of 16 analog values (vector components), vectors representing known patterns as diverse as spectra, histograms, graphs, or brightnesses of pixels in images. Chips exploit parallel nature of vector quantization architecture to implement highly parallel processing in relatively simple computational cells. Through collective action, cells classify input pattern in fraction of microsecond while consuming power of few microwatts.

  18. A fast charge integrating and shaping circuit

    International Nuclear Information System (INIS)

    Kulka, Z.; Szoncso, F.

    1990-01-01

    The development of a low cost fast charge integrating and shaping circuit (FCISC) was motivated by the need for an interface between the photomultipliers of an existing hadronic calorimeter and recently developed new readout electronics designed to match the output of small ionization chambers for the upgraded UA1 detector at the CERN proton-antiproton collider. This paper describes the design principles of gated and ungated charge integrating and shaping circuits. An FCISC prototype using discrete components was made and its properties were determined with a computerized test setup. Finally an SMD implementation of the FCISC is presented and the performance is reported. (orig.)

  19. RF and microwave coupled-line circuits

    CERN Document Server

    Mongia, R K; Bhartia, P; Hong, J; Gupta, K C

    2007-01-01

    This extensively revised edition of the 1999 Artech House classic, RF and Microwave Coupled-Line Circuits, offers you a thoroughly up-to-date understanding of coupled line fundamentals, explaining their applications in designing microwave and millimeter-wave components used in today's communications, microwave, and radar systems. The Second Edition includes a wealth of new material, particularly relating to applications. You find brand new discussions on a novel simple design technique for multilayer coupled circuits, high pass filters using coupled lines, software packages used for filter des

  20. Estimating the short-circuit impedance

    DEFF Research Database (Denmark)

    Nielsen, Arne Hejde; Pedersen, Knud Ole Helgesen; Poulsen, Niels Kjølstad

    1997-01-01

    and current are derived each period, and the short-circuit impedance is estimated from variations in these components created by load changes in the grid. Due to the noisy and dynamic grid with high harmonic distortion it is necessary to threat the calculated values statistical. This is done recursively...... through a RLS-algorithm. The algorithms have been tested and implemented on a PC at a 132 kV substation supplying a rolling mill. Knowing the short-circuit impedance gives the rolling mill an opportunity to adjust the arc furnace operation to keep flicker below a certain level. Therefore, the PC performs...

  1. Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

    International Nuclear Information System (INIS)

    Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

    2014-01-01

    The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

  2. The management of the Primary Coolant in HANARO (1998-2006)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Wook; Lee, Mun; Park, Se Il; Hur, Soon Ock; Lim, Kyeong Hwan; Ahn, Guk Hoon

    2007-11-15

    The primary coolant quality management in HANARO is important to prevent corrosion of nuclear fuel and reactor structure material. The purity of the primary coolant has monitored by measuring the electrical conductivity of the primary purification system and pH periodically. The status of the primary coolant quality management was investigated from 1998 to 2006 with the measured data. The electrical conductivity of the primary coolant becomes low as much as the flow-rate of the primary purification system increased because the performance of the resin is in proportion to flow rate through resin. In general, the pH and the electrical conductivity of the primary coolant have been managed within 1 {mu}S/cm which is a target criteria of HANARO.

  3. Technical findings related to Generic Issue 23: Reactor coolant pump seal failure

    International Nuclear Information System (INIS)

    Ruger, C.J.; Luckas, W.J. Jr.

    1989-03-01

    Reactor coolant pumps contain mechanical seals to limit the leakage of pressurized coolant from the reactor coolant system to the containment. These seals have the potential to leak, and a few have degraded and even failed resulting in a small break loss of coolant accident (LOCA). As a result, ''Reactor Coolant Pump Seal Failure,'' Generic Issue 23 was established. This report summarizes the findings of a technical investigation generated as part of the program to resolve this issue. These technical findings address the various fact-finding issue tasks developed for the action plan associated with the generic issue, namely background information on seal failure, evaluation of seal cooling, and mechanical- and maintenance-induced failure mechanisms. 46 refs., 15 figs., 14 tabs

  4. TRANSENERGY S: computer codes for coolant temperature prediction in LMFBR cores during transient events

    International Nuclear Information System (INIS)

    Glazer, S.; Todreas, N.; Rohsenow, W.; Sonin, A.

    1981-02-01

    This document is intended as a user/programmer manual for the TRANSENERGY-S computer code. The code represents an extension of the steady state ENERGY model, originally developed by E. Khan, to predict coolant and fuel pin temperatures in a single LMFBR core assembly during transient events. Effects which may be modelled in the analysis include temporal variation in gamma heating in the coolant and duct wall, rod power production, coolant inlet temperature, coolant flow rate, and thermal boundary conditions around the single assembly. Numerical formulations of energy equations in the fuel and coolant are presented, and the solution schemes and stability criteria are discussed. A detailed description of the input deck preparation is presented, as well as code logic flowcharts, and a complete program listing. TRANSENERGY-S code predictions are compared with those of two different versions of COBRA, and partial results of a 61 pin bundle test case are presented

  5. A New Simple Chaotic Circuit Based on Memristor

    Science.gov (United States)

    Wu, Renping; Wang, Chunhua

    In this paper, a new memristor is proposed, and then an emulator built from off-the-shelf solid state components imitating the behavior of the proposed memristor is presented. Multisim simulation and breadboard experiment are done on the emulator, exhibiting a pinched hysteresis loop in the voltage-current plane when the emulator is driven by a periodic excitation voltage. In addition, a new simple chaotic circuit is designed by using the proposed memristor and other circuit elements. It is exciting that this circuit with only a linear negative resistor, a capacitor, an inductor and a memristor can generate a chaotic attractor. The dynamical behaviors of the proposed chaotic system are analyzed by Lyapunov exponents, phase portraits and bifurcation diagrams. Finally, an electronic circuit is designed to implement the chaotic system. For the sake of simple circuit topology, the proposed chaotic circuit can be easily manufactured at low cost.

  6. Sensitivity Enhancement of Wheatstone Bridge Circuit for Resistance Measurement

    Directory of Open Access Journals (Sweden)

    Tarikul ISLAM

    2009-08-01

    Full Text Available Present work deals with the development of a low cost, appreciably accurate and sensitive electronic circuit for resistive sensor. The circuit is based on the modification of the Wheatstone bridge using active devices. It helps measurement of the incremental resistance precisely and linearly. It requires few components for its hardware implementation and found to be suitable in case there is small change in resistance due to change in physical quantity or chemical analytes to be measured. Theory of the proposed bridge circuit has been discussed and experimental results have been compared with conventional full bridge circuit. Experiments have been conducted with metallic strain gauge sensor but it can be utilized to other resistive sensors. Results show that the output of the circuit is almost four times more than usual full Wheatstone bridge circuit. Experimental results show that the errors due to the effects of the ambient temperature and connecting lead resistance are minimized.

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

    Science.gov (United States)

    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.

  8. Receiver Gain Modulation Circuit

    Science.gov (United States)

    Jones, Hollis; Racette, Paul; Walker, David; Gu, Dazhen

    2011-01-01

    A receiver gain modulation circuit (RGMC) was developed that modulates the power gain of the output of a radiometer receiver with a test signal. As the radiometer receiver switches between calibration noise references, the test signal is mixed with the calibrated noise and thus produces an ensemble set of measurements from which ensemble statistical analysis can be used to extract statistical information about the test signal. The RGMC is an enabling technology of the ensemble detector. As a key component for achieving ensemble detection and analysis, the RGMC has broad aeronautical and space applications. The RGMC can be used to test and develop new calibration algorithms, for example, to detect gain anomalies, and/or correct for slow drifts that affect climate-quality measurements over an accelerated time scale. A generalized approach to analyzing radiometer system designs yields a mathematical treatment of noise reference measurements in calibration algorithms. By treating the measurements from the different noise references as ensemble samples of the receiver state, i.e. receiver gain, a quantitative description of the non-stationary properties of the underlying receiver fluctuations can be derived. Excellent agreement has been obtained between model calculations and radiometric measurements. The mathematical formulation is equivalent to modulating the gain of a stable receiver with an externally generated signal and is the basis for ensemble detection and analysis (EDA). The concept of generating ensemble data sets using an ensemble detector is similar to the ensemble data sets generated as part of ensemble empirical mode decomposition (EEMD) with exception of a key distinguishing factor. EEMD adds noise to the signal under study whereas EDA mixes the signal with calibrated noise. It is mixing with calibrated noise that permits the measurement of temporal-functional variability of uncertainty in the underlying process. The RGMC permits the evaluation of EDA by

  9. Liquid level measurement on coolant pipeline using Raman distributed temperature sensor

    International Nuclear Information System (INIS)

    Kasinathan, M.; Sosamma, S.; Babu Rao, C.; Murali, N.; Jayakumar, T.

    2011-01-01

    Optical fibre based Raman Distributed Temperature Sensor (RDTS) has been widely used for temperature monitoring in oil pipe line, power cable and environmental monitoring. Recently it has gained importance in nuclear reactor owing to its advantages like continuous, distributed temperature monitoring and immunity from electromagnetic interference. It is important to monitor temperature based level measurement in sodium capacities and in coolant pipelines for Fast Breeder Reactor (FBR). This particular application is used for filling and draining sodium in storage tank of sodium circuits of Fast breeder reactor. There are different conventional methods to find out the sodium level in the storage tank of sodium cooled reactors. They are continuous level measurement and discontinuous level measurement. For continuous level measurement, mutual inductance type level probes are used. The disadvantage of using this method is it needs a temperature compensation circuit. For discontinuous level measurement, resistance type discontinuous level probe and mutual inductance type discontinuous level probe are used. In resistance type discontinuous level probe, each level needs a separate probe. To overcome these disadvantages, RDTS is used for level measurement based distributed temperature from optical fibre as sensor. The feasibility of using RDTS for measurement of temperature based level measurement sensor is studied using a specially designed test set-up and using hot water, instead of sodium. The test set-up consist of vertically erected Stainless Steel (SS) pipe of length 2m and diameter 10cm, with provision for filling and draining out the liquid. Bare graded index multimode fibre is laid straight along the length of the of the SS pipe. The SS pipe is filled with hot water at various levels. The hot water in the SS pipe is maintained at constant temperature by insulating the SS pipe. The temperature profile of the hot water at various levels is measured using RDTS. The

  10. Small circuits for cryptography.

    Energy Technology Data Exchange (ETDEWEB)

    Torgerson, Mark Dolan; Draelos, Timothy John; Schroeppel, Richard Crabtree; Miller, Russell D.; Anderson, William Erik

    2005-10-01

    This report examines a number of hardware circuit design issues associated with implementing certain functions in FPGA and ASIC technologies. Here we show circuit designs for AES and SHA-1 that have an extremely small hardware footprint, yet show reasonably good performance characteristics as compared to the state of the art designs found in the literature. Our AES performance numbers are fueled by an optimized composite field S-box design for the Stratix chipset. Our SHA-1 designs use register packing and feedback functionalities of the Stratix LE, which reduce the logic element usage by as much as 72% as compared to other SHA-1 designs.

  11. Electronic circuits fundamentals & applications

    CERN Document Server

    Tooley, Mike

    2015-01-01

    Electronics explained in one volume, using both theoretical and practical applications.New chapter on Raspberry PiCompanion website contains free electronic tools to aid learning for students and a question bank for lecturersPractical investigations and questions within each chapter help reinforce learning Mike Tooley provides all the information required to get to grips with the fundamentals of electronics, detailing the underpinning knowledge necessary to appreciate the operation of a wide range of electronic circuits, including amplifiers, logic circuits, power supplies and oscillators. The

  12. Offset cancelling circuit

    OpenAIRE

    Wiegerink, Remco J.; Seevinck, Evert; de Jager, Wim

    1989-01-01

    A monolithic offset cancelling circuit to reduce the offset voltage at an integrated audio-amplifier output is described. This offset voltage is detected using a low-pass filter with a very large time constant for which only one small on-chip capacitor is needed. The circuit was realized with a bipolar cell-based semicustom array. Measurements have shown that a -3-dB bandwidth below 5 Hz can be realized with a capacitor value of 50 pF. The resulting offset voltage at the audio-amplifier outpu...

  13. Circuit design for reliability

    CERN Document Server

    Cao, Yu; Wirth, Gilson

    2015-01-01

    This book presents physical understanding, modeling and simulation, on-chip characterization, layout solutions, and design techniques that are effective to enhance the reliability of various circuit units.  The authors provide readers with techniques for state of the art and future technologies, ranging from technology modeling, fault detection and analysis, circuit hardening, and reliability management. Provides comprehensive review on various reliability mechanisms at sub-45nm nodes; Describes practical modeling and characterization techniques for reliability; Includes thorough presentation of robust design techniques for major VLSI design units; Promotes physical understanding with first-principle simulations.

  14. Circuit architecture derivation starting from a formal requirements specification considering a DDS as example

    Science.gov (United States)

    Garbe, H.; Richter, R.; Jentschel, H.-J.

    2004-05-01

    Based on a formal specification of a direct digital synthesis (DDS) and assuming the availability of a set of possible circuit architectures we derive a customised system configuration. e calculate the design parameters that can be used for the specification to synthesise the circuit components. We show how the derived parameters and the selected IC technology influence the complexity of the circuit implementation.

  15. Scaling of pneumatic digital logic circuits.

    Science.gov (United States)

    Duncan, Philip N; Ahrar, Siavash; Hui, Elliot E

    2015-03-07

    The scaling of integrated circuits to smaller dimensions is critical for achieving increased system complexity and speed. Digital logic circuits composed of pneumatic microfluidic components have to this point been limited to a circuit density of 2-4 gates cm(-2), constraining the complexity of the digital systems that can be achieved. We explored the use of precision machining techniques to reduce the size of pneumatic valves and resistors, and to achieve more accurate and efficient placement of ports and vias. In this way, we attained an order of magnitude increase in circuit density, reaching as high as 36 gates cm(-2). A 12-bit binary counter circuit composed of 96 gates was realized in an area of 360 mm(2). The reduction in size also brought an order of magnitude increase in speed. The frequency of a 13-stage ring oscillator increased from 2.6 Hz to 22.1 Hz, and the maximum clock frequency of a binary counter increased from 1/3 Hz to 6 Hz.

  16. Circuit for Communication Over Power Lines

    Science.gov (United States)

    Krasowski, Michael J.; Prokop, Normal F.; Greer, Lawrence C., III; Nappier, Jennifer

    2011-01-01

    Many distributed systems share common sensors and instruments along with a common power line supplying current to the system. A communication technique and circuit has been developed that allows for the simple inclusion of an instrument, sensor, or actuator node within any system containing a common power bus. Wherever power is available, a node can be added, which can then draw power for itself, its associated sensors, and actuators from the power bus all while communicating with other nodes on the power bus. The technique modulates a DC power bus through capacitive coupling using on-off keying (OOK), and receives and demodulates the signal from the DC power bus through the same capacitive coupling. The circuit acts as serial modem for the physical power line communication. The circuit and technique can be made of commercially available components or included in an application specific integrated circuit (ASIC) design, which allows for the circuit to be included in current designs with additional circuitry or embedded into new designs. This device and technique moves computational, sensing, and actuation abilities closer to the source, and allows for the networking of multiple similar nodes to each other and to a central processor. This technique also allows for reconfigurable systems by adding or removing nodes at any time. It can do so using nothing more than the in situ power wiring of the system.

  17. Circuit simulation and physical implementation for a memristor-based colpitts oscillator

    Science.gov (United States)

    Deng, Hongmin; Wang, Dongping

    2017-03-01

    This paper implements two kinds of memristor-based colpitts oscillators, namely, the circuit where the memristor is added into the feedback network of the oscillator in parallel and series, respectively. First, a MULTISIM simulation circuit for the memristive colpitts oscillator is built, where an emulator constructed by some off-the-shelf components is utilized to replace the memristor. Then the physical system is implemented in terms of the MULTISIM simulation circuit. Circuit simulation and experimental study show that this memristive colpitts oscillator can exhibit periodic, quasi-periodic, and chaotic behaviors with certain parameter's variances. Besides, in a sense, the circuit is robust with circuit parameters and device types.

  18. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

    2010-01-15

    According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

  19. Performance Analysis of Thermoelectric Based Automotive Waste Heat Recovery System with Nanofluid Coolant

    Directory of Open Access Journals (Sweden)

    Zhi Li

    2017-09-01

    Full Text Available Output performance of a thermoelectric-based automotive waste heat recovery system with a nanofluid coolant is analyzed in this study. Comparison between Cu-Ethylene glycol (Cu-EG nanofluid coolant and ethylene glycol with water (EG-W coolant under equal mass flow rate indicates that Cu-EG nanofluid as a coolant can effectively improve power output and thermoelectric conversion efficiency for the system. Power output enhancement for a 3% concentration of nanofluid is 2.5–8 W (12.65–13.95% compared to EG-Water when inlet temperature of exhaust varies within 500–710 K. The increase of nanofluid concentration within a realizable range (6% has positive effect on output performance of the system. Study on the relationship between total area of thermoelectric modules (TEMs and output performance of the system indicates that optimal total area of TEMs exists for maximizing output performance of the system. Cu-EG nanofluid as coolant can decrease optimal total area of TEMs compared with EG-W, which will bring significant advantages for the optimization and arrangement of TEMs whether the system space is sufficient or not. Moreover, power output enhancement under Cu-EG nanofluid coolant is larger than that of EG-W coolant due to the increase of hot side heat transfer coefficient of TEMs.

  20. Exhaust temperature analysis of four stroke diesel engine by using MWCNT/Water nanofluids as coolant

    Science.gov (United States)

    Muruganandam, M.; Mukesh Kumar, P. C.

    2017-10-01

    There has been a continuous improvement in designing of cooling system and in quality of internal combustion engine coolants. The liquid engine coolant used in early days faced many difficulties such as low boiling, freezing points and inherently poor thermal conductivity. Moreover, the conventional coolants have reached their limitations of heat dissipating capacity. New heat transfer fluids have been developed and named as nanofluids to try to replace traditional coolants. Moreover, many works are going on the application of nanofluids to avail the benefits of them. In this experimental investigation, 0.1, 0.3 and 0.5% volume concentrations of multi walled carbon nanotube (MWCNT)/water nanofluids have been prepared by two step method with surfactant and is used as a coolant in four stroke single cylinder diesel engine to assess the exhaust temperature of the engine. The nanofluid prepared is characterized with scanning electron microscope (SEM) to confirm uniform dispersion and stability of nanotube with zeta potential analyzer. Experimental tests are performed by various mass flow rate such as 270 300 330 LPH (litre per hour) of coolant nanofluids and by changing the load in the range of 0 to 2000 W and by keeping the engine speed constant. It is found that the exhaust temperature decreases by 10-20% when compared to water as coolant at the same condition.

  1. Physics study of Canada deuterium uranium lattice with coolant void reactivity analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Shin, Ho Cheol [Korea Hydro and Nuclear Power Central Research Institute (KHNP-CRI), Daejeon (Korea, Republic of)

    2017-02-15

    This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the 2 x 2 checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

  2. Physics Study of Canada Deuterium Uranium Lattice with Coolant Void Reactivity Analysis

    Directory of Open Access Journals (Sweden)

    Jinsu Park

    2017-02-01

    Full Text Available This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700 fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the 2 × 2 checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

  3. ESD analog circuits and design

    CERN Document Server

    Voldman, Steven H

    2014-01-01

    A comprehensive and in-depth review of analog circuit layout, schematic architecture, device, power network and ESD design This book will provide a balanced overview of analog circuit design layout, analog circuit schematic development, architecture of chips, and ESD design.  It will start at an introductory level and will bring the reader right up to the state-of-the-art. Two critical design aspects for analog and power integrated circuits are combined. The first design aspect covers analog circuit design techniques to achieve the desired circuit performance. The second and main aspect pres

  4. Het onzichtbare circuit

    NARCIS (Netherlands)

    Nauta, Bram

    2013-01-01

    De chip, of geïntegreerde schakeling, heeft in een razend tempo ons leven ingrijpend veranderd. Het lijkt zo vanzelfsprekend dat er weer een nieuwe generatie smartphones, tablets of computers is. Maar dat is het niet. Prof.dr.ir. Bram Nauta, hoogleraar Integrated Circuit Design, laat in zijn rede

  5. Closed Circuit Videoinstallationen

    DEFF Research Database (Denmark)

    Kacunko, Slavko

    be seen with only insignificant qualification as a specific characteristic of the medium. The closed-circuit video installations based on it represent the attest field of experiment for the assumptions on art and the theory and history of the medium that it might lead one make. In recent years...

  6. Streaming Reduction Circuit

    NARCIS (Netherlands)

    Gerards, Marco Egbertus Theodorus; Kuper, Jan; Kokkeler, Andre B.J.; Molenkamp, Egbert

    2009-01-01

    Reduction circuits are used to reduce rows of floating point values to single values. Binary floating point operators often have deep pipelines, which may cause hazards when many consecutive rows have to be reduced. We present an algorithm by which any number of consecutive rows of arbitrary lengths

  7. A Magnetic Circuit Demonstration.

    Science.gov (United States)

    Vanderkooy, John; Lowe, June

    1995-01-01

    Presents a demonstration designed to illustrate Faraday's, Ampere's, and Lenz's laws and to reinforce the concepts through the analysis of a two-loop magnetic circuit. Can be made dramatic and challenging for sophisticated students but is suitable for an introductory course in electricity and magnetism. (JRH)

  8. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

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

    Science.gov (United States)

    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.

  10. Investigation of circulating temperature fluctuations of the primary coolant in order to develop an enhanced MTC estimator for VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kiss, Sandor; Lipcsei, Sandor [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research - MTA

    2017-09-15

    Our aim was to develop a method based on noise diagnostics for the estimation of the moderator temperature coefficient of reactivity (MTC) for the Paks VVER-440 units in normal operation. The method requires determining core average neutron flux and temperature fluctuations. The circulation period of the primary coolant, transfer properties of the steam generators, as well as the source and the propagation of the temperature perturbations and the proportions of the perturbation components were investigated in order to estimate the feedback caused by the circulation of the primary coolant. Finally, after developing the new MTC estimator, determining its frequency range and optimal parameters, trends were produced based on an overall evaluation of measurements made with standard instrumentation during a one-year-long period at Paks NPP.

  11. A review of hydrodynamic instabilities and their relevance to mixing in molten fuel coolant interactions

    International Nuclear Information System (INIS)

    Fletcher, D.F.

    1984-03-01

    A review of the literature on Rayleigh-Taylor, Kelvin-Helmholtz and capillary instability is presented. The concept of Weber breakup is examined and found to involve a combination of the above instabilities. Sample calculations are given which show how these instabilities may contribute to the mixing of melt and coolant in a molten fuel coolant interaction. It is concluded that Rayleigh-Taylor instability is likely to be important as the melt falls into the coolant and that Kelvin-Helmholtz instability is likely to develop when significant vapour velocities occur. (author)

  12. Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

    Directory of Open Access Journals (Sweden)

    Navid Bozorgan

    2015-07-01

    Full Text Available Nanofluids are expected to be a promising coolant condidate in chemical processes for heat transfer system size reduction. This paper focuses on reducing the number of turns in a helical coil heat recovery exchanger with a given heat exchange capacity in a biomass heating plant using γ-Al2O3/n-decane nanofluid as coolant. The nanofluid flows through the tubes and the hot n-hexane flows through the shell. The numerical results show that using nanofluid as coolant in a helical coil heat exchanger can reduce the manufacturing cost of the heat exchanger and pumping power by reducing the number of turns of the coil.

  13. Lubrication analysis of the journal bearing in the main coolant pump of SMART

    International Nuclear Information System (INIS)

    Lee, J. S.; Park, J. S.; Kim, J. H.; Kim, J. I.; Jang, M. H.

    2000-01-01

    Special type journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel. The MCP bearings are lubricated with water without external lubricating oil supply. Long bearing with vertical grooves is designed with relatively large bearing clearance to accommodate the long shaft. Lubricational analysis method for journal bearing with vertical grooves in the main coolant pump of SMART is proposed, and lubricational characteristics of the bearings are examined in this paper

  14. Lubrication analysis of the thrust bearing in the main coolant pump of SMART

    International Nuclear Information System (INIS)

    Lee, J. S.; Park, J. S.; Kim, J. H.; Hur, H.; Kim, J. I.

    2001-01-01

    Thrust bearing and journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel and especially the MCP bearings are lubricated with water without external lubricating oil supply. Because axial load capacity of the thrust bearing can hardly meet requirement to acquire hydrodynamic or fluid film lubrication state, self-lubrication characteristics of silicon graphite meterials would be needed. Lubricational analysis method for thrust bearing for the main coolant pump of SMART is proposed, and lubricational characteristics of the bearing generated by solving the Reynolds equation are examined in this paper

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

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse

    2012-01-01

    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 threedimensional governing equations for the fluid flow and the heat...... heat sink configurations reduces the coolant pumping power in the system....

  16. Circuit simulation and physical implementation for a memristor-based colpitts oscillator

    OpenAIRE

    Hongmin Deng; Dongping Wang

    2017-01-01

    This paper implements two kinds of memristor-based colpitts oscillators, namely, the circuit where the memristor is added into the feedback network of the oscillator in parallel and series, respectively. First, a MULTISIM simulation circuit for the memristive colpitts oscillator is built, where an emulator constructed by some off-the-shelf components is utilized to replace the memristor. Then the physical system is implemented in terms of the MULTISIM simulation circuit. Circuit simulation an...

  17. Investigation of the compatibility between ADS target material with coolant

    International Nuclear Information System (INIS)

    Xu Yongli; Long Bin; Zhang Pingzhu

    2002-01-01

    The compatibility tests for the forging tungsten with coolants are performed in sodium at 500, 600 and 700 degree C, and in water at 100 degree C. The results show: a compact W x O y film is formed at the surface of the tungsten, its thickness is about 1 - 2 μm. There are also corrosion product Na x WO y on the surface of the specimens, the amount of Na x WO y dependents on the temperature and the oxygen content in the sodium. After test for more than 400 h, the matrix of tungsten is not attacked further by sodium and oxygen resulting from the protection of the W x O y film, the thickness of W x O y film and the weight loss become a constant; for the corrosion of the tungsten in water, a W x O y film at the specimen surface is formed at the beginning of the test, its thickness is about 0.8 μm; This film is porous and loose, and they peels off after test for more than 100 h. The new oxide film does not form again because of the lack of the oxygen, the weight loss of the specimens is near a constant

  18. Numerical simulation on coolant flow and heat transfer in core

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    To simulate the coolant flow and the heat transfer characteristics of a core, a computer code, THAPMA (Thermal Hydraulic Analysis Porous Medium Analysis) has been developed. In THAPMA code, conservation equations are based on a porous-medium formulation, which uses four parameters, i.e, volume porosity, directional surface porosity, distributed resistance, and distributed heat source (sink), to model the effects of fuel rods and other internal solid structures on flow and heat transfer. Because the scheme and the solution are very important in accuracy and speed of calculation, a new difference scheme (WSUC) has been used in the energy equation, and a modified PISO solution method have been employed to simulate the steady/transient states. The code has been proved reliable and can effectively solve the transient state problem by several numerical tests. According to the design of Qinshan NPP-II, the flow and heat transfer phenomena in reactor core have been numerically simulated. The distributions of the velocity and the temperature can provide a theoretical basis for core design and safety analysis

  19. A study on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tai; Choi, Y. D.; Choi, J. H.; Kim, T. J.; Jeong, K. C.; Kwon, S. W.; Kim, B. H.; Jeong, J. Y.; Park, J. H.; Kim, K. R.; Jo, B. R.

    1997-08-01

    A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H{sub 2} gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs.

  20. Evaluation of zinc addition to PWR primary coolant

    International Nuclear Information System (INIS)

    Pathania, R.; Yagnik, S.; Gold, R.E.; Dove, M.; Kolstad, E.

    1995-01-01

    Laboratory studies have shown that addition of zinc to a PWR environment reduces the general corrosion rates of materials in the primary system and delays the initiation of primary water stress corrosion cracking (PWSCC) in Alloy 600. Because of the potential benefits of zinc addition in reducing radiation fields and mitigating PWSCC of Alloy 600 a project was initiated to qualify zinc addition to a PWR. The objective of this work was to evaluate the effect of zinc addition on radiation fields, PWSCC of Alloy 600 and fuel cladding corrosion at the Farley-2 PWR. In order to provide an early warning of any potential adverse effects on the fuel cladding, corrosion studies were initiated at the Halden test reactor prior to zinc addition at Farley-2. This paper provides an overview of the scope of the zinc addition demonstration at Farley-2 and the fuel cladding corrosion tests at Halden. The zinc concentration in the Farley-2 coolant is approximately 40 ppb and that in Halden is 50 ppb. The paper presents initial results from these studies which are still in progress