WorldWideScience

Sample records for coolant circuit activity

  1. Primary coolant circuits in FBR type reactors

    International Nuclear Information System (INIS)

    Kutani, Masushiro.

    1985-01-01

    Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

  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. Coolant circuit water chemistry of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Tilky, Peter; Doma, Arpad

    1985-01-01

    The numerous advantages of the proper selection of water chemistry parameters including low corrosion rate of the structural materials, hence the low-level activity build-up, depositions, radiation doses were emphasized. Major characteristics of water chemistry applied to the primary coolant of pressurized water reactors including neutral, slightly basic and strong basic ones are discussed. Boric acid is widely used to control reactivity. Primary coolant water chemistry of WWER type reactors which is based on the addition of ammonia and potassium hydroxide to boric acid is compared with that of other reactors. The demineralization of the total condensate of the steam turbines became a general trend in the water chemistry of the secondary coolant circuits. (V.N.)

  4. Coolant clean-up system in the primary coolant circuit for nuclear reactor

    International Nuclear Information System (INIS)

    Saito, Michio.

    1981-01-01

    Purpose: To maintain the quality of coolants at a prescribed level by distillating coolants in the primary coolant circuit for a BWR type reactor to remove impurities therefrom, taking out the condensates from the top of the distillation column and extracting impurities in a concentrated state from the bottom. Constitution: Coolant water for cooling the core is recycled by a recycling pump by way of a recycling pipeway in a reactor. The coolants extracted from an extraction pipeway connected to the recycling pipeway are fed into a distillation column, where distillation is taken place. Impurities in the coolants, that is, in-core corrosion products, fission products generated in the reactor core, etc. are separated by the distillation, concentrated and solidified in the bottom of the distillation column. While on the other hand, condensates removed with the impurities, that is, coolants cleaned-up are recycled to the coolant water for cooling the reactor core. (Moriyama, K.)

  5. Method of decontaminating primary coolant circuits

    International Nuclear Information System (INIS)

    Ishibashi, Masaru; Sumi, Masao.

    1981-01-01

    Purpose: To eliminate hard contaminated layers as well as soft contaminated layers without injuring substrate materials, upon decontamination of radiation contaminated portions in equipments and pipeways constituting primary coolant circuits. Constitution: High pressure water from a high pressure pump is jetted out from the nozzle of a spray gun to the radiation contaminated portions in equipments, for example, to the surface of water chamber in a vapor evaporator. High pressure pure water or aqueous boric acid is jetted out from the periphery and boric oxide particles (of about 1 - 100 μ particle size) are jetted out from the center of the nozzle of the spray gun. The particles (blasting material) jetted out together with the high pressure water impinge on the contaminated surfaces to remove the contaminated layers. Upon impingement, the high pressure water acts as the shock absorber for the blasting material and, after the impingement, it flows down to the bottom of the water chamber, and the blasting material is dissolved in the high pressure water. (Horiuchi, T.)

  6. 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.)

  7. Radioactive corrosion products in circuit of fast reactor loop with dissociating coolant

    International Nuclear Information System (INIS)

    Dolgov, V.M.; Katanaev, A.O.

    1982-01-01

    The results of experimental investigation into depositions of radionuclides of corrosion origin on the surfaces of a reactor-in-pile loop facility with a dissociating coolant are presented. It is stated that the ratio of radionuclides in fixed depositions linearly decreases with decrease of the coolant temperature at the core-condenser section. The element composition of non-fixed compositions quantitatively and qualitatively differs from the composition of structural material, and it is more vividly displayed for the core-condenser section. The main mechanism of circuit contamination with radioactive corrosion products is substantiated: material corrosion in the zones of coolant phase transfer, their remove by the coolant in the core, deposition, activation and wash-out by the coolant from the core surfaces

  8. Application of damage function analysis to reactor coolant circuits

    International Nuclear Information System (INIS)

    MacDonald, D.D.

    2002-01-01

    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)

  9. 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)

  10. The electrochemistry of IGSCC mitigation in BWR coolant circuits

    International Nuclear Information System (INIS)

    Macdonald, D.D.

    2002-01-01

    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 2 , H 2 , H 2 O 2 ), external solution composition (concentrations of O 2 , H 2 O 2 , and H 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)

  11. 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)

  12. Filtering device for primary coolant circuits in BWR type reactors

    International Nuclear Information System (INIS)

    Tajima, Fumio; Yamamoto, Tetsuo.

    1985-01-01

    Purpose: To obtain a filtering device with a large filtering area and requiring less space. Constitution: A condensate inlet for introducing condensates to be filtered of primary coolant circuits, a filtrate exit, a backwash water exit and a bent tube are disposed to a container, and a plurality of hollow thread membrane modules are suspended in the container. The condensates are caused to flow through the condensate inlet, filtered through the hollow thread membrane and then discharged from the filtrate exit. When the filtering treatment is proceeded to some extent, since solid contents captured in the hollow thread membranes are accumulated, a differential pressure is produced between the condensate inlet and the filtrate exit. When the differential pressure reaches a predetermined value, the backwash is conducted to discharge the liquid cleaning wastes through the backwash exit. The bent tube disposed to the container body is used for water and air draining. The hollow thread membranes are formed with porous resin such as of polyethylene. (Kawakami, Y.)

  13. Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

    International Nuclear Information System (INIS)

    Liebig, E.

    1977-01-01

    A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

  14. 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

  15. Design on Hygrometry System of Primary Coolant Circuit of HTR-PM

    International Nuclear Information System (INIS)

    Sun Yanfei; Zhong Shuoping; Huang Xiaojin

    2014-01-01

    Helium is the primary coolant in HTR-PM. If vapor get into the helium in primary coolant circuit because of some special reasons, such as the broken of steam-generator tube, chemical reaction will take effect between the graphite in reactor core and vapor in primary coolant circuit, and the safety of the reactor operation will be influenced. So the humidity of the helium in primary coolant circuit is one key parameter of HTR-PM to be monitored in-line. Once the humidity is too high, trigger signal of turning off the reactor must be issued. The hygrometry system of HTR-PM is consisting of filter, cooler, hygrometry sensor, flow meter, and some valves and tube. Helium with temperature of 250℃ is lead into the hygrometry system from the outlet of the main helium blower. After measuring, the helium is re-injected back to the primary circuit. No helium loses in this processing, and no other pump is needed. Key factors and calculations in design on hygrometry system of HTR-PM are described. A sample instrument has been made. Results of experiments proves that this hygrometry system is suitable for monitoring the humidity of the primary coolant of HTR-PM. (author)

  16. 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

  17. Secondary coolant circuit operation tests: steam generator feedwater supply

    International Nuclear Information System (INIS)

    Beroux, M.

    1985-01-01

    No one important accident occurred during the start-up tests of the 1300MWe P4 series, concerning the feedwater system of steam generators (SG). This communication comments on some incidents, that the tests allowed to detect very soon and which had no consequences on the operation of units: 1) Water hammer in feedwater tubes, and incidents met in the emergency steam generator water supply circuit. The technological differences between SG 900 and 1300 are pointed out, and the measures taken to prevent this problem are presented. 2) Incidents met on the emergency feedwater supply circuit of steam generators; mechanical or functional modifications involved by these incidents [fr

  18. Helium impurities in a PNP-primary coolant circuit

    International Nuclear Information System (INIS)

    Reif, M.

    1981-01-01

    The concentration of impurities to be expected have been defined in consideration of recent findings concerning the rates of infiltration and formation and the reaction mechanisms of the impurity components in the circuit. The data obtained correspond with the requirements on the metallic high-temperature components as well as with the requirements of limited graphite corrosion. (DG) [de

  19. 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

  20. Simplified model of a PWR primary coolant circuit

    International Nuclear Information System (INIS)

    Souza, A.L. de; Faya, A.J.G.

    1988-01-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analysed by a nodal model. Average and hot channels are treated so that the bulk response of the core and DNBR can be evaluated. A Homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  1. 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

  2. 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.)

  3. 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)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  5. Q-factor of coolant flow in the primary circuit of NPP with pressurised water reactors

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Belikov, S.O.; Novikov, K.S.

    2011-01-01

    Systems of preoperational vibration dynamic monitoring in of WWER are presented. The results of measurements during commission of NPP with WWER are presented. The paper provides the result of the research, that estimation of coolant fluctuations caused by pulse perturbation of pressure in the primary circuit NPP. It is shown that results could be received at known value of a Q - factor of acoustical oscillatory system only. The research demonstrates the results of dependence of the sound speed from the mass steam content in the coolant flow thru reactor core. The worked out results can be used for identification of the reasons of abnormal growth of level of vibrations of fuel assembly, fuel rod, equipment and internals, and for forecasting the operation conditions which provide of vibration - acoustical resonances in the primary loop equipment. (author)

  6. Experience in vibro-acoustic control of primary coolant circuit aggregates

    International Nuclear Information System (INIS)

    Sedov, V.K.; Adamenkov, K.A.

    1977-01-01

    Fundamental principles and possibilities of vibro-acoustic control of the primary coolant circuit in nuclear power plants for detecting failures (slack parts, penetration of foreign bodies, crack formation, etc.) are presented. As a result of pressure and flow rate fluctuations such failures give rise to characteristic changes in apmplitude and frequency of vibration and technological noise from the different aggregates with respect to a 'calibration' spectrum taken in the intact state. Nature and location of the failures may be determined by statistical analysis of the signals recorded from pressure and acceleration gauges. Certain parts of the primary circuit are controlled, especially the main circulation pumps. Additionally, neutron noise has been measured in order to control the core insertions. The method is illustrated by means of measurements performed in the units 1 to 4 of the Novovoronezh nuclear power plant during start-up operation and continuous operation. (author)

  7. Experience in vibro-acoustic control of primary coolant circuit aggregates

    Energy Technology Data Exchange (ETDEWEB)

    Sedov, V K; Adamenkov, K A [Nuclear power plant Novo-Voronesh (USSR)

    1977-10-01

    Fundamental principles and possibilities of vibro-acoustic control of the primary coolant circuit in nuclear power plants for detecting failures (slack parts, penetration of foreign bodies, crack formation, etc.) are presented. As a result of pressure and flow rate fluctuations such failures give rise to characteristic changes in apmplitude and frequency of vibration and technological noise from the different aggregates with respect to a 'calibration' spectrum taken in the intact state. Nature and location of the failures may be determined by statistical analysis of the signals recorded from pressure and acceleration gauges. Certain parts of the primary circuit are controlled, especially the main circulation pumps. Additionally, neutron noise has been measured in order to control the core insertions. The method is illustrated by means of measurements performed in the units 1 to 4 of the Novovoronezh nuclear power plant during start-up operation and continuous operation.

  8. Fact and fiction in ECP measurement and control in boiling water reactor primary coolant circuits

    International Nuclear Information System (INIS)

    Macdonald, D.D.

    2005-01-01

    A review is presented of various electrochemical potentials, including the electrochemical corrosion potential (ECP), that are used in the mitigation of stress corrosion cracking in the primary coolant circuits of boiling water reactors (BWRs). Attention is paid to carefully defining each potential in terms of fundamental electrochemical concepts, so as to counter the confusion that has arisen due to the misuse of previously accepted terminology. A brief discussion is also included of reference electrodes and it is shown on the basis of experimental data that the use of a platinum redox sensor as a reference electrode in the monitoring of ECP in BWR primary coolant circuits is inappropriate and should be discouraged. If platinum is used as a reference electrode, because of extenuating circumstances (e.g., potential measurements in high dose regions in a reactor core), the onus must be placed on the user to demonstrate quantitatively that the electrode behaves as an equilibrium electrode under the specified conditions and/or that its potential is invariant with changes in the independent variables of the system. Preferably, a means should also be demonstrated of transferring the measured potential to the standard hydrogen electrode (SHE) scale. (orig.)

  9. Factors governing particulate corrosion product adhesion to surfaces in water reactor coolant circuits

    International Nuclear Information System (INIS)

    1979-03-01

    Gravity, van der Waals, magnetic, electrical double layer and hydrodynamic forces are considered as potential contributors to the adhesion of particulate corrosion products to surfaces in water reactor coolant circuits. These forces are renewed and evaluated, and the following are amongst the conclusions drawn; adequate theories are available to estimate the forces governing corrosion product particle adhesion to surfaces in single phase flow in water reactor coolant circuits. Some uncertainty is introduced by the geometry of real particle-surface systems. The major uncertainties are due to inadequate data on the Hamaker constant and the zeta potential for the relevant materials, water chemistry and radiation chemistry at 300 0 C; van der Waals force is dominant over the effect of gravity for particles smaller than about 100 m; quite modest zeta potentials, approximately 50mV, are capable of inhibiting particle deposition throughout the size range relevant to water reactors; for surfaces exposed to typical water reactor flow conditions, particles smaller than approximately 1 m will be stable against resuspension in the absence of electrical double layer repulsion; and the magnitude of the electrical double layer repulsion for a given potential depends on whether the interaction is assumed to occur at constant potential or constant change. (author)

  10. Low-activation lead coolant for advanced small modular NPP

    International Nuclear Information System (INIS)

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

    2001-01-01

    The purpose of the paper is in studying perspectives of a new heavy liquid metal coolant for a small fast reactor (FR) concept. To reduce the post irradiation activity of the coolant the using of lead isotope, Pb-206, instead of natural lead, Pb-nat, is offered. In this case the accumulation of such hazardous radionuclides, as Po-210, Bi-208, Bi-207, essentially decreases. The interval of the lead-206 coolant cost which does not exceed 20% of the overall FR cost is estimated. The possibility of lead-206 obtaining for FR needs with the centrifugal separation technique is pointed out. (author)

  11. 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.

  12. 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

  13. Studies of iodine adsorption and desorption on HTGR coolant circuit materials

    International Nuclear Information System (INIS)

    Osborne, M.F.; Compere, E.L.; de Nordwall, H.J.

    1976-04-01

    Safety studies of the HTGR system indicate that radioactive iodine, released from the fuel to the helium coolant, may pose a problem of concern if no attenuation of the amount of iodine released occurs in the coolant circuit. Since information on iodine behavior in this system was incomplete, iodine adsorption on HTGR materials was studied in vacuum as a function of iodine pressure and of adsorber temperature. Iodine coverages on Fe 3 O 4 and Cr 2 O 3 approached maxima of about 2 x 10 14 and 1 x 10 14 atoms/cm 2 , respectively, whereas the iodine coverage on graphite under similar conditions was found to be less by a factor of about 100. Iodine desorption from the same materials into vacuum or flowing helium was investigated, on a limited basis, as a function of iodine coverage, of adsorber temperature, and of dry vs wet helium. The rate of vacuum desorption from Fe 3 O 4 was related to the spectrum of energies of the adsorption sites. A small amount of water vapor in the helium enhanced desorption from iron powder but appeared to have less effect on desorption from the metal oxides

  14. Review of the OECD specialist meeting on continuous monitoring techniques for assuring coolant circuit integrity

    International Nuclear Information System (INIS)

    Thie, J.A.

    1986-01-01

    This article summarizes the OECD Specialist Meeting on Continuous Monitoring Techniques for Assuring Coolant Circuit Integrity held August 12-14, 1985, in London. The conference was organized by the Organization for Economic Cooperation and Development's (OECD's) Committee on the Safety for Nuclear Installations and hosted by Her Majesty's Nuclear Installation Inspectorate at King's College. Many other conferences have addressed analysis and inspection approaches to ensuring primary-system integrity, but the OECD meeting was structured to pay attention to the continuous monitoring approach - possibly the first conference to be so designed. The specific technologies represented were vibrations, noise (i.e., random fluctuations in signals), leaks, acoustic emission, and cyclic fatigue. Although water reactors dominate the papers, all reactor types were included. A diverse group of about 50 attendees from 11 countries participated, including representatives from utilities, suppliers, regulators, and researchers

  15. Actively controlling coolant-cooled cold plate configuration

    Science.gov (United States)

    Chainer, Timothy J.; Parida, Pritish R.

    2015-07-28

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

  16. Advances on the analysis of fast reactor core and coolant circuit structures

    International Nuclear Information System (INIS)

    Livolant, M.; Imazu, A.; Chang, Y.W.; Eggen, D.T.

    1989-01-01

    For the 10th SMiRT Conference, it has been decided to make general reviews of the accomplishments throughout the conferences. The aim of this paper is to make such a review in the field of fast reactor core and coolant circuit structures, which is now fully treated in division E. That was not true in the past: at the earliest conferences up to the 5th, the division E dealt with accidental studies among which the hypothetical core disruptive accident was the most important. So, to cover the subject from the first SMiRT to now, it has been necessary to search into all the past division in order to recover the studies fitting into the scope of the present division E. This has allowed a table showing the number of presented papers on the various topics at the SMiRT conferences to be set up (table I). Then, some significant topics have been studied in detail, highlighting the main accomplishments, but trying also to point out the shortcomings and the work still to be done, in view of the present state of art

  17. Thermal-hydraulic model of the primary coolant circuits for the full-scale training facility with WWER-1000

    International Nuclear Information System (INIS)

    Kroshilin, A.E.; Zhukavin, A.P.; Pryakhin, V.N.

    1992-01-01

    The mathematical model realized in the full-scale educational facility for NPP operator training is described. The RETACT computational complex providing real time process simulation for all regimes including the maximum credible accident is used for calculation of thermohydraulic parameters of the primary coolant circuits and steam generator under stationary and transient conditions. The two-velocity two-temperature model of one-dimensional steam-water flow containing uncondensed gases is realized in the program

  18. Assessment of the heat carrier movement in the primary coolant circuit by its own momentum

    International Nuclear Information System (INIS)

    Kadalev, Stoyan

    2014-01-01

    Highlights: • We model the heat carrier flow alteration after the circulation pump(s) stop. • The general mathematical model used is described in details. • The model is adapted and applied to a particular example research reactor. • Assessment is presented in detail, step by step with references. • The information provided is enough to apply calculations to another facility. - Abstract: In the presented paper is considered the approach to an assessment of the heat carrier flow alteration in the primary water–water reactor coolant circuit after the circulation pump(s) stop. This topic is highly relevant trough advanced and increased nuclear safety requirements because such a process is observed in case of black-out accident or damaged pump(s). The general mathematical model used is described; enabling preparation of this evaluation adapted and applied to a particular example facility namely a pool type research reactor. The factors influencing to the heat carrier movement by its own momentum are examined. The evaluation measures and includes the factors influencing the heat carrier flow rate from the moment the pump(s) stops down to a negligible value. Assessment is presented in detail, step by step and where needed with references to specific data and/or formulae from reference books to allow repetition of the calculations and/or apply to another facility. The calculations are presented utilizing all necessary data according to the design and technological documentation. No account is given to the pressure of the natural circulation caused by the residual heat generation in the fuel after the reactor scram system extinction of the fission reaction

  19. 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

  20. 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.

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

    International Nuclear Information System (INIS)

    Saunin, Yuri V.; Dobrotvorski, Alexander N.; Semenikhin, Alexander V.; Korolev, Alexander S.

    2017-01-01

    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.

  2. Draft of diagnostic techniques for primary coolant circuit facilities using control computer

    International Nuclear Information System (INIS)

    Suchy, R.; Procka, V.; Murin, V.; Rybarova, D.

    A method is proposed of in-service on-line diagnostics of primary circuit selected parts by means of a control computer. Computer processing will involve the measurements of neutron flux, pressure difference in pumps and in the core, and the vibrations of primary circuit mechanical parts. (H.S.)

  3. Development of an automated system for CANDU secondary coolant circuit chemistry control

    International Nuclear Information System (INIS)

    Dean, J.R.; Stewart, R.B.

    1978-04-01

    This report is a summary of work done to develop a means for automated control of the secondary coolant chemistry of CANDU 600 MW(e) power reactors using on-line analyzers and a minicomputer. The development work was carried out in cooperation with Saskatchewan Power Corporation at Estevan. Results and conclusions of the program are included, as are recommendations for a prototype installation in a domestic CANDU 600 MW steam generator. (author)

  4. Analytical and sampling problems in primary coolant circuits of PWR-type reactors

    International Nuclear Information System (INIS)

    Illy, H.

    1980-10-01

    Details of recent analytical methods on the analysis and sampling of a PWR primary coolant are given in the order as follows: sampling and preparation; analysis of the gases dissolved in the water; monitoring of radiating substances; checking of boric acid concentration which controls the reactivity. The bibliography of this work and directions for its use are published in a separate report: KFKI-80-48 (1980). (author)

  5. Evaluation of specific activity in the primary circuit of SMART-P

    International Nuclear Information System (INIS)

    Kim, Ah Young; Choi, Byung Seon; Kim, Seong Hoon; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    SMART-P is a soluble boron free reactor, and the ammonia is used as a pH reagent. The titanium alloy, which has a high corrosion resistance, is chosen as a steam generator tube material. Despite these design features to achieve the corrosion reduction, it is expected that SMART-P exhibits a relatively high specific activity in the coolant due to the lack of purification during the power operation. The main reason for the high specific activity is the activation and transportation of the corrosion products that released from the primary circuit surfaces. The objective of this work is to analyze the corrosion product activity in the primary circuit of SMART-P using a multi-region model, KORA. This model, which is incorporated with the mass and activity transport between the dissolved corrosion products in the coolant and the surface, describes the specific activity of corrosion products in coolant and on the surfaces according to the operation modes

  6. Integral forged pump casing for the primary coolant circuit of a nuclear reactor: Development in design, forging technology, and material

    International Nuclear Information System (INIS)

    Austel, W.; Korbe, H.

    1986-01-01

    Developments in the forging of large casings for primary circuit coolant pumps for light water reactors in Germany are demonstrated beginning with the multiple forging fabricated version and ending with the integral forged type. This version is the result of the joint efforts of the pump manufacturer and the forgemaster after a cost-gain evaluation and represents an optimum solution in view of its functional and economical performance and also considering the high requirements for mechanical-technological properties, including homogeneity of the material. The development from 22 NiMoCr 3 7/A 508 Class 2 to 20 MnMoNi 5 5/A 508 Class 3 and their optimization will be demonstrated. This development is based mainly on minimizing the sulfur content and on vacuum carbon deoxidation (VCD), which results in a reduction of the A-segregations, in improving fracture toughness and isotropy, and in the desired fine-grain structure

  7. Major activated corrosion products cobalt, silver and antimony in the primary coolant of PWR power plants

    International Nuclear Information System (INIS)

    Xu Mingxia

    2012-01-01

    The production of the major activated corrosion products such as cobalt, silver and antimony in the primary coolant of PWR power plants and the impacts on the increase of the dose rates caused by these corrosion products during the shutdown are described in the paper. Investigating the corrosion product behavior during the operation and shutdown periods aims at detecting the appearance of these radiological pollutants in the early time and searching relevant solutions that may enable eventually to decrease the dose rate. The solutions may include: Replacing critical material in the primary system's equipment and components, which contact with primary coolant circuit to possibly limit the source term, Elaborating strictly the specific chemical and shutdown procedure to optimize the purification capacity and to minimize the over-contaminations; Improving purification techniques according to the real operation circumstance, and limiting the impacts of these pollutants. It is obvious in the real practices that implementing appropriate solution will be benefit to decrease or limit the pollutants species like cobalt, silver and antimony. (author)

  8. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

    2017-06-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  9. Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

    International Nuclear Information System (INIS)

    Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav

    2017-01-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

  10. Effect of high-temperature filtration on impurity composition in the primary circuit coolant of power units with WWER-1000 reactors

    International Nuclear Information System (INIS)

    Efimov, A.A.; Moskvin, L.N.; Gusev, B.A.; Leont'ev, G.G.; Nekrest'yanov, S.N.

    1992-01-01

    The effects of high-temperature filtration on changes in dispersive, chemical, radioisotope and phase compositions of impurities in primary circuit coolant of NPP with the WWER-1000 reactor are studied. Special filters are used for the studies. The data obtained confirm the applicability of high-temperature filtration for purification of WWER reactor water and steam separators at NPPs with RBMK reactors

  11. 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....

  12. 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)

  13. Correlating activity incorporation with properties of oxide films formed on material samples exposed to BWR and PWR coolants in Finnish nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Sirkiae, P. [VTT Industrial Systems, Espoo (Finland); Buddas, T.; Halin, M.; Kvarnstroem, R.; Tompuri, K. [Fortum Power and Heat Oy, Loviisa Power Plant, Loviisa (Finland); Helin, M.; Muttilainen, E.; Reinvall, A. [Teollisuuden Voima Oy, Olkiluoto (Finland)

    2002-07-01

    The extent of activity incorporation on primary circuit surfaces in nuclear power plants is connected to the chemical composition of the coolant, to the corrosion behaviour of the material surfaces and to the structure and properties of oxide films formed on circuit surfaces due to corrosion. Possible changes in operational conditions may induce changes in the structure of the oxide films and thus in the rate of activity incorporation. To predict these changes, experimental correlations between water chemistry, oxide films and activity incorporation, as well as mechanistic understanding of the related phenomena need to be established. In order to do this, flow-through cells with material samples and facilities for high-temperature water chemistry monitoring have been installed at Olkiluoto unit 1 (BWR) and Loviisa unit 1 (PWR) in spring 2000. The cells are being used for two major purposes: To observe the changes in the structure and activity levels of oxide films formed on material samples exposed to the primary coolant. Correlating these observations with the abundant chemical and radiochemical data on coolant composition, dose rates etc. collected routinely by the plant, as well as with high-temperature water chemistry monitoring data such as the corrosion potentials of relevant material samples, the redox potential and the high-temperature conductivity of the primary coolant. We describe in this paper the scope of the work, give examples of the observations made and summarize the results on oxide films that have been obtained during one full fuel cycle at both plants. (authors)

  14. Active Trimming of Hybrid Integrated Circuits

    OpenAIRE

    Németh, P.; Krémer, P.

    1984-01-01

    One of the more important fields of the microelectronics industry is the manufacturing of hybrid integrated circuits.An important part of the manufacturing process is concerned with the trimming of the hybrid integratedl circuits. This article deals with the basic principles of active trimming and introduces a microprocessor controlled trimming machine. By comparing active trimming with passive techniques, it can be shown that the active system has some advantages. This article outlines these...

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

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2006-01-01

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

  16. 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)

  17. Estimation of aluminum and argon activation sources in the HANARO coolant

    International Nuclear Information System (INIS)

    Jun, Byung Jin; Lee, Byung Chul; Kim, Myong Seop

    2010-01-01

    The activation products of aluminum and argon are key radionuclides for operational and environmental radiological safety during the normal operation of open-tank-in-pool type research reactors using aluminum-clad fuels. Their activities measured in the primary coolant and pool surface water of HANARO have been consistent. We estimated their sources from the measured activities and then compared these values with their production rates obtained by a core calculation. For each aluminum activation product, an equivalent aluminum thickness (EAT) in which its production rate is identical to its release rate into the coolant is determined. For the argon activation calculation, the saturated argon concentration in the water at the temperature of the pool surface is assumed. The EATs are 5680, 266 and 1.2 nm, respectively, for Na-24, Mg-27 and Al-28, which are much larger than the flight lengths of the respective recoil nuclides. These values coincide with the water solubility levels and with the half-lives. The EAT for Na-24 is similar to the average oxide layer thickness (OLT) of fuel cladding as well; hence, the majority of them in the oxide layer may be released to the coolant. However, while the average OLT clearly increases with the fuel burn-up during an operation cycle, its effect on the pool-top radiation is not distinguishable. The source of Ar-41 is in good agreement with the calculated reaction rate of Ar-40 dissolved in the coolant

  18. Computer programmes of the Power Research Institute for the analysis of processes in the primary coolant circuit and in the containment of a WWER plant in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Misak, J.

    1976-01-01

    A brief description is given of computer programmes for the analysis of loss-of-coolant accidents (LOCA) in WWER type reactors. The LENKA programme is intended for the thermal and hydraulic analysis of the consequences of such accidents in the primary coolant circuit. The SICHTA programme is intended for the detailed calculation of the time dependence of the axial and radial distribution of heat in fuel rods from steady-state to the flooding of the core. CHEMLOC is intended for the analysis of the heat history of the core and the extent of chemical reactions in LOCA when the emergency core cooling system is not operating. The TRACO I is intended for the analysis of the initial stage of the transient process in a full-pressure containment after LOCA (the computation of the time and spatial dependences of pressures and temperatures). TRACO III is intended for the computation of the long-term time dependence of pressure and temperature in the full-pressure containment after LOCA. (B.S.)

  19. Evaluation of the corrosion, reactivity and chemistry control aspects for the selection of an alternative coolant in the secondary circuit of sodium fast reactors

    International Nuclear Information System (INIS)

    Brissonneau, L.; Simon, N.; Balbaud-Celerier, F.; Courouau, J.L.; Martinelli, L.; Grabon, V.; Capitaine, A.; Conocar, O.; Blat, M.

    2009-01-01

    Full text of publication follows: Sodium Fast Reactors are promising fourth generation reactors as they can contribute to reduce resource demand in uranium and considerably reduce waste level due to their fast spectrum. However, progress can be obtained for these reactors on the investment cost and on safety improvement. To achieve these goals, one of the innovative solutions consists in eliminating the reaction of sodium with water in the steam generators, by replacing the sodium in the secondary circuit by another coolant. A work group composed of experts from CEA, Areva NP and EdF was in charge to evaluate several alternative coolants as Heavy Liquid Metals (HLM), nitrate salts and hydroxide mixtures, through a multi-criteria analysis. Three important criteria for the selection of one coolant are its 'Interactions with the structures', and its 'chemistry control', and 'Reactivity with fluids' which are strongly correlated. The assessment, mainly based on the state-of-art from published literature on these points, is detailed in this paper. The mechanisms of corrosion of steels by the HLM depend on the oxygen content. For Pb-Bi, it has been modelled for oxidation and release domains. The corrosion of steels by nitrate salts presents similarity with the oxidation induced by HLM. The highly corrosive hydroxide mixture requires the use of nickel base alloys, for which oxidation and mass transfer are nevertheless significant. The HLM requires a fine regulation of oxygen content, through measurements and control systems, both to prevent lead oxide precipitation at high level and release corrosion at low level. Nitrate salts decompose into nitrites at sufficiently high temperature, which might induce pressure build-up in the circuit. The hydroxides must be kept under reducing atmosphere to lower the corrosion rate. Though these coolants are relatively inert to air and water, one of the main drawbacks of HLM and nitrate salts are their reactivity with sodium. Bismuth

  20. The application of release models to the interpretation of rare gas coolant activities

    International Nuclear Information System (INIS)

    Wise, C.

    1985-01-01

    Much research is carried out into the release of fission products from UO 2 fuel and from failed pins. A significant application of this data is to define models of release which can be used to interpret measured coolant activities of rare gas isotopes. Such interpretation is necessary to extract operationally relevant parameters, such as the number and size of failures in the core and the 131 I that might be released during depressurization faults. The latter figure forms part of the safety case for all operating CAGRs. This paper describes and justifies the models which are used in the ANAGRAM program to interpret CAGR coolant activities, highlighting any remaining uncertainties. The various methods by which the program can extract relevant information from the measurements are outlined, and examples are given of the analysis of coolant data. These analyses point to a generally well understood picture of fission gas release from low temperature failures. Areas of higher temperature release are identified where further research would be beneficial to coolant activity analysis. (author)

  1. ACFA - a versatile activation code for coolant and structural materials

    International Nuclear Information System (INIS)

    Brockmann, H.; Ohlig, U.

    1983-09-01

    The ACFA code calculates the neutron-induced activation, afterheat, transmutation, gas production, biological hazard potential, and activation gamma ray spectra in the components of a nuclear system. The quantities of interest may be computed by spatial interval and zone or only by zone of the system considered. To calculate the transmutation coefficients for the neutron-induced reactions the code uses multigroup activation cross sections and space-dependent multigroup neutron fluxes in one- or two-dimensional geometry. The neutron reaction types incorporated in the code are: (n,n'), (n,2n), (n,γ), (n,p), (n,α), (n,n'p), (n,n'α)sub(,) (n,t), (n,3n), (n,He-3), (n,d), and (n,n'd) considering both reactions to the ground state and to isomeric states. The code uses a variable dimensioning technique to adapt the core data storage requirements to the particular problem considered and uses the FIDO input system to read the input data. The numerical methods for establishing and solving the decay chain equations are taken from the ORIGEN code. To test the ACFA code and the nuclear data libraries used, the activation, composition change, and gas production in the first wall of the UWMAK-I fusion reactor are calculated. The results of the activation calculation are compared with earlier results of the University of Wisconsin Fusion Study Group. (orig.)

  2. Evaluation of Specific Activity in the Primary Coolant of PWRs by using SAEP

    International Nuclear Information System (INIS)

    Kim, Ha Yong; Song, Jae Seung; Kim, Keung Ku; Kim, Kyo Youn

    2008-07-01

    SAEP(Specific Activity Evaluation Program) to evaluate specific activities in the primary coolant of reactors due to fission products has been developed, which can be applied to the new concept nuclear reactor such as SMART as well as commercial PWRs in existence. Specific activities in the primary coolant were evaluated by using SAEP against reactor plants which are being operated currently in South Korea, respectively. We study the possibility of being applied to the developing commercial PWRs and the new concept reactors through the comparison the results by using SAEP with the results mentioned in the FSARs. We also verify SAEP itself through this evaluation. From the evaluation results, we know that the general trend is agreed with each other from the viewpoint of order of magnitude and that SAEP correctly executes the evaluation of specific activities in the primary coolant of reactor due to fission products for several reactor types, regardless of a reactor type. Therefore, SAEP can widely be applied to the new concept nuclear reactor development phase as well as already developed PWRs

  3. The experimental definition of the acoustic standing wave series shapes, formed in the coolant of the primary circuit of VVER-440 type reactor

    International Nuclear Information System (INIS)

    Bulavin, V.V.; Pavelko, V.I.

    1995-01-01

    On the basis of pressure fluctuation measurements in some primary circuit loops at 2 nd Unit of Kola NPP with VVER-440 type reactors, the shapes of acoustic standing waves (ASW) were determined at frequencies corresponding to four minimal oscillation eigenfrequencies in the primary circuit coolant. On identification of the ASW modes and properties, experimental results based on six circulating loops in symmetric arrangement allowed determination of the three-dimensional space structure of the wave nodes and antinodes inside and outside of the reactor vessel (RV). As part of this analysis, the geometric features of the primary circuit that caused the formation of these standing waves were identified. Differences in each ASW shape were shown to cause different individual effects on the neutron field in the reactor core and on fuel assembly vibration. This has been partially confirmed by ex-core neutron ionization chamber noise analysis. One type of ASW, possessing an antinode inside the RV, can be used for measurement of the pressure coefficient of reactivity. However, this must be done with care to avoid the potential for incorrect results in some cases. The results presented in this paper can be readily extended to other VVER type reactors with both odd and even number of loops. (author)

  4. 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.

  5. Liquid metal coolant disposal from UKAEA reactors at Dounreay

    International Nuclear Information System (INIS)

    Adam, E.R.

    1997-01-01

    As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)

  6. 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.

  7. Measurement of vibrations in the primary coolant circuit and in the vertical experimental channel of the RA reactor

    International Nuclear Information System (INIS)

    Ristic, B.; Rakic, R.; Milosevic, M.; Jerkovic, M.

    1966-01-01

    Full text: Beginning of the work dates from 1962 with the initial objective: study of the wear-out of the bearings of the centrifugal pumps in the heavy water system. It has been expected that the increase of wear-out would initiate increase of vibration amplitudes and noise. During further study the initial task was broadened to other fields, mainly appearance of material fatigue in components of the heavy water coolant system. During operation mechanical energy is generated due to non existing equilibrium of the pump rotor, wear-out of the bearing, turbulence in the pump, cavitation process and pulsation of the operating environment. This energy is transformed into noise and vibration energy which is spread through surrounding walls and pipes causing noise finally. Obtained results were only qualitatively tested at present. For quantitative testing it would be necessary to obtain data about the material, in addition to the diagrams obtained by measurements. It would be possible to calculate the fatigue of the material at measuring points as well as estimation of the time when material fatigue would become critical [sr

  8. Calculated model of radioactive fission and corrosion product accumulation and distribution in a fast reactor sodium coolant circuit

    International Nuclear Information System (INIS)

    Kizin, V.D.; Konyashov, V.V.

    1987-01-01

    A simple calculation procedure of radioactive products accumulation and distribution in a primary circuit has been developed on the basis of experimental investigations at the BOR-60 reactor. Common knowledge on the impurity products transfer at the liquid-solid and liquid-gas phase boundary is taken. Use is made of the typical in reactor physics relationships for the description of the products transition to the equipment surfaces, of fission products release, metal corrosion and others. Satisfactory agreement of the calculation data with the experimental ones has been obtained. (orig.)

  9. Improving the ALUeS diagnostic system for determining the coolant leak place from the WWER-440 primary circuit

    International Nuclear Information System (INIS)

    Markosyan, G.R.; Petrosyan, V.G.; Shakhverdyan, S.V.; Aslanyan, M.A.

    2000-01-01

    The new algorithm for localizing the leakage from the WWER-440 primary circuit, intended for operation in the Siemens ALUeS system, is proposed. The results of the algorithm realization in the leakage control system (the ALUeS system copy), installed at the Armenian NPP power unit-2, are presented. The leakage localization algorithm proposed was tested in other experiments. The leakage position in the majority of cases is determined exactly. Small (up to 5 m) deviations, the cause whereof were incorrect readings of the transducers, were observed [ru

  10. Spectral analysis of coolant activity from a commercial nuclear generating station

    International Nuclear Information System (INIS)

    Swann, J.D.; Lewis, B.J.; Ip, M.

    2008-01-01

    In support of the development of a real-time on-line fuel failure monitoring system for the CANDU reactor, actual gamma spectroscopy data files from the gaseous fission product (GFP) monitoring system were acquired from almost four years of operation at a commercial Nuclear Generating Station (NGS). Several spectral analysis techniques were used to process the data files. Radioisotopic activity from the plant information (PI) system was compared to an in-house C++ code that was used to determine the photopeak area and to a separate analysis with commercial software from Canberra-Aptec. These various techniques provided for a calculation of the coolant activity concentration of the noble gas and iodine species in the primary heat transport system. These data were then used to benchmark the Visual DETECT code, a user friendly software tool which can be used to characterize the defective fuel state based on a coolant activity analysis. Acceptable agreement was found with the spectral techniques when compared to the known defective bundle history at the commercial reactor. A more generalized method of assessing the fission product release data was also considered with the development of a pre-processor to evaluate the radioisotopic release rate from mass balance considerations. The release rate provided a more efficient means to characterize the occurrence of a defect and was consistent with the actual defect situation at the power plant as determined from in-bay examination of discharged fuel bundles. (author)

  11. Research on coolant radiochemistry

    International Nuclear Information System (INIS)

    Yeon, Jei Won; Kim, W. H.; Park, Y. J.; Im, J. K.; Jung, Y. J.; Jee, K. Y.; Choi, K. C.

    2004-04-01

    The final objective of this study is to develop the technology on the reduction of radioactive material formed in reactor coolant circuit. The contents of this study are composed of the simulation of primary cooling system, chemistry measurement technology in the high-temperature high-pressure environments, and coolant chemistry control technology. The main results are as follows; High-temperature and high-pressure loop system was designed and fabricated, which is to inducing CRUD growth condition on the surface of cladding. The high-temperature pH measurement system was established with YSZ sensing electrode and Ag/AgCl reference electrode. The performance of pH electrode was confirmed in the temperature range 200∼280 .deg. C. Coolant chemistry control technologies such as the neutron irradiation technique of boric acid solution, the evaluation on high-temperature electrochemical behavior of coolant, and the measurement of physicochemical properties of micro-particles were developed. The results of this study can be useful for the understanding of chemical phenomena occurred in reactor coolant and for the study on the reduction of radioactive material in primary coolant, which will be carried out in the next research stage

  12. An Activity for Demonstrating the Concept of a Neural Circuit

    Science.gov (United States)

    Kreiner, David S.

    2012-01-01

    College students in two sections of a general psychology course participated in a demonstration of a simple neural circuit. The activity was based on a neural circuit that Jeffress proposed for localizing sounds. Students in one section responded to a questionnaire prior to participating in the activity, while students in the other section…

  13. BR-5 primary circuit decontamination

    International Nuclear Information System (INIS)

    Efimov, I.A.; Nikulin, M.P.; Smirnov-Averin, A.P.; Tymosh, B.S.; Shereshkov, V.S.

    1976-01-01

    Results and methodology of steam-water and acid decontamination of the primary coolant circuit SBR-5 reactor in 1971 are discussed. Regeneration process in a cold trap of the primary coolant circuit is discussed

  14. Reactor coolant cleanup device

    International Nuclear Information System (INIS)

    Igarashi, Noboru.

    1986-01-01

    Purpose: To enable to introduce reactor water at high temperature and high pressure as it is, as well as effectively adsorb to eliminate cobalt in reactor water. Constitution: The coolant cleanup device comprises a vessel main body inserted to coolant pipeway circuits in a water cooled reactor power plant and filters contained within the vessel main body. The filters are prepared by coating and baking powder of metal oxides such as manganese ferrite having a function capable of adsorbing cobalt in the coolants onto the surface of supports made of metals or ceramics resistant to strong acids and alkalies in the form of three-dimensional network structure, for example, zircaloy-2, SUS 303 and the zirconia (baking) to form a basic filter elements. The basic filter elements are charged in plurality to the vessel main body. (Kawaiami, Y.)

  15. Normalizing the maximum permissible seal failure of the fuel cladding of VVER and the activity of the fission products in the coolant

    International Nuclear Information System (INIS)

    Luzanova, L.M.; Miglo, V.N.; Slavyagin, P.D.

    1993-01-01

    In most countries developing a nuclear power industry based on pressurized water reactors, one of the conditions for issuing a license under normal operating conditions for issuing a license stipulates that the fuel elements may not lose their hermetic seal either under normal operating conditions or during presumable disturbances of the conditions of normal use. At a conference on radiation safety the ALARA principle was taken to be fundamental, it being attempted to keep the activity of the coolant of the primary circuit, including the fission products emerging from unsealed fuel elements, to a level as low as reasonably possible. As many years of experience in the nuclear power industry have shown, nuclear power stations are in many cases operated with nonhermetic fuel elements in the core. Therefore, from the point of view of safety and economy, the best way to operate a power plant is to try to ensure maximum burnup of the fuel of the unsealed elements as they operate within the limits of safe activity of the fission products in the fuel circuits

  16. Research on Coolant Radiochemistry

    International Nuclear Information System (INIS)

    Ha, Yeong Keong; Kim, W. H.; Yeon, J. W.; Jung, Y. J.; Choi, K. C.; Choi, K. S.; Park, Y. J.; Cho, Y. H.

    2007-06-01

    The final objective of this study is to develop a method for reducing radioactive materials formed in the reactor coolant circuit. This second stage research was categorized into the following three subgroups: the development of the estimation technique of microscopic chemical variation at high temperatures and pressures, the fundamental study on the thermodynamics at high temperatures and pressures, and the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD. First, in the development of the estimation technique of microscopic chemical change at high temperatures and pressures, the technique for measuring coolant chemistry such as pH, conductivity and Eh was developed to be appropriate for the high temperature and pressure condition. The coolant chemistry measuring system including the self-devised high temperature pH sensor can be applied to the field of nuclear reactor and contribute on a large scale in the automation of the coolant chemistry control and the establishment of the real-time on-line measuring technique. Secondly, the dissociation constant of water and the solubility of metal oxides were measured in the fundamental study on the thermodynamics at high temperatures and pressures. Finally, in the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD, the careful investigation of the deposition phenomena of micro particles on the cladding surface showed that subcooled boiling and the dissolved hydrogen are the main factors responsible for the growth of CRUD. In addition, the basis was provided for the construction of a new particle behavior model in the reactor coolant circuit

  17. Measurements of the Effects of Smoke on Active Circuits

    International Nuclear Information System (INIS)

    Tanaka, T.J.

    1999-01-01

    Smoke has long been recognized as the most common source of fire damage to electrical equipment; however, most failures have been analyzed after the fire was out and the smoke vented. The effects caused while the smoke is still in the air have not been explored. Such effects have implications for new digital equipment being installed in nuclear reactors. The U.S. Nuclear Regulatory Commission is sponsoring work to determine the impact of smoke on digital instrumentation and control. As part of this program, Sandia National Laboratories has tested simple active circuits to determine how smoke affects them. These tests included the study of three possible failure modes on a functional board: (1) circuit bridging, (2) corrosion (metal loss), and (3) induction of stray capacitance. The performance of nine different circuits was measured continuously on bare and conformably coated boards during smoke exposures lasting 1 hour each and continued for 24 hours after the exposure started. The circuit that was most affected by smoke (100% change in measured values) was the one most sensitive to circuit bridging. Its high impedance (50 MOmega) was shorted during the exposure, but in some cases recovered after the smoke was vented. The other two failure modes, corrosion and induced stray capacitance, caused little change in the function of the circuits. The smoke permanently increased resistance of the circuit tested for corrosion, implying that the cent acts were corroded. However, the change was very small (< 2%). The stray-capacitance test circuit showed very little change after a smoke exposure in either the short or long term. The results of the tests suggest that conformal coatings and type of circuit are major considerations when designing digital circuitry to be used in critical control systems

  18. Measurements of the effects of smoke on active circuits

    International Nuclear Information System (INIS)

    Tanaka, T.J.

    1998-01-01

    Smoke has long been recognized as the most common source of fire damage to electrical equipment; however, most failures have been analyzed after the fire was out and the smoke vented. The effects caused while the smoke is still in the air have not been explored. Such effects have implications for new digital equipment being installed in nuclear reactors. The US Nuclear Regulatory Commission is sponsoring work to determine the impact of smoke on digital instrumentation and control. As part of this program, Sandia National Laboratories has tested simple active circuits to determine how smoke affects them. These tests included the study of three possible failure modes on a functional board: (1) circuit bridging, (2) corrosion (metal loss), and (3) induction of stray capacitance. The performance of nine different circuits was measured continuously on bare and conformally coated boards during smoke exposures lasting 1 hour each and continued for 24 hours after the exposure started. The circuit that was most affected by smoke (100% change in measured values) was the one most sensitive to circuit bridging. Its high impedance (50 Mohm) was shorted during the exposure, but in some cases recovered after the smoke was vented. The other two failure modes, corrosion and induced stray capacitance, caused little change in the function of the circuits. The smoke permanently increased resistance of the circuit tested for corrosion, implying that the contacts were corroded. However, the change was very small (< 2%). The stray capacitance test circuit showed very little change after a smoke exposure in either the short or long term. The results of the tests suggest that conformal coatings and type of circuit are major considerations when designing digital circuitry to be used in critical control systems

  19. Activity-regulated genes as mediators of neural circuit plasticity.

    Science.gov (United States)

    Leslie, Jennifer H; Nedivi, Elly

    2011-08-01

    Modifications of neuronal circuits allow the brain to adapt and change with experience. This plasticity manifests during development and throughout life, and can be remarkably long lasting. Evidence has linked activity-regulated gene expression to the long-term structural and electrophysiological adaptations that take place during developmental critical periods, learning and memory, and alterations to sensory map representations in the adult. In all these cases, the cellular response to neuronal activity integrates multiple tightly coordinated mechanisms to precisely orchestrate long-lasting, functional and structural changes in brain circuits. Experience-dependent plasticity is triggered when neuronal excitation activates cellular signaling pathways from the synapse to the nucleus that initiate new programs of gene expression. The protein products of activity-regulated genes then work via a diverse array of cellular mechanisms to modify neuronal functional properties. Synaptic strengthening or weakening can reweight existing circuit connections, while structural changes including synapse addition and elimination create new connections. Posttranscriptional regulatory mechanisms, often also dependent on activity, further modulate activity-regulated gene transcript and protein function. Thus, activity-regulated genes implement varied forms of structural and functional plasticity to fine-tune brain circuit wiring. Copyright © 2011 Elsevier Ltd. All rights reserved.

  20. Neuronal Population Activity in Spinal Motor Circuits

    DEFF Research Database (Denmark)

    Berg, Rune W.

    2017-01-01

    The core elements of stereotypical movements such as locomotion, scratching and breathing are generated by networks in the lower brainstem and the spinal cord. Ensemble activities in spinal motor networks had until recently been merely a black box, but with the emergence of ultra-thin Silicon multi......-electrode technology it was possible to reveal the spiking activity of larger parts of the network. A series of experiments revealed unexpected features of spinal networks, such as multiple spiking regimes and lognormal firing rate distributions. The lognormality renders the widespread idea of a typical firing rate...

  1. The atmospheric electric global circuit. [thunderstorm activity

    Science.gov (United States)

    Kasemir, H. W.

    1979-01-01

    The hypothesis that world thunderstorm activity represents the generator for the atmospheric electric current flow in the earth atmosphere between ground and the ionosphere is based on a close correlation between the magnitude and the diurnal variation of the supply current (thunderstorm generator current) and the load current (fair weather air-earth current density integrated over the earth surface). The advantages of using lightning survey satellites to furnish a base for accepting or rejecting the thunderstorm generator hypothesis are discussed.

  2. Active Match Load Circuit Intended for Testing Piezoelectric Transformers

    DEFF Research Database (Denmark)

    Andersen, Thomas; Rødgaard, Martin Schøler; Andersen, Michael A. E.

    2012-01-01

    An adjustable high voltage active load circuit for voltage amplitudes above 100 volts, especially intended for resistive matching the output impedance of a piezoelectric transformer (PT) is proposed in this paper. PTs have been around for over 50 years, were C. A. Rosen is common known for his...

  3. Activity-dependent modulation of neural circuit synaptic connectivity

    Directory of Open Access Journals (Sweden)

    Charles R Tessier

    2009-07-01

    Full Text Available In many nervous systems, the establishment of neural circuits is known to proceed via a two-stage process; 1 early, activity-independent wiring to produce a rough map characterized by excessive synaptic connections, and 2 subsequent, use-dependent pruning to eliminate inappropriate connections and reinforce maintained synapses. In invertebrates, however, evidence of the activity-dependent phase of synaptic refinement has been elusive, and the dogma has long been that invertebrate circuits are “hard-wired” in a purely activity-independent manner. This conclusion has been challenged recently through the use of new transgenic tools employed in the powerful Drosophila system, which have allowed unprecedented temporal control and single neuron imaging resolution. These recent studies reveal that activity-dependent mechanisms are indeed required to refine circuit maps in Drosophila during precise, restricted windows of late-phase development. Such mechanisms of circuit refinement may be key to understanding a number of human neurological diseases, including developmental disorders such as Fragile X syndrome (FXS and autism, which are hypothesized to result from defects in synaptic connectivity and activity-dependent circuit function. This review focuses on our current understanding of activity-dependent synaptic connectivity in Drosophila, primarily through analyzing the role of the fragile X mental retardation protein (FMRP in the Drosophila FXS disease model. The particular emphasis of this review is on the expanding array of new genetically-encoded tools that are allowing cellular events and molecular players to be dissected with ever greater precision and detail.

  4. Fundamentals for the development of a low-activation lead coolant with isotopic enrichment for advanced nuclear power facilities

    International Nuclear Information System (INIS)

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

    2002-01-01

    The purpose of this paper is to study the prospects of new coolants for fast reactors and accelerator driven systems. The main focus is on their improvement using the isotopic tailoring technique to reduce post-irradiation activity. Calculations using the FISPACT-3 code show that irradiating natural lead (Pb-nat) for 30 years leads to the accumulation of long-lived toxic radionuclides, 207 Bi, 208 Bi and 210 Pb, which extends the cooling down period to the clearance level. This time can be shortened by using the lead isotope 206 Pb instead of Pb-nat. This substantially decreases the concentration of the most toxic polonium isotope, 210 Po. Calculations for lead activation in the hard proton-neutron ADS spectrum were performed using the CASCADE/SNT code. The time-dependent activity of the 207 Bi produced in Pb-nat and 206 Pb after irradiation for one year with a proton beam having an energy of 0.8 GeV and a current of 30 mA is given. The activity of 207 Bi is decreased by four orders of magnitude when 206 Pb is used instead of natural lead as a coolant for ADS targets. The production of such radiotoxic nuclides as 210 Po is also substantially diminished. (author)

  5. 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

  6. Active filter for the DESY III dipole circuit

    International Nuclear Information System (INIS)

    Bothe, W.

    1991-01-01

    The DESY 3 dipole circuit is now operated in a ramp mode cycle with 3.6 s repetition rate. Excitation is done by a 12-pulse thyristor converter, followed by a passive filter. The existing current control could be improved by addition of an active filter. The use of a more efficient passive filter reduces the size of the active filter and does not deteriorate the dynamic behavior. The design of the control loops and the results of the simulation are presented

  7. Estimation of activity in primary coolant heat exchanger of Apsara reactor after 50 years of reactor operation

    International Nuclear Information System (INIS)

    Prasad, S.K.; Anilkumar, S.; Vajpayee, L.K.; Belhe, M.S.; Yadav, R.K.B.; Deolekar, S.S.

    2012-01-01

    The primary coolant heat exchanger of Apsara Reactor was in operation for 53 years and as a part of partial decommissioning of Apsara Primary Coolant Heat Exchanger (PHEx) was decommissioned and disposed off as active waste. The long lived component deposited in the SS tubes inside the heat exchanger was assessed by taking the scrape samples and in situ gamma spectrometry technique employing NaI(Tl) detector. The data obtained by experimental measurements were validated by Monte Carlo simulation method. From the present studies, it was shown that 137 Cs and 144 Ce as the major isotopes deposited on the SS tube of heat exchanger. In this paper the authors describes the details of the methodology adopted for the assessment of radioactivity content and the results obtained. This give a reliable method to estimate the activity disposed for waste management accounting purpose in a long and heavy reactor component. The upper bound of total activity in PHEx 39.0μCi. (author)

  8. Profiling bacterial kinase activity using a genetic circuit

    DEFF Research Database (Denmark)

    van der Helm, Eric; Bech, Rasmus; Lehning, Christina Eva

    Phosphorylation is a post-translational modification that regulates the activity of several key proteins in bacteria and eukaryotes. Accordingly, a variety of tools has been developed to measure kinase activity. To couple phosphorylation to an in vivo fluorescent readout we used the Bacillus...... subtilis kinase PtkA, transmembrane activator TkmA and the repressor FatR to construct a genetic circuit in E. coli. By tuning the repressor and kinase expression level at the same time, we were able to show a 4.2-fold increase in signal upon kinase induction. We furthermore validated that the previously...... reported FatR Y45E mutation1 attenuates operator repression. This genetic circuit provides a starting point for computational protein design and a metagenomic library-screening tool....

  9. Evaluation of an optimized coolant circuit conception in a thermal whole vehicle environment with respect to the consumption of primary energy; Bewertung eines optimierten Kuehlmittelkreislaufkonzeptes in einer thermischen Gesamtfahrzeugumgebung hinsichtlich des Primaerenergieverbrauchs

    Energy Technology Data Exchange (ETDEWEB)

    Schulze, Mirko; Neumann, Alexander; Tilch, Benjamin; Eilts, Peter [Technische Univ. Braunschweig (Germany). Inst. fuer Verbrennungskraftmaschinen; Niedersaechsisches Forschungszentrum Fahrzeugtechnik (NFF), Braunschweig (Germany)

    2012-11-01

    This work deals with a co-simulation vehicle environment developed by the institute of internal combustion engines (ivb) of the Technical University Braunschweig as a tool to analyze the thermal effects in the power train during the warm-up phase, especially on the fuel consumption. This allows evaluating new drive train concepts in early stages of development by using power train thermal management techniques (TMM). Therefore you are able to give an objective statement for these techniques by analyzing the changes in fuel consumption. The used simulation models will be introduced and the mechanical and thermal behavior is verified using test bench data. An optimized coolant circuit concept in GT Suite {sup registered}, developed at the institute is identified and coupled to a thermal engine model. In this paper, the potentials for reducing primary energy consumption in the New European Driving Cycle (NEDC) are presented. (orig.)

  10. Using perturbations to identify the brain circuits underlying active vision.

    Science.gov (United States)

    Wurtz, Robert H

    2015-09-19

    The visual and oculomotor systems in the brain have been studied extensively in the primate. Together, they can be regarded as a single brain system that underlies active vision--the normal vision that begins with visual processing in the retina and extends through the brain to the generation of eye movement by the brainstem. The system is probably one of the most thoroughly studied brain systems in the primate, and it offers an ideal opportunity to evaluate the advantages and disadvantages of the series of perturbation techniques that have been used to study it. The perturbations have been critical in moving from correlations between neuronal activity and behaviour closer to a causal relation between neuronal activity and behaviour. The same perturbation techniques have also been used to tease out neuronal circuits that are related to active vision that in turn are driving behaviour. The evolution of perturbation techniques includes ablation of both cortical and subcortical targets, punctate chemical lesions, reversible inactivations, electrical stimulation, and finally the expanding optogenetic techniques. The evolution of perturbation techniques has supported progressively stronger conclusions about what neuronal circuits in the brain underlie active vision and how the circuits themselves might be organized.

  11. Evaluation of CRUDTRAN code to predict transport of corrosion products and radioactivity in the PWR primary coolant system

    International Nuclear Information System (INIS)

    Lee, C.B.

    2002-01-01

    CRUDTRAN code is to predict transport of the corrosion products and their radio-activated nuclides such as cobalt-58 and cobalt-60 in the PWR primary coolant system. In CRUDTRAN code the PWR primary circuit is divided into three principal sections such as the core, the coolant and the steam generator. The main driving force for corrosion product transport in the PWR primary coolant comes from coolant temperature change throughout the system and a subsequent change in corrosion product solubility. As the coolant temperature changes around the PWR primary circuit, saturation status of the corrosion products in the coolant also changes such that under-saturation in steam generator and super-saturation in the core. CRUDTRAN code was evaluated by comparison with the results of the in-reactor loop tests simulating the PWR primary coolant system and PWR plant data. It showed that CRUDTRAN could predict variations of cobalt-58 and cobalt-60 radioactivity with time, plant cycle and coolant chemistry in the PWR plant. (author)

  12. Coolant controls of a PEM fuel cell system

    Science.gov (United States)

    Ahn, Jong-Woo; Choe, Song-Yul

    When operating the polymer electrolyte membrane (PEM) fuel cell stack, temperatures in the stack continuously change as the load current varies. The temperature directly affects the rate of chemical reactions and transport of water and reactants. Elevated temperature increases the mobility of water vapor, which reduces the ohmic over-potential in the membrane and eases removal of water produced. Adversely, the high temperature might impose thermal stress on the membrane and cathode catalyst and cause degradation. Conversely, excessive supply of coolants lowers the temperature in the stack and reduces the rate of the chemical reactions and water activity. Corresponding parasitic power dissipated at the electrical coolant pump increases and overall efficiency of the power system drops. Therefore, proper design of a control for the coolant flow plays an important role in ensuring highly reliable and efficient operations of the fuel cell system. Herein, we propose a new temperature control strategy based on a thermal circuit. The proposed thermal circuit consists of a bypass valve, a radiator with a fan, a reservoir and a coolant pump, while a blower and inlet and outlet manifolds are components of the air supply system. Classic proportional and integral (PI) controllers and a state feedback control for the thermal circuit were used in the design. In addition, the heat source term, which is dependent upon the load current, was feed-forwarded to the closed loop and the temperature effects on the air flow rate were minimized. The dynamics and performance of the designed controllers were evaluated and analyzed by computer simulations using developed dynamic fuel cell system models, where a multi-step current and an experimental current profile measured at the federal urban driving schedule (FUDS) were applied. The results show that the proposed control strategy cannot only suppress a temperature rise in the catalyst layer and prevent oxygen starvation, but also reduce the

  13. 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

  14. Thermal hydraulic analyses of accidents associated with coolant leak from the primary circuit through a hole 10 mm equivalent diameter for the needs of PTS

    International Nuclear Information System (INIS)

    Krhounkova, J.; Kral, P.; Parduba, Z.

    1999-10-01

    The conservative assumptions of the analyses were oriented towards a worsening of the process with respect to the pressurized thermal shock (PTS). Four variants were treated, viz. leaks from the cold or hot leg, each at the rated power or zero power. Since the temperature of water supplied to the primary circuit by the emergency core cooling system is an important parameter with respect to a PTS, the calculations were performed by the iterative procedure: the basic thermal hydraulic calculation was performed by the RELAP5/MOD3.2.1 code which calculates the behaviour of the primary and secondary circuits, whereas the MELCOR code was used to calculate the behaviour of the parameters in the hermetic rooms. The calculation by the RELAP code was then repeated using data from the MELCOR calculations. Interventions by the reactor operators were also considered. (P.A.)

  15. International Space Station Active Thermal Control Sub-System On-Orbit Pump Performance and Reliability Using Liquid Ammonia as a Coolant

    Science.gov (United States)

    Morton, Richard D.; Jurick, Matthew; Roman, Ruben; Adamson, Gary; Bui, Chinh T.; Laliberte, Yvon J.

    2011-01-01

    The International Space Station (ISS) contains two Active Thermal Control Sub-systems (ATCS) that function by using a liquid ammonia cooling system collecting waste heat and rejecting it using radiators. These subsystems consist of a number of heat exchangers, cold plates, radiators, the Pump and Flow Control Subassembly (PFCS), and the Pump Module (PM), all of which are Orbital Replaceable Units (ORU's). The PFCS provides the motive force to circulate the ammonia coolant in the Photovoltaic Thermal Control Subsystem (PVTCS) and has been in operation since December, 2000. The Pump Module (PM) circulates liquid ammonia coolant within the External Active Thermal Control Subsystem (EATCS) cooling the ISS internal coolant (water) loops collecting waste heat and rejecting it through the ISS radiators. These PM loops have been in operation since December, 2006. This paper will discuss the original reliability analysis approach of the PFCS and Pump Module, comparing them against the current operational performance data for the ISS External Thermal Control Loops.

  16. Active quenching circuit for a InGaAs single-photon avalanche diode

    International Nuclear Information System (INIS)

    Zheng Lixia; Wu Jin; Xi Shuiqing; Shi Longxing; Liu Siyang; Sun Weifeng

    2014-01-01

    We present a novel gated operation active quenching circuit (AQC). In order to simulate the quenching circuit a complete SPICE model of a InGaAs SPAD is set up according to the I–V characteristic measurement results of the detector. The circuit integrated with aROIC (readout integrated circuit) is fabricated in an CSMC 0.5 μm CMOS process and then hybrid packed with the detector. Chip measurement results show that the functionality of the circuit is correct and the performance is suitable for practical system applications. (semiconductor integrated circuits)

  17. Derivation of criteria for primary circuit activity in an HTGR

    International Nuclear Information System (INIS)

    Su, S.D.; Barsell, A.W.

    1980-11-01

    This paper derives specific criteria for the circulating and plateout activity in the primary circuit for a 2170-MW(t) high temperature gas-cooled reactor-gas turbine (HTGR-GT) plant. Results show that for a design basis, (1) the circulating activity should be limited to 14,000 Ci Kr-88 (a principal nuclide) to meet both offsite dose and containment access constraint during normal operation and depressurization accidents, and (2) the plateout inventories for those important nuclides affecting shutdown maintenance should not exceed 10,000 Ci Ag-110m, 45,000 Ci Cs-134 and 130,000 Ci Cs-137. This paper presents bases and methodology for deriving such criteria and compares them with light water reactors. 5 tables

  18. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system; Transport des radionucleides dans le circuit primaire d`un REP: comparaison des codes MAAP et ECART

    Energy Technology Data Exchange (ETDEWEB)

    Hervouet, C.; Ranval, W. [Electricite de France (EDF), 92 - Clamart (France); Parozzi, F.; Eusebi, M. [Ente Nazionale per l`Energia Elettrica, Rome (Italy)

    1996-04-01

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO{sub 2} and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs.

  19. 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)

  20. On the origin of reproducible sequential activity in neural circuits

    Science.gov (United States)

    Afraimovich, V. S.; Zhigulin, V. P.; Rabinovich, M. I.

    2004-12-01

    Robustness and reproducibility of sequential spatio-temporal responses is an essential feature of many neural circuits in sensory and motor systems of animals. The most common mathematical images of dynamical regimes in neural systems are fixed points, limit cycles, chaotic attractors, and continuous attractors (attractive manifolds of neutrally stable fixed points). These are not suitable for the description of reproducible transient sequential neural dynamics. In this paper we present the concept of a stable heteroclinic sequence (SHS), which is not an attractor. SHS opens the way for understanding and modeling of transient sequential activity in neural circuits. We show that this new mathematical object can be used to describe robust and reproducible sequential neural dynamics. Using the framework of a generalized high-dimensional Lotka-Volterra model, that describes the dynamics of firing rates in an inhibitory network, we present analytical results on the existence of the SHS in the phase space of the network. With the help of numerical simulations we confirm its robustness in presence of noise in spite of the transient nature of the corresponding trajectories. Finally, by referring to several recent neurobiological experiments, we discuss possible applications of this new concept to several problems in neuroscience.

  1. Monolithic microwave integrated circuit devices for active array antennas

    Science.gov (United States)

    Mittra, R.

    1984-01-01

    Two different aspects of active antenna array design were investigated. The transition between monolithic microwave integrated circuits and rectangular waveguides was studied along with crosstalk in multiconductor transmission lines. The boundary value problem associated with a discontinuity in a microstrip line is formulated. This entailed, as a first step, the derivation of the propagating as well as evanescent modes of a microstrip line. The solution is derived to a simple discontinuity problem: change in width of the center strip. As for the multiconductor transmission line problem. A computer algorithm was developed for computing the crosstalk noise from the signal to the sense lines. The computation is based on the assumption that these lines are terminated in passive loads.

  2. 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)

  3. Radiation hazards due to activated corrosion and neutron sputtering products in fusion reactor coolant and tritium breeding fluids

    International Nuclear Information System (INIS)

    Klein, A.C.; Vogelsang, W.F.

    1985-01-01

    The accumulation of radioactive corrosion and neutron sputtering products on the surfaces of components in fusion reactor coolant and tritium breeding systems can cause significant personnel access problems. Remote maintenance techniques or special treatment may be required to limit the amount of radiation exposure to plant operational and maintenance personnel. A computer code, RAPTOR, has been developed to estimate the transport of this activated material throughout a fusion heat transfer and/or tritium breeding material loop. A method is devised which treats the components of the loop individually and determines the source rates, deposition and erosion rates, decay rates, and purification rates of these radioactive materials. RAPTOR has been applied to the MARS and Starfire conceptual reactor designs to determine the degree of the possible radiation hazard due to these products. Due to the very high corrosion release rate by HT-9 when exposed to LiPb in the MARS reactor design, the radiation fields surrounding the primary system will preclude direct contact maintenance even after shutdown. Even the removal of the radioactive LiPb from the system will not decrease the radiation fields to reasonable levels. The Starfire primary system will exhibit radiation fields similar to those found in present pressurized water reactors. (orig.)

  4. The radionuclides of primary coolant in HANARO and the recent activities performed to reduce the radioactivity or reactor pool water

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minjin [HANARO Research Reactor Centre, Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1998-10-01

    In HANARO reactor, there have been activities to identify the principal radionuclides and to quantify them under the normal operation. The purposes of such activities were to establish the measure by which we can reduce the radioactivity of the reactor pool water and detect, in early stage, the abnormal symptoms due to the leakage of radioactive materials from the irradiation sample or the damage of the nuclear fuel, etc. The typical radionuclides produced by the activation of reactor coolant are N{sup 16} and Ar{sup 41}. The radionuclides produced by the activation of the core structural material consist of Na{sup 24}, Mn{sup 56}, and W{sup 187}. Of the various radionuclides, governing the radiation level at the pool surface are Na{sup 24}, Ar{sup 41}, Mn{sup 58}, and W{sup 187}. By establishing the hot water layer system on the pool surface, we expected that the radionuclides such as Ar{sup 41} and Mn{sup 56} whose half-life are relatively short could be removed to a certain extent. Since the content of radioactivity of Na{sup 24} occupies about 60% of the total radioactivity, we assumed that the total radiation level would be greatly reduced if we could decrease the radiation level of Na{sup 24}. However the actual radiation level has not been reduced as much as we expected. Therefore, some experiments have been carried out to find the actual causes afterwards. What we learned through the experiments are that any disturbance in reactor pool water layer causes increase of the pool surface radiation level and even if we maintain the hot water layer well, reactor shutdown will be very much likely to happen once the hot water layer is disturbed. (author)

  5. Advanced active quenching circuit for ultra-fast quantum cryptography.

    Science.gov (United States)

    Stipčević, Mario; Christensen, Bradley G; Kwiat, Paul G; Gauthier, Daniel J

    2017-09-04

    Commercial photon-counting modules based on actively quenched solid-state avalanche photodiode sensors are used in a wide variety of applications. Manufacturers characterize their detectors by specifying a small set of parameters, such as detection efficiency, dead time, dark counts rate, afterpulsing probability and single-photon arrival-time resolution (jitter). However, they usually do not specify the range of conditions over which these parameters are constant or present a sufficient description of the characterization process. In this work, we perform a few novel tests on two commercial detectors and identify an additional set of imperfections that must be specified to sufficiently characterize their behavior. These include rate-dependence of the dead time and jitter, detection delay shift, and "twilighting". We find that these additional non-ideal behaviors can lead to unexpected effects or strong deterioration of the performance of a system using these devices. We explain their origin by an in-depth analysis of the active quenching process. To mitigate the effects of these imperfections, a custom-built detection system is designed using a novel active quenching circuit. Its performance is compared against two commercial detectors in a fast quantum key distribution system with hyper-entangled photons and a random number generator.

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

    International Nuclear Information System (INIS)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A.

    1997-01-01

    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

  7. 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.

  8. 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)

    1998-12-31

    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.

  9. Coolant leakage detection device

    International Nuclear Information System (INIS)

    Ito, Takao.

    1983-01-01

    Purpose: To surely detect the coolant leakage at a time when the leakage amount is still low in the intra-reactor inlet pipeway of FBR type reactor. Constitution: Outside of the intra-reactor inlet piping for introducing coolants at low temperature into a reactor core, an outer closure pipe is furnished. The upper end of the outer closure pipe opens above the liquid level of the coolants in the reactor, and a thermocouple is inserted to the opening of the upper end. In such a structure, if the coolants in the in-reactor piping should leak to the outer closure pipe, coolants over-flows from the opening thereof, at which the thermocouple detects the temperature of the coolants at a low temperature, thereby enabling to detect the leakage of the coolants at a time when it is still low. (Kamimura, M.)

  10. Process and device for monitoring active and highly active liquids for specific nuclides, particularly in the primary coolant of boiling water and pressurized water reactors

    International Nuclear Information System (INIS)

    Paffrath; Haag; Hanstein, W.

    1986-01-01

    The invention provides for a direct continuous measurement of amounts of activity or concentration and the automatic execution of analysis cycles for specific nuclides, which occur depending on the continuously determined total gamma value. The continuously determined total gamma values in a first measuring circuit are used to decide the frequency of sampling, the volume of samples, the measuring times, the doses of separating chemicals and the sample remnants. These parameters are provided for the control of the execution of an analysis cycle for specific nuclides occurring in a second measuring circuit. (orig./HP) [de

  11. Experimental study on cryogenic adsorption of methane by activated carbon for helium coolant purification of High-Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Chang, Hua; Wu, Zong-Xin; Jia, Hai-Jun

    2017-01-01

    Highlights: • The cryogenic CH 4 adsorption on activated carbon was studied for design of HTGR. • The breakthrough curves at different conditions were analyzed by the MTZ model. • The CH 4 adsorption isotherm was fitted well by the Toth model and the D-R model. • The work provides valuable reference data for helium coolant purification of HTGR. - Abstract: The cryogenic adsorption behavior of methane on activated carbon was investigated for helium coolant purification of high-temperature gas-cooled reactor by using dynamic column breakthrough method. With helium as carrier gas, experiments were performed at −196 °C and low methane partial pressure range of 0–120 Pa. The breakthrough curves at different superficial velocities and different feed concentrations were measured and analyzed by the mass-transfer zone model. The methane single-component adsorption isotherm was obtained and fitted well by the Toth model and the Dubinin-Radushkevich model. The adsorption heat of methane on activated carbon was estimated. The cryogenic adsorption process of methane on activated carbon has been verified to be effective for helium coolant purification of high-temperature gas-cooled reactor.

  12. Copper is an endogenous modulator of neural circuit spontaneous activity.

    Science.gov (United States)

    Dodani, Sheel C; Firl, Alana; Chan, Jefferson; Nam, Christine I; Aron, Allegra T; Onak, Carl S; Ramos-Torres, Karla M; Paek, Jaeho; Webster, Corey M; Feller, Marla B; Chang, Christopher J

    2014-11-18

    For reasons that remain insufficiently understood, the brain requires among the highest levels of metals in the body for normal function. The traditional paradigm for this organ and others is that fluxes of alkali and alkaline earth metals are required for signaling, but transition metals are maintained in static, tightly bound reservoirs for metabolism and protection against oxidative stress. Here we show that copper is an endogenous modulator of spontaneous activity, a property of functional neural circuitry. Using Copper Fluor-3 (CF3), a new fluorescent Cu(+) sensor for one- and two-photon imaging, we show that neurons and neural tissue maintain basal stores of loosely bound copper that can be attenuated by chelation, which define a labile copper pool. Targeted disruption of these labile copper stores by acute chelation or genetic knockdown of the CTR1 (copper transporter 1) copper channel alters the spatiotemporal properties of spontaneous activity in developing hippocampal and retinal circuits. The data identify an essential role for copper neuronal function and suggest broader contributions of this transition metal to cell signaling.

  13. 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

  14. Leak detection system for RBMK coolant circuit

    Energy Technology Data Exchange (ETDEWEB)

    Cherkashov, Ju M; Strelkov, B P; Korolev, Yu V; Eperin, A P; Kozlov, E P; Belyanin, L A; Vanukov, V N [Leningrad Nuclear Power Plant, Leningrad (Russian Federation). Research and Development Inst. of Power Engineering

    1997-12-31

    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.

  15. HANARO secondary coolant management

    International Nuclear Information System (INIS)

    Kim, Seon Duk.

    1998-02-01

    In this report, the basic theory for management of water quality, environmental factors influencing to the coolant, chemicals and its usage for quality control of coolant are mentioned, and water balance including the loss rate by evaporation (34.3 m 3 /hr), discharge rate (12.665 m 3 /hr), concentration ratio and feed rate (54.1 m 3 /hr) are calculated at 20 MW operation. Also, the analysis data of HANSU Limited for HANARO secondary coolant (feed water and circulating coolant) - turbidity, pH, conductivity, M-alkalinity, Ca-hardness, chloride ion, total iron ion, phosphoric ion and conversion rate are reviewed. It is confirmed that the feed water has good quality and the circulating coolant has been maintained within the control specification in general, but some items exceeded the control specification occasionally. Therefore it is judged that more regular discharge of coolant is needed. (author). 6 refs., 17 tabs., 18 figs

  16. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

  17. Passive and active elements using fractional Lβ C α circuit

    KAUST Repository

    Radwan, Ahmed G.; Salama, Khaled N.

    2011-01-01

    this illustration to verify the cases discussed in part one; develop the generalized parameters for the bandpass filter response of the L βCα circuit, such as the resonance frequency and quality factor versus α-β plane; discuss sensitivity analyses with respect

  18. Nuclear reactor of pressurized liquid coolant type

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

  19. A High Step-Down Interleaved Buck Converter with Active-Clamp Circuits for Wind Turbines

    Directory of Open Access Journals (Sweden)

    Chih-Lung Shen

    2012-12-01

    Full Text Available In this paper, a high step-down interleaved buck coupled-inductor converter (IBCC with active-clamp circuits for wind energy conversion has been studied. In high step-down voltage applications, an IBCC can extend duty ratio and reduce voltage stresses on active switches. In order to reduce switching losses of active switches to improve conversion efficiency, a IBCC with soft-switching techniques is usually required. Compared with passive-clamp circuits, the IBCC with active-clamp circuits have lower switching losses and minimum ringing voltage of the active switches. Thus, the proposed IBCC with active-clamp circuits for wind energy conversion can significantly increase conversion efficiency. Finally, a 240 W prototype of the proposed IBCC with active-clamp circuits was built and implemented. Experimental results have shown that efficiency can reach as high as 91%. The proposed IBCC with active-clamp circuits is presented in high step-down voltage applications to verify the performance and the feasibility for energy conversion of wind turbines.

  20. Potential for low fracture toughness and lamellar tearing on PWR steam generator and reactor coolant pump supports. Resolution of generic technical activity A-12

    International Nuclear Information System (INIS)

    Snaider, R.P.; Hodge, J.M.; Levin, H.A.; Zudans, J.J.

    1979-10-01

    This report summarizes work performed by the Nuclear Regulatory Commission staff and its contractor, Sandia Laboratories, in the resolution of Generic Technical Activity A-12, ''Potential for Low Fracture Toughness and Lamellar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports.'' The report describes the technical issues, the technical studies performed by Sandia describes the technical issues, the technical studies performed by Sandia Laboratories, the NRC staff's technical positions based on these studies, and the staff's plan for implementing its technical positions. It also provides recommendations for further work. The complete technical input from Sandia Laboratories is appended to the report

  1. Calculation of activity concentration and dose rates from online radioactivity measurement in primary coolant channel of TAPS-III and IV

    International Nuclear Information System (INIS)

    Chaudhury, Sanhita; Agarwal, Chhavi; Goswami, A.; Mhatre, Amol; Chaturvedi, T.P.; Tawde, N.; Gathibandhe, Manohar; Dash, S.C.

    2011-05-01

    Radioactivity measurement using CdZnTe detector and dose measurement using teletector were done at several locations of primary heat transport (PHT) system of the Tarapur Atomic Power Station-III and IV reactor during shut down as well as operating condition of the reactors. The detector efficiency for the required counting geometry was simulated using MCNP code. Using this simulated efficiency and the experimental count rate (cps), the activity concentrations (Bq/mL) of different radionuclides in coolant water were calculated. The dose rates for the counted locations were also simulated using Monte Carlo code and it matched well with the experimentally obtained dose rate. (author)

  2. Passive and active elements using fractional Lβ C α circuit

    KAUST Repository

    Radwan, Ahmed G.

    2011-10-01

    This paper introduces a qualitative revision of the traditional LC tank circuit in the fractional domain. The paper can be divided into six major parts, aiming in turn to establish the various conditions under which L βCα impedance may act as a resistor, negative resistor, or a positive or negative pure imaginary inductor or capacitor, in accordance to new frequency definitions; illustrate the process by which the phase response chooses the shortest path from initial to final phase, and use this illustration to verify the cases discussed in part one; develop the generalized parameters for the bandpass filter response of the L βCα circuit, such as the resonance frequency and quality factor versus α-β plane; discuss sensitivity analyses with respect to the fractional orders, as well as the time domain analyses for the impulse and step responses with their analytical formulas; and lastly, to propose some possible applications for this generalized circuit. Mathematical and PSpice simulation results are included to validate the discussion. © 2011 IEEE.

  3. Conductance and activation energy for electron transport in series and parallel intramolecular circuits.

    Science.gov (United States)

    Hsu, Liang-Yan; Wu, Ning; Rabitz, Herschel

    2016-11-30

    We investigate electron transport through series and parallel intramolecular circuits in the framework of the multi-level Redfield theory. Based on the assumption of weak monomer-bath couplings, the simulations depict the length and temperature dependence in six types of intramolecular circuits. In the tunneling regime, we find that the intramolecular circuit rule is only valid in the weak monomer coupling limit. In the thermally activated hopping regime, for circuits based on two different molecular units M a and M b with distinct activation energies E act,a > E act,b , the activation energies of M a and M b in series are nearly the same as E act,a while those in parallel are nearly the same as E act,b . This study gives a comprehensive description of electron transport through intramolecular circuits from tunneling to thermally activated hopping. We hope that this work can motivate additional studies to design intramolecular circuits based on different types of building blocks, and to explore the corresponding circuit laws and the length and temperature dependence of conductance.

  4. Coolant cleanup method in a nuclear reactor

    International Nuclear Information System (INIS)

    Kubota, Masayoshi; Nishimura, Shigeoki; Takahashi, Sankichi; Izumi, Kenkichi; Motojima, Kenji.

    1983-01-01

    Purpose : To effectively adsorb to remove low molecular weight organic substances from iron exchange resins for use in the removal of various radioactive nucleides contained in reactor coolants. Method : Reactor coolants are recycled by a main recyling pump in a nuclear reactor and a portion of the coolants is cooled and, thereafter, purified in a coolant desalter. While on the other hand, high pressure steams generated from the reactor are passed through a turbine, cooled in a condensator, eliminated with claddings or the likes by the passage through a filtration desalter using powderous ion exchange resins and then further passed through a desalter (filled with granular ion exchange resins). For instance, an adsorption and removing device for organic substances (resulted through the decomposition of ion exchange resins) precoated with activated carbon powder or filled with granular activated carbon is disposed at the downstream for each of the desalters. In this way, the organic substances in the coolants are eliminated to prevent the reduction in the desalting performance of the ion exchange resins caused by the formation of complexes between organic substances and cobalt in the coolants, etc. In this way, the coolant cleanup performance is increased and the amount of wasted ion exchange resins can be decreased. (Horiuchi, T.)

  5. The sodium coolant

    International Nuclear Information System (INIS)

    Rodriguez, G.

    2004-01-01

    The sodium is the best appropriate coolant for the fast neutrons reactors technology. Thus the fast neutrons reactors development is intimately bound to the sodium technology. This document presents the sodium as a coolant point of view: atomic structure and characteristics, sodium impacts on the fast neutron reactors technology, chemical properties of the sodium and the consequences, quality control in a nuclear reactor, sodium treatment. (A.L.B.)

  6. 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.)

  7. Extended Life Coolant Testing

    Science.gov (United States)

    2016-06-06

    number. PLEASE DO NOT RETURN YOUR FORM TO THE ABOVE ADDRESS. 1. REPORT DATE (DD-MM-YYYY) 06-06-2016 2. REPORT TYPE Interim Report 3. DATES COVERED ... Corrosion Testing of Traditional and Extended Life Coolants 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Hansen, Gregory A. T...providing vehicle specific coolants. Several laboratory corrosion tests were performed according to ASTM D1384 and D2570, but with a 2.5x extended time

  8. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  9. Hydraulically-activated operating system for an electric circuit breaker

    Science.gov (United States)

    Imam, Imdad; Barkan, Philip

    1979-01-01

    This operating system comprises a fluid motor having a piston, a breaker-opening space at one side of the piston, and a breaker-closing space at its opposite side. An accumulator freely communicates with the breaker-opening space for supplying pressurized fluid thereto during a circuit-breaker opening operation. A normally-closed valve located on the breaker-closing-side of the piston is openable to release liquid from the breaker-closing space so that pressurized liquid in the breaker-opening space can drive the piston in an opening direction. Means is provided for restoring the valve to its closed position following the circuit-breaker opening operation. An impeded passage affords communication between the accumulator and the breaker-closing space to allow pressurized liquid to flow from the accumulator to the breaker-closing space and develop a pressure therein substantially equal to accumulator pressure when the valve is restored to closed position following breaker-opening. This passage is so impeded that the flow therethrough from the accumulator into the breaker-closing space is sufficiently low during initial opening motion of the piston through a substantial portion of its opening stroke as to avoid interference with said initial opening motion of the piston.

  10. Persistent activity in a recurrent circuit underlies courtship memory in Drosophila

    Science.gov (United States)

    Zhao, Xiaoliang; Lenek, Daniela; Dag, Ugur; Dickson, Barry J

    2018-01-01

    Recurrent connections are thought to be a common feature of the neural circuits that encode memories, but how memories are laid down in such circuits is not fully understood. Here we present evidence that courtship memory in Drosophila relies on the recurrent circuit between mushroom body gamma (MBγ), M6 output, and aSP13 dopaminergic neurons. We demonstrate persistent neuronal activity of aSP13 neurons and show that it transiently potentiates synaptic transmission from MBγ>M6 neurons. M6 neurons in turn provide input to aSP13 neurons, prolonging potentiation of MBγ>M6 synapses over time periods that match short-term memory. These data support a model in which persistent aSP13 activity within a recurrent circuit lays the foundation for a short-term memory. PMID:29322941

  11. Persistent activity in a recurrent circuit underlies courtship memory in Drosophila.

    Science.gov (United States)

    Zhao, Xiaoliang; Lenek, Daniela; Dag, Ugur; Dickson, Barry J; Keleman, Krystyna

    2018-01-11

    Recurrent connections are thought to be a common feature of the neural circuits that encode memories, but how memories are laid down in such circuits is not fully understood. Here we present evidence that courtship memory in Drosophila relies on the recurrent circuit between mushroom body gamma (MBγ), M6 output, and aSP13 dopaminergic neurons. We demonstrate persistent neuronal activity of aSP13 neurons and show that it transiently potentiates synaptic transmission from MBγ>M6 neurons. M6 neurons in turn provide input to aSP13 neurons, prolonging potentiation of MB γ >M6 synapses over time periods that match short-term memory. These data support a model in which persistent aSP13 activity within a recurrent circuit lays the foundation for a short-term memory. © 2018, Zhao et al.

  12. Coolant leakage detecting device

    International Nuclear Information System (INIS)

    Yamauchi, Kiyoshi; Kawai, Katsunori; Ishihara, Yoshinao.

    1995-01-01

    The device of the present invention judges an amount of leakage of primary coolants of a PWR power plant at high speed. Namely, a mass of coolants contained in a pressurizer, a volume controlling tank and loop regions is obtained based on a preset relational formula and signals of each of process amount, summed up to determine the total mass of coolants for every period of time. The amount of leakage for every period of time is calculated by a formula of Karman's filter based on the total mass of the primary coolants for every predetermined period of time, and displays it on CRT. The Karman's filter is formed on every formula for several kinds of states formed based on the preset amount of the leakage, to calculate forecasting values for every mass of coolants. An adaptable probability for every preset leakage amount is determined based on the difference between the forecast value and the observed value and the scattering thereof. The adaptable probability is compared with a predetermined threshold value, which is displayed on the CRT. This device enables earlier detection of leakage and identification of minute leakage amount as compared with the prior device. (I.S.)

  13. 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

  14. 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)

  15. 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

  16. Experiences of activity measurements of primary circuit materials in a WWR-SM research reactor

    International Nuclear Information System (INIS)

    Elek, A.; Toth, M.; Bakos, L.; Vizdos, G.

    1980-01-01

    The activity of water and gas samples taken from the primary circuit have been measured nondestructively for more than two years to monitor the technological parameters of the reactor. In the primary water samples 17 fission products and seven activated traces, as well as six radioactive conponents in the gas samples were determined routinely by Ge/Li gamma-spectrometry. (author)

  17. The Electron Runaround: Understanding Electric Circuit Basics Through a Classroom Activity

    Science.gov (United States)

    Singh, Vandana

    2010-05-01

    Several misconceptions abound among college students taking their first general physics course, and to some extent pre-engineering physics students, regarding the physics and applications of electric circuits. Analogies used in textbooks, such as those that liken an electric circuit to a piped closed loop of water driven by a water pump, do not completely resolve these misconceptions. Mazur and Knight,2 in particular, separately note that such misconceptions include the notion that electric current on either side of a light bulb in a circuit can be different. Other difficulties and confusions involve understanding why the current in a parallel circuit exceeds the current in a series circuit with the same components, and include the role of the battery (where students may assume wrongly that a dry cell battery is a fixed-current rather than a fixed-voltage device). A simple classroom activity that students can play as a game can resolve these misconceptions, providing an intellectual as well as a hands-on understanding. This paper describes the "Electron Runaround," first developed by the author to teach extremely bright 8-year-old home-schooled children the basics of electric circuits and subsequently altered (according to the required level of instruction) and used for various college physics courses.

  18. Coolant system decontamination

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  19. Sensitive detection of proteasomal activation using the Deg-On mammalian synthetic gene circuit.

    Science.gov (United States)

    Zhao, Wenting; Bonem, Matthew; McWhite, Claire; Silberg, Jonathan J; Segatori, Laura

    2014-04-08

    The ubiquitin proteasome system (UPS) has emerged as a drug target for diverse diseases characterized by altered proteostasis, but pharmacological agents that enhance UPS activity have been challenging to establish. Here we report the Deg-On system, a genetic inverter that translates proteasomal degradation of the transcriptional regulator TetR into a fluorescent signal, thereby linking UPS activity to an easily detectable output, which can be tuned using tetracycline. We demonstrate that this circuit responds to modulation of UPS activity in cell culture arising from the inhibitor MG-132 and activator PA28γ. Guided by predictive modelling, we enhanced the circuit's signal sensitivity and dynamic range by introducing a feedback loop that enables self-amplification of TetR. By linking UPS activity to a simple and tunable fluorescence output, these genetic inverters will enable a variety of applications, including screening for UPS activating molecules and selecting for mammalian cells with different levels of proteasome activity.

  20. Integrated Printed Circuit Board (PCB) Active Cooling With Piezoelectric Actuator

    Science.gov (United States)

    2012-09-01

    prepreg epoxy. Individual FR-4 lamina were mechanically machined to pattern each layer. The layers were aligned, stacked, and laminated to form the...The cooler substrate is a laminated multilayer FR-4 substrate. Individual layers are patterned to support the active element, form a resonant... laminated with 70/30 copper-nickel alloy or 80/20 nickel-chrome alloy and patterned by means of photolithographic techniques and wet etching in a ferric

  1. Practical aspects of 13C surface receive coils with active decoupling and tuning circuit

    DEFF Research Database (Denmark)

    Nilsson, Daniel; Mohr, Johan Jacob; Zhurbenko, Vitaliy

    2012-01-01

    is based on application-specified coil profile and includes impedance matching and balancing circuits. Active decoupling is implemented in order to minimize the influence of the receiving coil on the homogeneity of the transmit-coil field. Measurement results for a coil prototype are presented, including...

  2. 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)

  3. Chemistry of liquid metal coolants and sensors

    International Nuclear Information System (INIS)

    Gnanasekaran, T.

    2015-01-01

    Liquid sodium is the coolant of choice for the current generation fast breeder reactors. When sodium contains low levels of dissolved non-metallic impurities, it is highly compatible with structural steels. When the dissolved oxygen level is high, corrosion and mass transfer in sodium-steel circuits are enhanced and this involves formation of NaxMyOz type of species (M = alloying components in steels). Experience has shown that this enhancement of corrosion in a sodium circuit with all austenitic steel structural materials would not be encountered if oxygen level in sodium is below ~ 5ppm. For understanding this observation, a complete knowledge on the phase diagrams of Na-M-O systems and the thermochemical data of all relevant NaxMyOz compounds is essential. This presentation would highlight the work carried out at IGCAR on the chemistry of liquid sodium and heavy liquid metal coolants. Work carried out on various sensors for their use in these liquid metal circuits would be described and their current status would be discussed

  4. 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

  5. Compartmentalized safety coolant injection system

    International Nuclear Information System (INIS)

    Johnson, F.T.

    1983-01-01

    A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

  6. Radiolysis of the VVER-1000 reactor coolant: An experimental study and mathematical modeling

    International Nuclear Information System (INIS)

    Arkhipov, O.P.; Bugaenko, V.L.; Kabakchi, S.A.

    1995-01-01

    Variations in the composition of the coolant for the primary circuit of a VVER-1000 reactor of the Kalinin nuclear power plant upon transition from power-level operation to shutdown was studied experimentally. The data obtained were used for verification of the MORAVA-H2 program developed earlier for simulation of the coolant state in pressurized-water power reactors

  7. 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)

  8. Oncogenes Activate an Autonomous Transcriptional Regulatory Circuit That Drives Glioblastoma

    Directory of Open Access Journals (Sweden)

    Dinesh K. Singh

    2017-01-01

    Full Text Available Efforts to identify and target glioblastoma (GBM drivers have primarily focused on receptor tyrosine kinases (RTKs. Clinical benefits, however, have been elusive. Here, we identify an SRY-related box 2 (SOX2 transcriptional regulatory network that is independent of upstream RTKs and capable of driving glioma-initiating cells. We identified oligodendrocyte lineage transcription factor 2 (OLIG2 and zinc-finger E-box binding homeobox 1 (ZEB1, which are frequently co-expressed irrespective of driver mutations, as potential SOX2 targets. In murine glioma models, we show that different combinations of tumor suppressor and oncogene mutations can activate Sox2, Olig2, and Zeb1 expression. We demonstrate that ectopic co-expression of the three transcription factors can transform tumor-suppressor-deficient astrocytes into glioma-initiating cells in the absence of an upstream RTK oncogene. Finally, we demonstrate that the transcriptional inhibitor mithramycin downregulates SOX2 and its target genes, resulting in markedly reduced proliferation of GBM cells in vivo.

  9. Bidirectional global spontaneous network activity precedes the canonical unidirectional circuit organization in the developing hippocampus.

    Science.gov (United States)

    Shi, Yulin; Ikrar, Taruna; Olivas, Nicholas D; Xu, Xiangmin

    2014-06-15

    Spontaneous network activity is believed to sculpt developing neural circuits. Spontaneous giant depolarizing potentials (GDPs) were first identified with single-cell recordings from rat CA3 pyramidal neurons, but here we identify and characterize a large-scale spontaneous network activity we term global network activation (GNA) in the developing mouse hippocampal slices, which is measured macroscopically by fast voltage-sensitive dye imaging. The initiation and propagation of GNA in the mouse is largely GABA-independent and dominated by glutamatergic transmission via AMPA receptors. Despite the fact that signal propagation in the adult hippocampus is strongly unidirectional through the canonical trisynaptic circuit (dentate gyrus [DG] to CA3 to CA1), spontaneous GNA in the developing hippocampus originates in distal CA3 and propagates both forward to CA1 and backward to DG. Photostimulation-evoked GNA also shows prominent backward propagation in the developing hippocampus from CA3 to DG. Mouse GNA is strongly correlated to electrophysiological recordings of highly localized single-cell and local field potential events. Photostimulation mapping of neural circuitry demonstrates that the enhancement of local circuit connections to excitatory pyramidal neurons occurs over the same time course as GNA and reveals the underlying pathways accounting for GNA backward propagation from CA3 to DG. The disappearance of GNA coincides with a transition to the adult-like unidirectional circuit organization at about 2 weeks of age. Taken together, our findings strongly suggest a critical link between GNA activity and maturation of functional circuit connections in the developing hippocampus. Copyright © 2013 Wiley Periodicals, Inc.

  10. 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_D_S = f(V_D_S, V_G_S, v_B_S) 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 thesis. The first

  11. Fluorescence-based monitoring of in vivo neural activity using a circuit-tracing pseudorabies virus.

    Directory of Open Access Journals (Sweden)

    Andrea E Granstedt

    Full Text Available The study of coordinated activity in neuronal circuits has been challenging without a method to simultaneously report activity and connectivity. Here we present the first use of pseudorabies virus (PRV, which spreads through synaptically connected neurons, to express a fluorescent calcium indicator protein and monitor neuronal activity in a living animal. Fluorescence signals were proportional to action potential number and could reliably detect single action potentials in vitro. With two-photon imaging in vivo, we observed both spontaneous and stimulated activity in neurons of infected murine peripheral autonomic submandibular ganglia (SMG. We optically recorded the SMG response in the salivary circuit to direct electrical stimulation of the presynaptic axons and to physiologically relevant sensory stimulation of the oral cavity. During a time window of 48 hours after inoculation, few spontaneous transients occurred. By 72 hours, we identified more frequent and prolonged spontaneous calcium transients, suggestive of neuronal or tissue responses to infection that influence calcium signaling. Our work establishes in vivo investigation of physiological neuronal circuit activity and subsequent effects of infection with single cell resolution.

  12. A voltage control method for an active capacitive DC-link module with series-connected circuit

    DEFF Research Database (Denmark)

    Wang, Haoran; Wang, Huai; Blaabjerg, Frede

    2017-01-01

    Many efforts have been made to improve the performance of power electronic systems with active capacitive DC-link module in terms of power density as well as reliability. One of the attractive solution is an active capacitive DC-link with the series-connected circuit because of handling small......-rated power. However, in the existing control method of this circuit, the DC-link current of the backward-stage or forward-stage need to be sensed for extracting the ripple components, which limits the flexibility of the active DC-link module. Thus, in this paper, a voltage control method of an active...... capacitive DC-link module is proposed. Current sensor at the DC-link will be cancel from the circuit. The controller of the series-connected circuit requires internal voltage signals of the DC-link module only, making it possible to be fully independent without any additional connection to the main circuit...

  13. Leak detection device for reactor coolant

    International Nuclear Information System (INIS)

    Oshima, Koichiro.

    1990-01-01

    In a light water cooled reactor, if reactor coolants are leaked from pipelines in a pipeline chamber, activated products (N-16) are diffused together to an atmosphere in the pipeline chamber. N-16 is sucked from an extracting tube which is always sucking the atmosphere in the pipeline chamber to a sucking blower. Then, β-rays released from N-16 are monitored by a radiation monitor in a measuring chamber which is radiation-shielded from the pipeline chamber. Accordingly, since the radiation monitor can detect even slight leakage, the slight leakage of reactor coolants in the pipelines can be detected at an early stage. (I.N.)

  14. Simulation of a loss of coolant accident

    International Nuclear Information System (INIS)

    1987-06-01

    An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. With this objective in mind, the Central Research Institute for Physics (CRIP) of the Hungarian Academy of Sciences designed and constructed the PMK-NVH (Paks Model Circuit) test facility, a scaled down model of the WWER-440 Paks nuclear power plant. Hungary with the aim of strengthening the international co-operation on nuclear safety, made the PMK-NVH facility available to the IAEA to conduct a standard problem exercise. In this exercise, experimental data from the simulation of a 7.4% break loss of coolant accident were compared with analytical predictions of the behaviour of the facility calculated with computer codes. This document presents a complete overview of the Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation

  15. Understanding Activation Patterns in Shared Circuits: Toward a Value Driven Model

    Directory of Open Access Journals (Sweden)

    Lisa Aziz-Zadeh

    2018-05-01

    Full Text Available Over the past decade many studies indicate that we utilize our own motor system to understand the actions of other people. This mirror neuron system (MNS has been proposed to be involved in social cognition and motor learning. However, conflicting findings regarding the underlying mechanisms that drive these shared circuits make it difficult to decipher a common model of their function. Here we propose adapting a “value-driven” model to explain discrepancies in the human mirror system literature and to incorporate this model with existing models. We will use this model to explain discrepant activation patterns in multiple shared circuits in the human data, such that a unified model may explain reported activation patterns from previous studies as a function of value.

  16. 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

  17. Metabolic activation of amygdala, lateral septum and accumbens circuits during food anticipatory behavior.

    Science.gov (United States)

    Olivo, Diana; Caba, Mario; Gonzalez-Lima, Francisco; Rodríguez-Landa, Juan F; Corona-Morales, Aleph A

    2017-01-01

    When food is restricted to a brief fixed period every day, animals show an increase in temperature, corticosterone concentration and locomotor activity for 2-3h before feeding time, termed food anticipatory activity. Mechanisms and neuroanatomical circuits responsible for food anticipatory activity remain unclear, and may involve both oscillators and networks related to temporal conditioning. Rabbit pups are nursed once-a-day so they represent a natural model of circadian food anticipatory activity. Food anticipatory behavior in pups may be associated with neural circuits that temporally anticipate feeding, while the nursing event may produce consummatory effects. Therefore, we used New Zealand white rabbit pups entrained to circadian feeding to investigate the hypothesis that structures related to reward expectation and conditioned emotional responses would show a metabolic rhythm anticipatory of the nursing event, different from that shown by structures related to reward delivery. Quantitative cytochrome oxidase histochemistry was used to measure regional brain metabolic activity at eight different times during the day. We found that neural metabolism peaked before nursing, during food anticipatory behavior, in nuclei of the extended amygdala (basolateral, medial and central nuclei, bed nucleus of the stria terminalis), lateral septum and accumbens core. After pups were fed, however, maximal metabolic activity was expressed in the accumbens shell, caudate, putamen and cortical amygdala. Neural and behavioral activation persisted when animals were fasted by two cycles, at the time of expected nursing. These findings suggest that metabolic activation of amygdala-septal-accumbens circuits involved in temporal conditioning may contribute to food anticipatory activity. Copyright © 2016 Elsevier B.V. All rights reserved.

  18. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, V; Rosenberg, R [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  19. Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system

    International Nuclear Information System (INIS)

    Samoilov, O.B.; Kuul, V.S.; Malamud, V.A.; Tarasov, G.I.

    1996-01-01

    The development of a small power (up to 240 MWe) integral PWR for nuclear co-generation power plants has been carried out. The distinctive features of this advanced reactor are: primary circuit arrangement in a single pressure vessel; natural coolant circulation; passive safety systems with self-activated control devices; use of a second (guard) vessel housing the reactor; favourable conditions for the most severe accident management. A passive steam condensing channel has been developed which is activated by the direct action of the primary circuit pressure without an automatic controlling action or manual intervention for emergency cooling of an integral reactor with an in-built pressurizer. In an emergency situation as pressure rises in the reactor a self-activated device blows out non-condensable gases from the condenser tube bundle and returns them in the steam-condensing mode of the operation with the returing primary coolant condensate into the reactor. The thermo-physical test facility is constructed and the experimental development of the steam-condensing channels is performed aiming at the verification of mathematical models for these channels operation in integral reactors both at loss-of-heat removal and LOCA accidents. (orig.)

  20. Toolbox for the design of LiNbO3-based passive and active integrated quantum circuits

    Science.gov (United States)

    Sharapova, P. R.; Luo, K. H.; Herrmann, H.; Reichelt, M.; Meier, T.; Silberhorn, C.

    2017-12-01

    We present and discuss perspectives of current developments on advanced quantum optical circuits monolithically integrated in the lithium niobate platform. A set of basic components comprising photon pair sources based on parametric down conversion (PDC), passive routing elements and active electro-optically controllable switches and polarisation converters are building blocks of a toolbox which is the basis for a broad range of diverse quantum circuits. We review the state-of-the-art of these components and provide models that properly describe their performance in quantum circuits. As an example for applications of these models we discuss design issues for a circuit providing on-chip two-photon interference. The circuit comprises a PDC section for photon pair generation followed by an actively controllable modified mach-Zehnder structure for observing Hong-Ou-Mandel interference. The performance of such a chip is simulated theoretically by taking even imperfections of the properties of the individual components into account.

  1. A Voltage Doubler Circuit to Extend the Soft-switching Range of Dual Active Bridge Converters

    DEFF Research Database (Denmark)

    Qin, Zian; Shen, Yanfeng; Wang, Huai

    2017-01-01

    A voltage doubler circuit is realized to extend the soft-switching range of Dual Active Bridge (DAB) converters. No extra hardware is added to the DAB to form this circuit, since it is composed of the dc blocking capacitor and the low side full bridge converter, which already exist in DAB....... With the voltage doubler, the DAB converter can achieve soft switching and high efficiency when the low side dc voltage is close to 2 pu (1 pu is the high side dc voltage divided by the transformer turn ratio), which can be realized only when the low side dc voltage is close to 1 pu by using the conventional phase...... shift modulation in DAB. Thus the soft switching range is extended. The soft switching boundary conditions are derived. A map to show the soft switching or hard switching in the full load and voltage range is obtained. The feasibility and effectiveness of the proposed method is finally verified...

  2. Multichannel brain recordings in behaving Drosophila reveal oscillatory activity and local coherence in response to sensory stimulation and circuit activation

    Science.gov (United States)

    Paulk, Angelique C.; Zhou, Yanqiong; Stratton, Peter; Liu, Li

    2013-01-01

    Neural networks in vertebrates exhibit endogenous oscillations that have been associated with functions ranging from sensory processing to locomotion. It remains unclear whether oscillations may play a similar role in the insect brain. We describe a novel “whole brain” readout for Drosophila melanogaster using a simple multichannel recording preparation to study electrical activity across the brain of flies exposed to different sensory stimuli. We recorded local field potential (LFP) activity from >2,000 registered recording sites across the fly brain in >200 wild-type and transgenic animals to uncover specific LFP frequency bands that correlate with: 1) brain region; 2) sensory modality (olfactory, visual, or mechanosensory); and 3) activity in specific neural circuits. We found endogenous and stimulus-specific oscillations throughout the fly brain. Central (higher-order) brain regions exhibited sensory modality-specific increases in power within narrow frequency bands. Conversely, in sensory brain regions such as the optic or antennal lobes, LFP coherence, rather than power, best defined sensory responses across modalities. By transiently activating specific circuits via expression of TrpA1, we found that several circuits in the fly brain modulate LFP power and coherence across brain regions and frequency domains. However, activation of a neuromodulatory octopaminergic circuit specifically increased neuronal coherence in the optic lobes during visual stimulation while decreasing coherence in central brain regions. Our multichannel recording and brain registration approach provides an effective way to track activity simultaneously across the fly brain in vivo, allowing investigation of functional roles for oscillations in processing sensory stimuli and modulating behavior. PMID:23864378

  3. Multichannel brain recordings in behaving Drosophila reveal oscillatory activity and local coherence in response to sensory stimulation and circuit activation.

    Science.gov (United States)

    Paulk, Angelique C; Zhou, Yanqiong; Stratton, Peter; Liu, Li; van Swinderen, Bruno

    2013-10-01

    Neural networks in vertebrates exhibit endogenous oscillations that have been associated with functions ranging from sensory processing to locomotion. It remains unclear whether oscillations may play a similar role in the insect brain. We describe a novel "whole brain" readout for Drosophila melanogaster using a simple multichannel recording preparation to study electrical activity across the brain of flies exposed to different sensory stimuli. We recorded local field potential (LFP) activity from >2,000 registered recording sites across the fly brain in >200 wild-type and transgenic animals to uncover specific LFP frequency bands that correlate with: 1) brain region; 2) sensory modality (olfactory, visual, or mechanosensory); and 3) activity in specific neural circuits. We found endogenous and stimulus-specific oscillations throughout the fly brain. Central (higher-order) brain regions exhibited sensory modality-specific increases in power within narrow frequency bands. Conversely, in sensory brain regions such as the optic or antennal lobes, LFP coherence, rather than power, best defined sensory responses across modalities. By transiently activating specific circuits via expression of TrpA1, we found that several circuits in the fly brain modulate LFP power and coherence across brain regions and frequency domains. However, activation of a neuromodulatory octopaminergic circuit specifically increased neuronal coherence in the optic lobes during visual stimulation while decreasing coherence in central brain regions. Our multichannel recording and brain registration approach provides an effective way to track activity simultaneously across the fly brain in vivo, allowing investigation of functional roles for oscillations in processing sensory stimuli and modulating behavior.

  4. 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

  5. A multi coding technique to reduce transition activity in VLSI circuits

    International Nuclear Information System (INIS)

    Vithyalakshmi, N.; Rajaram, M.

    2014-01-01

    Advances in VLSI technology have enabled the implementation of complex digital circuits in a single chip, reducing system size and power consumption. In deep submicron low power CMOS VLSI design, the main cause of energy dissipation is charging and discharging of internal node capacitances due to transition activity. Transition activity is one of the major factors that also affect the dynamic power dissipation. This paper proposes power reduction analyzed through algorithm and logic circuit levels. In algorithm level the key aspect of reducing power dissipation is by minimizing transition activity and is achieved by introducing a data coding technique. So a novel multi coding technique is introduced to improve the efficiency of transition activity up to 52.3% on the bus lines, which will automatically reduce the dynamic power dissipation. In addition, 1 bit full adders are introduced in the Hamming distance estimator block, which reduces the device count. This coding method is implemented using Verilog HDL. The overall performance is analyzed by using Modelsim and Xilinx Tools. In total 38.2% power saving capability is achieved compared to other existing methods. (semiconductor technology)

  6. Coolant channel module CCM

    International Nuclear Information System (INIS)

    Hoeld, Alois

    2007-01-01

    A complete and detailed description of the theoretical background of an '(1D) thermal-hydraulic drift-flux based mixture-fluid' coolant channel model and its resulting module CCM will be presented. The objective of this module is to simulate as universally as possible the steady state and transient behaviour of the key characteristic parameters of a single- or two-phase fluid flowing within any type of heated or non-heated coolant channel. Due to the possibility that different flow regimes can appear along any channel, such a 'basic (BC)' 1D channel is assumed to be subdivided into a number of corresponding sub-channels (SC-s). Each SC can belong to only two types of flow regime, an SC with just a single-phase fluid, containing exclusively either sub-cooled water or superheated steam, or an SC with a two-phase mixture flow. After an appropriate nodalisation of such a BC (and therefore also its SC-s) a 'modified finite volume method' has been applied for the spatial discretisation of the partial differential equations (PDE-s) which represent the basic conservation equations of thermal-hydraulics. Special attention had to be given to the possibility of variable SC entrance or outlet positions (which describe boiling boundaries or mixture levels) and thus the fact that an SC can even disappear or be created anew. The procedure yields for each SC type (and thus the entire BC), a set of non-linear ordinary 1st order differential equations (ODE-s). To link the resulting mean nodal with the nodal boundary function values, both of which are present in the discretised differential equations, a special quadratic polygon approximation procedure (PAX) had to be constructed. Together with the very thoroughly tested packages for drift-flux, heat transfer and single- and two-phase friction factors this procedure represents the central part of the here presented 'Separate-Region' approach, a theoretical model which provides the basis to the very effective working code package CCM

  7. 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)

  8. Liquid argon as active shielding and coolant for bare germanium detectors. A novel background suppression method for the GERDA 0νββ experiment

    International Nuclear Information System (INIS)

    Peiffer, J.P.

    2007-01-01

    Two of the most important open questions in particle physics are whether neutrinos are their own anti-particles (Majorana particles) as required by most extensions of the StandardModel and the absolute values of the neutrino masses. The neutrinoless double beta (0νββ) decay, which can be investigated using 76 Ge (a double beta isotope), is the most sensitive probe for these properties. There is a claim for an evidence for the 0νββ decay in the Heidelberg-Moscow (HdM) 76 Ge experiment by a part of the HdM collaboration. The new 76 Ge experiment Gerda aims to check this claim within one year with 15 kg.y of statistics in Phase I at a background level of ≤10 -2 events/(kg.keV.y) and to go to higher sensitivity with 100 kg.y of statistics in Phase II at a background level of ≤10 -3 events/(kg.keV.y). In Gerda bare germanium semiconductor detectors (enriched in 76 Ge) will be operated in liquid argon (LAr). LAr serves as cryogenic coolant and as high purity shielding against external background. To reach the background level for Phase II, new methods are required to suppress the cosmogenic background of the diodes. The background from cosmogenically produced 60 Co is expected to be ∝2.5.10 -3 events/(kg.keV.y). LAr scintillates in UV (λ=128 nm) and a novel concept is to use this scintillation light as anti-coincidence signal for background suppression. In this work the efficiency of such a LAr scintillation veto was investigated for the first time. In a setup with 19 kg active LAr mass a suppression of a factor 3 has been achieved for 60 Co and a factor 17 for 232 Th around Q ββ = 2039 keV. This suppression will further increase for a one ton active volume (factor O(100) for 232 Th and 60 Co). LAr scintillation can also be used as a powerful tool for background diagnostics. For this purpose a new, very stable and robust wavelength shifter/reflector combination for the light detection has been developed, leading to a photo electron (pe) yield of as much as

  9. Liquid argon as active shielding and coolant for bare germanium detectors. A novel background suppression method for the GERDA 0{nu}{beta}{beta} experiment

    Energy Technology Data Exchange (ETDEWEB)

    Peiffer, J.P.

    2007-07-25

    Two of the most important open questions in particle physics are whether neutrinos are their own anti-particles (Majorana particles) as required by most extensions of the StandardModel and the absolute values of the neutrino masses. The neutrinoless double beta (0{nu}{beta}{beta}) decay, which can be investigated using {sup 76}Ge (a double beta isotope), is the most sensitive probe for these properties. There is a claim for an evidence for the 0{nu}{beta}{beta} decay in the Heidelberg-Moscow (HdM) {sup 76}Ge experiment by a part of the HdM collaboration. The new {sup 76}Ge experiment Gerda aims to check this claim within one year with 15 kg.y of statistics in Phase I at a background level of {<=}10{sup -2} events/(kg.keV.y) and to go to higher sensitivity with 100 kg.y of statistics in Phase II at a background level of {<=}10{sup -3} events/(kg.keV.y). In Gerda bare germanium semiconductor detectors (enriched in {sup 76}Ge) will be operated in liquid argon (LAr). LAr serves as cryogenic coolant and as high purity shielding against external background. To reach the background level for Phase II, new methods are required to suppress the cosmogenic background of the diodes. The background from cosmogenically produced {sup 60}Co is expected to be {proportional_to}2.5.10{sup -3} events/(kg.keV.y). LAr scintillates in UV ({lambda}=128 nm) and a novel concept is to use this scintillation light as anti-coincidence signal for background suppression. In this work the efficiency of such a LAr scintillation veto was investigated for the first time. In a setup with 19 kg active LAr mass a suppression of a factor 3 has been achieved for {sup 60}Co and a factor 17 for {sup 232}Th around Q{sub {beta}}{sub {beta}} = 2039 keV. This suppression will further increase for a one ton active volume (factor O(100) for {sup 232}Th and {sup 60}Co). LAr scintillation can also be used as a powerful tool for background diagnostics. For this purpose a new, very stable and robust wavelength

  10. Secondary coolant purification system

    International Nuclear Information System (INIS)

    Stiteler, F.Z.; Donohue, J.P.

    1978-01-01

    The present invention combines the attributes of volatile chemical addition, continuous blowdown, and full flow condensate demineralization. During normal plant operation (defined as no primary to secondary leakage) condensate from the condenser is pumped through a full flow condensate demineralizer system by the condensate pumps. Volatile chemical additions are made. Dissolved and suspended solids are removed in the condensate polishers by ion exchange and/or filtration. At the same time a continuous blowdown of approximately 1 percent of the main steaming rate of the steam generators is maintained. Radiation detectors monitor the secondary coolant. If these monitors indicate no primary to secondary leakage, the blowdown is cooled and returned directly to the condensate pump discharge. If one of the radiation monitors should indicate a primary to secondary leak, when the temperature of the effluent exiting from the blowdown heat exchanger is compatible with the resin specifications of the ion exchangers, the bypass valve causes the blowdown flow to pass through the blowdown ion exchangers

  11. Decontamination of main coolant pumps

    International Nuclear Information System (INIS)

    Roofthooft, R.

    1988-01-01

    Last year a number of main coolant pumps in Belgian nuclear power plants were decontaminated. A new method has been developed to reduce the time taken for decontamination and the volume of waste to be treated. The method comprises two phases: Oxidation with permanganate in nitric acid and dissolution in oxalic acid. The decontamination of main coolant pumps can now be achieved in less than one day. The decontamination factors attained range between 15 and 150. (orig.) [de

  12. Butyrate reduces appetite and activates brown adipose tissue via the gut-brain neural circuit.

    Science.gov (United States)

    Li, Zhuang; Yi, Chun-Xia; Katiraei, Saeed; Kooijman, Sander; Zhou, Enchen; Chung, Chih Kit; Gao, Yuanqing; van den Heuvel, José K; Meijer, Onno C; Berbée, Jimmy F P; Heijink, Marieke; Giera, Martin; Willems van Dijk, Ko; Groen, Albert K; Rensen, Patrick C N; Wang, Yanan

    2017-11-03

    Butyrate exerts metabolic benefits in mice and humans, the underlying mechanisms being still unclear. We aimed to investigate the effect of butyrate on appetite and energy expenditure, and to what extent these two components contribute to the beneficial metabolic effects of butyrate. Acute effects of butyrate on appetite and its method of action were investigated in mice following an intragastric gavage or intravenous injection of butyrate. To study the contribution of satiety to the metabolic benefits of butyrate, mice were fed a high-fat diet with butyrate, and an additional pair-fed group was included. Mechanistic involvement of the gut-brain neural circuit was investigated in vagotomised mice. Acute oral, but not intravenous, butyrate administration decreased food intake, suppressed the activity of orexigenic neurons that express neuropeptide Y in the hypothalamus, and decreased neuronal activity within the nucleus tractus solitarius and dorsal vagal complex in the brainstem. Chronic butyrate supplementation prevented diet-induced obesity, hyperinsulinaemia, hypertriglyceridaemia and hepatic steatosis, largely attributed to a reduction in food intake. Butyrate also modestly promoted fat oxidation and activated brown adipose tissue (BAT), evident from increased utilisation of plasma triglyceride-derived fatty acids. This effect was not due to the reduced food intake, but explained by an increased sympathetic outflow to BAT. Subdiaphragmatic vagotomy abolished the effects of butyrate on food intake as well as the stimulation of metabolic activity in BAT. Butyrate acts on the gut-brain neural circuit to improve energy metabolism via reducing energy intake and enhancing fat oxidation by activating BAT. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2017. All rights reserved. No commercial use is permitted unless otherwise expressly granted.

  13. A comparative neutronic analysis of KALIMER breeder core using Na or Pb-Bi coolant

    International Nuclear Information System (INIS)

    Yoo, J. W.; Kim, S. J.; Kim, Y. I.

    2000-01-01

    A comparative neutronic study has been conducted on KALIMER breeder core according to the replacement of sodium coolant by Pb-Bi coolant. Since the atomic weight of Pb and Bi is about 9 times heavier than that of Na, the energy loss by neutron colliding with Pb-Bi nucleus will be very small. Therefore, the reactor with Pb-Bi coolant will have a harder neutron spectrum than that with Na coolant. Consequently, the breeding ratio and burnup reactivity swing is expected to be enhanced. In addition, when Pb-Bi coolant is voided, a negative coolant void coefficient can be obtained by the net effects of smaller spectrum hardening and large neutron leakage. As a result, the breeding ratio was increased from 1.18 to 1.23 and burnup reactivity swing was reduced from 631 pcm to 150 pcm. When the coolant in the whole region of active core is voided, the coolant void coefficient was found to be -539 and -264 pcm at BOEC and EOEC, respectively. In the local voided case, the smaller coolant void coefficient was obtained than that of Na coolant. Accordingly, the use of Pb-Bi coolant in KALIMER gives an advantage of higher breeding ratio, smaller burnup reactivity swing and negative coolant void coefficient without any significant degradation of nuclear performance

  14. DQ reference frame modeling and control of single-phase active power decoupling circuits

    DEFF Research Database (Denmark)

    Tang, Yi; Qin, Zian; Blaabjerg, Frede

    2015-01-01

    . This paper presents the dq synchronous reference frame modeling of single-phase power decoupling circuits and a complete model describing the dynamics of dc-link ripple voltage is presented. The proposed model is universal and valid for both inductive and capacitive decoupling circuits, and the input...... of decoupling circuits can be either dependent or independent of its front-end converters. Based on this model, a dq synchronous reference frame controller is designed which allows the decoupling circuit to operate in two different modes because of the circuit symmetry. Simulation and experimental results...... are presented to verify the effectiveness of the proposed modeling and control method....

  15. 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)

  16. Advances in Forecasting and Prevention of Resonances Between Coolant Acoustical Oscillations and Fuel Rod Vibrations

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, Konstantin Nicolaevich [NPP, NPEI, 14, Krasnokazarmennaya str. Moscow, 111250 (Russian Federation)

    2009-06-15

    would be important for NPP life time management purposes. In a similar way it is possible to lead estimation of EFCPO, Q - factors of coolant acoustic oscillatory circuit and PBF for any of updating NPP with PWR including NPP with supercritical parameters. Certainly, the quantitative characteristics of EFCPO, Q - factors and PBF will be various for each class of the nuclear reactor. Paper shows what operating control influences are necessary to remove EFCPO from area of resonant interaction with vibrations FR, FA etc. It is offered to use instrumentation and control systems to prevent operating of NPP at capacity level which provides increasing in amplitudes of pulsations of pressure. The increase in demand of the safety of NPP requires further increase of adequacy between a model and an object. The integrated PSB-VVER test facility is the 1:300 replica of the prototype reactor VVER with respect to power capacity and volume. The height evaluations of the test facility are the same as those of the original. The maximum power of heat released by an assembly of fuel rod simulators is 10 MW. PSB-VVER consists of four loops closed to the reactor model; the latter consists of a down comer section with the lower mixing chamber, a model of the reactor core (a channel with fuel rod simulators), a bypass of the reactor core model, and the upper mixing chamber. Each loop contains a reactor coolant pump, a steam generator, and a cold and hot pipeline. The test facility also includes a pressurizer and an ECCS consisting of three subsystems: a passive one, which incorporates four hydro accumulators and two active ones (a high-pressure ECCS and a low pressure ECCS). Test facility description, scheme and the measuring system are presented. Using such systems the transient processes have been investigated in accident with loss of coolant from the primary cooling system. The basic mathematical models for calculation of EFCPO are achieved. These models are intended for both one-phase and

  17. 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)

  18. Exposure to Forced Swim Stress Alters Local Circuit Activity and Plasticity in the Dentate Gyrus of the Hippocampus

    Directory of Open Access Journals (Sweden)

    Mouna Maroun

    2008-02-01

    Full Text Available Studies have shown that, depending on its severity and context, stress can affect neural plasticity. Most related studies focused on synaptic plasticity and long-term potentiation (LTP of principle cells. However, evidence suggests that following high-frequency stimulation, which induces LTP in principal cells, modifications also take place at the level of complex interactions with interneurons within the dentate gyrus, that is, at the local circuit level. So far, the possible effects of stress on local circuit activity and plasticity were not studied. Therefore, we set out to examine the possible alterations in local circuit activity and plasticity following exposure to stress. Local circuit activity and plasticity were measured by using frequency dependant inhibition (FDI and commissural modulation protocols following exposure to a 15 minute-forced swim trial. Exposure to stress did not alter FDI. The application of theta-burst stimulation (TBS reduced FDI in both control and stressed rats, but this type of plasticity was greater in stressed rats. Commissural-induced inhibition was significantly higher in stressed rats both before and after applying theta-burst stimulation. These findings indicate that the exposure to acute stress affects aspects of local circuit activity and plasticity in the dentate gyrus. It is possible that these alterations underlie some of the behavioral consequences of the stress experience.

  19. Exposure to Forced Swim Stress Alters Local Circuit Activity and Plasticity in the Dentate Gyrus of the Hippocampus

    Science.gov (United States)

    Yarom, Orli; Maroun, Mouna; Richter-Levin, Gal

    2008-01-01

    Studies have shown that, depending on its severity and context, stress can affect neural plasticity. Most related studies focused on synaptic plasticity and long-term potentiation (LTP) of principle cells. However, evidence suggests that following high-frequency stimulation, which induces LTP in principal cells, modifications also take place at the level of complex interactions with interneurons within the dentate gyrus, that is, at the local circuit level. So far, the possible effects of stress on local circuit activity and plasticity were not studied. Therefore, we set out to examine the possible alterations in local circuit activity and plasticity following exposure to stress. Local circuit activity and plasticity were measured by using frequency dependant inhibition (FDI) and commissural modulation protocols following exposure to a 15 minute-forced swim trial. Exposure to stress did not alter FDI. The application of theta-burst stimulation (TBS) reduced FDI in both control and stressed rats, but this type of plasticity was greater in stressed rats. Commissural-induced inhibition was significantly higher in stressed rats both before and after applying theta-burst stimulation. These findings indicate that the exposure to acute stress affects aspects of local circuit activity and plasticity in the dentate gyrus. It is possible that these alterations underlie some of the behavioral consequences of the stress experience. PMID:18301720

  20. Some experimental justifications of constructions of nuclear reactors with the use of solid coolant

    International Nuclear Information System (INIS)

    Deniskin, V.; Nalivaev, V.; Fedik, I.; Vishnevski, U.; Dmitriev, A.

    2003-01-01

    the solid coolant are: 1. Pressure in the primary circuit of the reactor is below the atmospheric one and, as a consequence, there is small steel intensity and cost of the facility. There is a possibility to build a large-power capacity reactor with a low specific power density of the core and high critical margins, which would spare efforts and money to manufacture a complicated and costly equipment and augmented equipment. It means that it is feasible to reduce a possibility of emergency state, to augment safety during all possible accidents, including depressurisation. 2. High temperature of the reactor primary circuit allows obtaining a high thermal efficiency coefficient. 3. A circumstance of importance is that there are no practically any corrosion related problems while using the solid coolant and the erosion issue may be minimised. In its turn, this means that the system for the coolant treatment and recovery may be simple in design, cheap and cost-efficient in operation. 4. The reactor plant may be designed in such a way that its cost and dismalting complexity would be significantly lower than that of existing PWRs. Radioactive waste generated in the course of dismalting of such a reactor would have a specific radioactivity level and total radioactivity hundreds of times less than that of the existing reactor systems. This does not pose a problem with building a new reactor on the decommissioned site and allows reduction of the number of NPP sites. 5. Such reactor practically does not generate liquid waste, and degasifiers may dispose of the minimum amount of gaseous waste generated. The solid low activity operational waste does not incur large storage costs. 6. The reactor will have good neutron and physical properties. (author)

  1. Magneto Rheological Semi-Active Damper with External By-pass Circuit in Modular Structure

    Directory of Open Access Journals (Sweden)

    Alexandru Boltoşi

    2010-10-01

    Full Text Available In order to perform experimentally studies, in the paper it is presented a simple method which was elaborated to realize reliable, at low cost and reproducible semi-active dampers with magnetorheological fluids, having external magnetic circuit. The main components are common constitutive elements of industrial hydraulic and pneumatic drivers, having the supplementary advantages being manufactured in a large scale of overall dimensions and demanding minimal modifications. As accumulator, a similar type of hydraulic or pneumatic cylinder was used. The work of the whole damper can be optimized by modifying the nitrogen pressure and interior volume of accumulator. Another important advantage of this conception is the possibility to realize a modular structure composed by the damper, accumulator and magnetic field generator, interconnected by flexible elements.

  2. Microwave amplifier and active circuit design using the real frequency technique

    CERN Document Server

    Jarry, Pierre

    2016-01-01

    This book focuses on the authors' Real Frequency Technique (RFT) and its application to a wide variety of multi-stage microwave amplifiers and active filters, and passive equalizers for radar pulse shaping and antenna return loss applications. The first two chapters review the fundamentals of microwave amplifier design and provide a description of the RFT. Each subsequent chapter introduces a new type of amplifier or circuit design, reviews its design problems, and explains how the RFT can be adapted to solve these problems. The authors take a practical approach by summarizing the design steps and giving numerous examples of amplifier realizations and measured responses. Provides a complete description of the RFT as it is first used to design multistage lumped amplifiers using a progressive optimization of the equalizers, leading to a small umber of parameters to optimize simultaneously Presents modifications to the RFT to design trans-impedance microwave amplifiers that are used for photodiodes acti...

  3. 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

  4. Sensitivity analysis of an LCL-filter-based three-phase active rectifier via a virtual circuit approach

    DEFF Research Database (Denmark)

    Blaabjerg, Frede; Chiarantoni, Ernesto; Aquila, Antonio Dell’

    2004-01-01

    Three-phase active rectifiers based on the voltage source converter topology can successfully replace traditional thyristor based rectifiers or diode bridge plus chopper in interfacing dc-systems to the grid. However, if the application in which they are employed has a high safety issue......, to the grid side stiffness and to the parameters of the controller has never been detailed considered. In this paper the experimental results of an LCL-filter-based three-phase active rectifier are analysed with the circuit theory approach. A ?virtual circuit? is synthesized in role of the digital controller...

  5. Reactor coolant pumps for nuclear reactors

    International Nuclear Information System (INIS)

    Harand, E.; Richter, G.; Tschoepel, G.

    1975-01-01

    A brake for the pump rotor of a main coolant pump or a shutoff member on the pump are provided in order to prevent excess speeds of the pump rotor. Such excess speeds may occur in PWR type reactors with water at a pressure below, e.g., 150 bars if there is leakage from a coolant line associated with the main coolant pump. As a brake, a centrifugal brake depending upon the pump speed or a brake ring arranged on the pump housing and acting on the pump rotor, which ring would be activated by pressure differentials in the pump, may be used. If the pressure differences between suction and pressure sockets are very small, a controlled hydraulic increase of the pressure force on the brake may also be provided. Furthermore, a turbine brake may be provided. A slide which is automatically movable in closing position along the pump rotor axis is used as a shutoff element. It is of cylindrical configuration and is arranged concentrically with the rotor axis. (DG) [de

  6. Reactor coolant pump seal leakage monitoring

    International Nuclear Information System (INIS)

    Stevens, D.M.; Spencer, J.W.; Morris, D.J.; James, W.; Shugars, H.G.

    1986-01-01

    Problems with reactor coolant pump seals have historically accounted for a large percentage of unscheduled outages. Studies performed for the Electric Power Research Institute (EPRI) have shown that the replacement of coolant pump seals has been one of the leading causes of nuclear plant unavailability over the last ten years. Failures of coolant pump seals can lead to primary coolant leakage rates of 200-500 gallons per minute into the reactor building. Airborne activity and high surface contamination levels following these failures require a major cleanup effort and increases the time and personnel exposure required to refurbish the pump seals. One of the problems in assessing seal integrity is the inability to accurately measure seal leakage. Because seal leakage flow is normally very small, it cannot be sensed directly with normal flow instrumentation, but must be inferred from several other temperature and flow measurements. In operating plants the leakage rate has been quantified with a tipping-bucket gauge, a device which indicates when one quart of water has been accumulated. The tipping-bucket gauge has been used for most rainfall-intensity monitoring. The need for a more accurate and less expensive gauge has been addressed. They have developed a drop-counter precipitation sensor has been developed and optimized. The applicability of the drop-counter device to the problem of measuring seal leakage is being investigated. If a review of system specification and known drop-counter performance indicates that this method is feasible for measuring seal leak rates, a drop-counter gauge will be fabricated and tested in the laboratory. If laboratory tests are successful the gauge will be demonstrated in a pump test loop at Ontario Hydro and evaluated under simulated plant conditions. 3 references, 2 figures

  7. Reactor coolant pump transportation incident

    International Nuclear Information System (INIS)

    Noce, D.

    1992-01-01

    This paper reports on an incident, which occurred on August 27, 1991, in which a Reactor Coolant Pump motor en route from Surry Power Station to Westinghouse repair facilities struck the overpass at the junction of Interstate 64 and Jefferson Avenue in Newport News, Virginia. The transport container that housed the reactor coolant pump motor failed to clear the overpass. The force of the impact dislodged the container and motor from the truck bed, and it landed on the acceleration land and road shoulder. Upon impact, the container broke open and exposed the reactor coolant pump motor. Incidental radioactively contaminated water that remained in the motor coolers drained onto the road, contaminating the aggregate as well as the underlying gravel

  8. 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

  9. Organization of Functional Long-Range Circuits Controlling the Activity of Serotonergic Neurons in the Dorsal Raphe Nucleus

    Directory of Open Access Journals (Sweden)

    Li Zhou

    2017-03-01

    Full Text Available Serotonergic neurons play key roles in various biological processes. However, circuit mechanisms underlying tight control of serotonergic neurons remain largely unknown. Here, we systematically investigated the organization of long-range synaptic inputs to serotonergic neurons and GABAergic neurons in the dorsal raphe nucleus (DRN of mice with a combination of viral tracing, slice electrophysiological, and optogenetic techniques. We found that DRN serotonergic neurons and GABAergic neurons receive largely comparable synaptic inputs from six major upstream brain areas. Upon further analysis of the fine functional circuit structures, we found both bilateral and ipsilateral patterns of topographic connectivity in the DRN for the axons from different inputs. Moreover, the upstream brain areas were found to bidirectionally control the activity of DRN serotonergic neurons by recruiting feedforward inhibition or via a push-pull mechanism. Our study provides a framework for further deciphering the functional roles of long-range circuits controlling the activity of serotonergic neurons in the DRN.

  10. Anabolic steroids alter the physiological activity of aggression circuits in the lateral anterior hypothalamus.

    Science.gov (United States)

    Morrison, T R; Sikes, R W; Melloni, R H

    2016-02-19

    Syrian hamsters exposed to anabolic/androgenic steroids (AAS) during adolescence consistently show increased aggressive behavior across studies. Although the behavioral and anatomical profiles of AAS-induced alterations have been well characterized, there is a lack of data describing physiological changes that accompany these alterations. For instance, behavioral pharmacology and neuroanatomical studies show that AAS-induced changes in the vasopressin (AVP) neural system within the latero-anterior hypothalamus (LAH) interact with the serotonin (5HT) and dopamine (DA) systems to modulate aggression. To characterize the electrophysiological profile of the AAS aggression circuit, we recorded LAH neurons in adolescent male hamsters in vivo and microiontophoretically applied agonists and antagonists of aggressive behavior. The interspike interval (ISI) of neurons from AAS-treated animals correlated positively with aggressive behaviors, and adolescent AAS exposure altered parameters of activity in regular firing neurons while also changing the proportion of neuron types (i.e., bursting, regular, irregular). AAS-treated animals had more responsive neurons that were excited by AVP application, while cells from control animals showed the opposite effect and were predominantly inhibited by AVP. Both DA D2 antagonists and 5HT increased the firing frequency of AVP-responsive cells from AAS animals and dual application of AVP and D2 antagonists doubled the excitatory effect of AVP or D2 antagonist administration alone. These data suggest that multiple DA circuits in the LAH modulate AAS-induced aggressive responding. More broadly, these data show that multiple neurochemical interactions at the neurophysiological level are altered by adolescent AAS exposure. Copyright © 2015 IBRO. Published by Elsevier Ltd. All rights reserved.

  11. Successful reconstruction of a physiological circuit with known connectivity from spiking activity alone.

    Directory of Open Access Journals (Sweden)

    Felipe Gerhard

    Full Text Available Identifying the structure and dynamics of synaptic interactions between neurons is the first step to understanding neural network dynamics. The presence of synaptic connections is traditionally inferred through the use of targeted stimulation and paired recordings or by post-hoc histology. More recently, causal network inference algorithms have been proposed to deduce connectivity directly from electrophysiological signals, such as extracellularly recorded spiking activity. Usually, these algorithms have not been validated on a neurophysiological data set for which the actual circuitry is known. Recent work has shown that traditional network inference algorithms based on linear models typically fail to identify the correct coupling of a small central pattern generating circuit in the stomatogastric ganglion of the crab Cancer borealis. In this work, we show that point process models of observed spike trains can guide inference of relative connectivity estimates that match the known physiological connectivity of the central pattern generator up to a choice of threshold. We elucidate the necessary steps to derive faithful connectivity estimates from a model that incorporates the spike train nature of the data. We then apply the model to measure changes in the effective connectivity pattern in response to two pharmacological interventions, which affect both intrinsic neural dynamics and synaptic transmission. Our results provide the first successful application of a network inference algorithm to a circuit for which the actual physiological synapses between neurons are known. The point process methodology presented here generalizes well to larger networks and can describe the statistics of neural populations. In general we show that advanced statistical models allow for the characterization of effective network structure, deciphering underlying network dynamics and estimating information-processing capabilities.

  12. Low-Noise Active Decoupling Circuit and its Application to 13C Cryogenic RF Coils at 3T

    DEFF Research Database (Denmark)

    Sanchez, Juan Diego; Søvsø Szocska Hansen, Esben; Laustsen, Christoffer

    2017-01-01

    We analyze the loss contributions in a small, 50-mm-diameter receive-only coil for carbon-13 (13C) magnetic resonance imaging at 3 T for 3 different circuits, which, including active decoupling, are compared in terms of their Q-factors and signal-to-noise ratio (SNR). The results show that a circ......We analyze the loss contributions in a small, 50-mm-diameter receive-only coil for carbon-13 (13C) magnetic resonance imaging at 3 T for 3 different circuits, which, including active decoupling, are compared in terms of their Q-factors and signal-to-noise ratio (SNR). The results show...... that a circuit using unsegmented tuning and split matching capacitors can provide 20% SNR enhancement at room temperature compared with that using more traditional designs. The performance of the proposed circuit was also measured when cryogenically cooled to 105 K, and an additional 1.6-fold SNR enhancement...... was achieved on a phantom. The enhanced circuit performance is based on the low capacitance needed to match to 50 when coil losses are low, which significantly reduces the proportion of the current flowing through the matching network and therefore minimizes this loss contribution. This effect makes...

  13. Radioisotopes in the primary circuit of a fast reactor

    International Nuclear Information System (INIS)

    Berlin, M.; Cauvin, M.

    1976-01-01

    In the frame of the research performed to understand the behaviour of the radioactive isotopes of iodine in the primary coolant circuit of fast reactor, a simple theoretical model is proposed. Results concerning PHENIX and RAPSODIE are given

  14. Development of Coolant Radioactivity Interpretation Code

    International Nuclear Information System (INIS)

    Kim, Kiyoung; Jung, Youngsuk; Kim, Kyounghyun; Kim, Jangwook

    2013-01-01

    In Korea, the coolant radioactivity analysis has been performed by using the computer codes of foreign companies such as CADE (Westinghouse), IODYNE and CESIUM (ABB-CE). However, these computer codes are too conservative and have involved considerable errors. Furthermore, since these codes are DOS-based program, their easy operability is not satisfactory. Therefore it is required development of an enhanced analysis algorithm applying an analytical method reflecting the change of operational environments of domestic nuclear power plants and a fuel failure evaluation software considering user' conveniences. We have developed a nuclear fuel failure evaluation code able to estimate the number of failed fuel rods and the burn-up of failed fuels during nuclear power plant operation cycle. A Coolant Radio-activity Interpretation Code (CRIC) for LWR has been developed as the output of the project 'Development of Fuel Reliability Enhanced Technique' organized by Korea Institute of Energy Technology Evaluation and Planning (KETEP). The CRIC is Windows based-software able to evaluate the number of failed fuel rods and the burn-up of failed fuel region by analyzing coolant radioactivity of LWR in operation. The CRIC is based on the model of fission products release commonly known as 'three region model' (pellet region, gap region, and coolant region), and we are verifying the CRIC results based on the cases of domestic fuel failures. CRIC users are able to estimate the number of failed fuel rods, burn-up and regions of failed fuel considered enrichment and power distribution of fuel region by using operational cycle data, coolant activity data, fuel loading pattern, Cs-134/Cs-137 ratio according to burn-up and U-235 enrichment provided in the code. Due to development of the CRIC, it is secured own unique fuel failure evaluation code. And, it is expected to have the following significant meaning. This is that the code reflecting a proprietary technique for quantitatively

  15. The Electron Runaround: Understanding Electric Circuit Basics through a Classroom Activity

    Science.gov (United States)

    Singh, Vandana

    2010-01-01

    Several misconceptions abound among college students taking their first general physics course, and to some extent pre-engineering physics students, regarding the physics and applications of electric circuits. Analogies used in textbooks, such as those that liken an electric circuit to a piped closed loop of water driven by a water pump, do not…

  16. 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.)

  17. 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

  18. On-chip active gate bias circuit for MMIC amplifier applications with 100% threshold voltage variation compensation

    NARCIS (Netherlands)

    Hek, A.P. de; Busking, E.B.

    2006-01-01

    In this paper the design and performance of an on-chip active gate bias circuit for application in MMIC amplifiers, which gives 100% compensation for threshold variation and at the same time is insensitive to supply voltage variations, is discussed. Design equations have been given. In addition, the

  19. Optimized circuit design for flexible 8-bit RFID transponders with active layer of ink-jet printed small molecule semiconductors

    NARCIS (Netherlands)

    Kjellander, B.K.C.; Smaal, W.T.T.; Myny, K.; Genoe, J.; Dehaene, W.; Heremans, P.; Gelinck, G.H.

    2013-01-01

    We ink-jet print a blend of 6,13-bis(triisopropyl-silylethynyl)pentacene and polystyrene as the active layer for flexible circuits. The discrete ink-jet printed transistors exhibit a saturation mobility of 0.5 cm2 V -1 s-1. The relative spread in transistor characteristics can be very large. This

  20. 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

  1. Vertical reactor coolant pump instabilities

    International Nuclear Information System (INIS)

    Jones, R.M.

    1985-01-01

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

  2. Altered regional and circuit resting-state activity associated with unilateral hearing loss.

    Directory of Open Access Journals (Sweden)

    Xingchao Wang

    Full Text Available The deprivation of sensory input after hearing damage results in functional reorganization of the brain including cross-modal plasticity in the sensory cortex and changes in cognitive processing. However, it remains unclear whether partial deprivation from unilateral auditory loss (UHL would similarly affect the neural circuitry of cognitive processes in addition to the functional organization of sensory cortex. Here, we used resting-state functional magnetic resonance imaging to investigate intrinsic activity in 34 participants with UHL from acoustic neuroma in comparison with 22 matched normal controls. In sensory regions, we found decreased regional homogeneity (ReHo in the bilateral calcarine cortices in UHL. However, there was an increase of ReHo in the right anterior insular cortex (rAI, the key node of cognitive control network (CCN and multimodal sensory integration, as well as in the left parahippocampal cortex (lPHC, a key node in the default mode network (DMN. Moreover, seed-based resting-state functional connectivity analysis showed an enhanced relationship between rAI and several key regions of the DMN. Meanwhile, lPHC showed more negative relationship with components in the CCN and greater positive relationship in the DMN. Such reorganizations of functional connectivity within the DMN and between the DMN and CCN were confirmed by a graph theory analysis. These results suggest that unilateral sensory input damage not only alters the activity of the sensory areas but also reshapes the regional and circuit functional organization of the cognitive control network.

  3. Curtailing effect of awakening on visual responses of cortical neurons by cholinergic activation of inhibitory circuits.

    Science.gov (United States)

    Kimura, Rui; Safari, Mir-Shahram; Mirnajafi-Zadeh, Javad; Kimura, Rie; Ebina, Teppei; Yanagawa, Yuchio; Sohya, Kazuhiro; Tsumoto, Tadaharu

    2014-07-23

    Visual responsiveness of cortical neurons changes depending on the brain state. Neural circuit mechanism underlying this change is unclear. By applying the method of in vivo two-photon functional calcium imaging to transgenic rats in which GABAergic neurons express fluorescent protein, we analyzed changes in visual response properties of cortical neurons when animals became awakened from anesthesia. In the awake state, the magnitude and reliability of visual responses of GABAergic neurons increased whereas the decay of responses of excitatory neurons became faster. To test whether the basal forebrain (BF) cholinergic projection is involved in these changes, we analyzed effects of electrical and optogenetic activation of BF on visual responses of mouse cortical neurons with in vivo imaging and whole-cell recordings. Electrical BF stimulation in anesthetized animals induced the same direction of changes in visual responses of both groups of neurons as awakening. Optogenetic activation increased the frequency of visually evoked action potentials in GABAergic neurons but induced the delayed hyperpolarization that ceased the late generation of action potentials in excitatory neurons. Pharmacological analysis in slice preparations revealed that photoactivation-induced depolarization of layer 1 GABAergic neurons was blocked by a nicotinic receptor antagonist, whereas non-fast-spiking layer 2/3 GABAergic neurons was blocked only by the application of both nicotinic and muscarinic receptor antagonists. These results suggest that the effect of awakening is mediated mainly through nicotinic activation of layer 1 GABAergic neurons and mixed nicotinic/muscarinic activation of layer 2/3 non-fast-spiking GABAergic neurons, which together curtails the visual responses of excitatory neurons. Copyright © 2014 the authors 0270-6474/14/3410122-12$15.00/0.

  4. 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

  5. Inward-rectifying potassium (Kir) channels regulate pacemaker activity in spinal nociceptive circuits during early life

    Science.gov (United States)

    Li, Jie; Blankenship, Meredith L.; Baccei, Mark L.

    2013-01-01

    Pacemaker neurons in neonatal spinal nociceptive circuits generate intrinsic burst-firing and are distinguished by a lower “leak” membrane conductance compared to adjacent, non-bursting neurons. However, little is known about which subtypes of leak channels regulate the level of pacemaker activity within the developing rat superficial dorsal horn (SDH). Here we demonstrate that a hallmark feature of lamina I pacemaker neurons is a reduced conductance through inward-rectifying potassium (Kir) channels at physiological membrane potentials. Differences in the strength of inward rectification between pacemakers and non-pacemakers indicate the presence of functionally distinct Kir currents in these two populations at room temperature. However, Kir currents in both groups showed high sensitivity to block by extracellular Ba2+ (IC50 ~ 10 µM), which suggests the presence of ‘classical’ Kir (Kir2.x) channels in the neonatal SDH. The reduced Kir conductance within pacemakers is unlikely to be explained by an absence of particular Kir2.x isoforms, as immunohistochemical analysis revealed the expression of Kir2.1, Kir2.2 and Kir2.3 within spontaneously bursting neurons. Importantly, Ba2+ application unmasked rhythmic burst-firing in ~42% of non-bursting lamina I neurons, suggesting that pacemaker activity is a latent property of a sizeable population of SDH cells during early life. In addition, the prevalence of spontaneous burst-firing within lamina I was enhanced in the presence of high internal concentrations of free Mg2+, consistent with its documented ability to block Kir channels from the intracellular side. Collectively, the results indicate that Kir channels are key modulators of pacemaker activity in newborn central pain networks. PMID:23426663

  6. Oscillator circuits

    CERN Document Server

    Graf, Rudolf F

    1996-01-01

    This series of circuits provides designers with a quick source for oscillator circuits. Why waste time paging through huge encyclopedias when you can choose the topic you need and select any of the specialized circuits sorted by application?This book in the series has 250-300 practical, ready-to-use circuit designs, with schematics and brief explanations of circuit operation. The original source for each circuit is listed in an appendix, making it easy to obtain additional information.Ready-to-use circuits.Grouped by application for easy look-up.Circuit source listing

  7. Measuring circuits

    CERN Document Server

    Graf, Rudolf F

    1996-01-01

    This series of circuits provides designers with a quick source for measuring circuits. Why waste time paging through huge encyclopedias when you can choose the topic you need and select any of the specialized circuits sorted by application?This book in the series has 250-300 practical, ready-to-use circuit designs, with schematics and brief explanations of circuit operation. The original source for each circuit is listed in an appendix, making it easy to obtain additional information.Ready-to-use circuits.Grouped by application for easy look-up.Circuit source listings

  8. Methods for the prediction and achievement of a low circuit activity

    International Nuclear Information System (INIS)

    Faircloth, R.; Knowles, A.N.

    1975-10-01

    While there are about 60,000 fuel pins in the core of an LWR power station, in a comparable HTR there would be over 10 10 coated particles. This five orders of magnitude disparity in the number of basic fission product retaining units has led to important differences in the approach to the prediction and achievement of a low circuit activity. With the HTR, the very large number of units requires statistical treatment of the failure mechanisms involved in the escape of key fission products from the particles and in the assessment of the distribution in the core of burnup, dose and temperature. The methods created to deal with these factors have proved to be very powerful, leading to a situation in which the distribution of fission products within the core and the probability of experiencing a given release are better known for the HTR than for the metal clad fuel of other thermal systems. This advantage has been exploited and reinforced in the UK studies, first by the development of the technology of particle design (a combination of irradiation testing and theoretical analysis) and second by refining the study of fission product movement. This is a continuing process, but already it has usefully illuminated several aspects of design and operation, ranging from the specification and quality control of the fuel itself to the philosophy of outer containments. This paper is intended to summarise this progress, the account being mainly concerned with normal operation. (author)

  9. Plasticity during Sleep Is Linked to Specific Regulation of Cortical Circuit Activity

    Directory of Open Access Journals (Sweden)

    Niels Niethard

    2017-09-01

    Full Text Available Sleep is thought to be involved in the regulation of synaptic plasticity in two ways: by enhancing local plastic processes underlying the consolidation of specific memories and by supporting global synaptic homeostasis. Here, we briefly summarize recent structural and functional studies examining sleep-associated changes in synaptic morphology and neural excitability. These studies point to a global down-scaling of synaptic strength across sleep while a subset of synapses increases in strength. Similarly, neuronal excitability on average decreases across sleep, whereas subsets of neurons increase firing rates across sleep. Whether synapse formation and excitability is down or upregulated across sleep appears to partly depend on the cell’s activity level during wakefulness. Processes of memory-specific upregulation of synapse formation and excitability are observed during slow wave sleep (SWS, whereas global downregulation resulting in elimination of synapses and decreased neural firing is linked to rapid eye movement sleep (REM sleep. Studies of the excitation/inhibition balance in cortical circuits suggest that both processes are connected to a specific inhibitory regulation of cortical principal neurons, characterized by an enhanced perisomatic inhibition via parvalbumin positive (PV+ cells, together with a release from dendritic inhibition by somatostatin positive (SOM+ cells. Such shift towards increased perisomatic inhibition of principal cells appears to be a general motif which underlies the plastic synaptic changes observed during sleep, regardless of whether towards up or downregulation.

  10. Dextromethorphan mediated bitter taste receptor activation in the pulmonary circuit causes vasoconstriction.

    Science.gov (United States)

    Upadhyaya, Jasbir D; Singh, Nisha; Sikarwar, Anurag S; Chakraborty, Raja; Pydi, Sai P; Bhullar, Rajinder P; Dakshinamurti, Shyamala; Chelikani, Prashen

    2014-01-01

    Activation of bitter taste receptors (T2Rs) in human airway smooth muscle cells leads to muscle relaxation and bronchodilation. This finding led to our hypothesis that T2Rs are expressed in human pulmonary artery smooth muscle cells and might be involved in regulating the vascular tone. RT-PCR was performed to reveal the expression of T2Rs in human pulmonary artery smooth muscle cells. Of the 25 T2Rs, 21 were expressed in these cells. Functional characterization was done by calcium imaging after stimulating the cells with different bitter agonists. Increased calcium responses were observed with most of the agonists, the largest increase seen for dextromethorphan. Previously in site-directed mutational studies, we have characterized the response of T2R1 to dextromethorphan, therefore, T2R1 was selected for further analysis in this study. Knockdown with T2R1 specific shRNA decreased mRNA levels, protein levels and dextromethorphan-induced calcium responses in pulmonary artery smooth muscle cells by up to 50%. To analyze if T2Rs are involved in regulating the pulmonary vascular tone, ex vivo studies using pulmonary arterial and airway rings were pursued. Myographic studies using porcine pulmonary arterial and airway rings showed that stimulation with dextromethorphan led to contraction of the pulmonary arterial and relaxation of the airway rings. This study shows that dextromethorphan, acting through T2R1, causes vasoconstrictor responses in the pulmonary circuit and relaxation in the airways.

  11. Dextromethorphan mediated bitter taste receptor activation in the pulmonary circuit causes vasoconstriction.

    Directory of Open Access Journals (Sweden)

    Jasbir D Upadhyaya

    Full Text Available Activation of bitter taste receptors (T2Rs in human airway smooth muscle cells leads to muscle relaxation and bronchodilation. This finding led to our hypothesis that T2Rs are expressed in human pulmonary artery smooth muscle cells and might be involved in regulating the vascular tone. RT-PCR was performed to reveal the expression of T2Rs in human pulmonary artery smooth muscle cells. Of the 25 T2Rs, 21 were expressed in these cells. Functional characterization was done by calcium imaging after stimulating the cells with different bitter agonists. Increased calcium responses were observed with most of the agonists, the largest increase seen for dextromethorphan. Previously in site-directed mutational studies, we have characterized the response of T2R1 to dextromethorphan, therefore, T2R1 was selected for further analysis in this study. Knockdown with T2R1 specific shRNA decreased mRNA levels, protein levels and dextromethorphan-induced calcium responses in pulmonary artery smooth muscle cells by up to 50%. To analyze if T2Rs are involved in regulating the pulmonary vascular tone, ex vivo studies using pulmonary arterial and airway rings were pursued. Myographic studies using porcine pulmonary arterial and airway rings showed that stimulation with dextromethorphan led to contraction of the pulmonary arterial and relaxation of the airway rings. This study shows that dextromethorphan, acting through T2R1, causes vasoconstrictor responses in the pulmonary circuit and relaxation in the airways.

  12. 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

  13. Radionuclide activity in the secondary cooling circuit of the NPP with WWER-440

    International Nuclear Information System (INIS)

    Voronin, L.M.; Volkov, A.P.; Kozlov, V.F.; Luzanova, L.M.; Pashevich, V.I.

    1979-01-01

    On the basis of the Kol'skaya NPP operating experience the radioactivity of water and steam fed to turbines and boilers has been determined. The radioactivity of water in the primary circuit is demonstrated to be mainly due to 24 Na, 42 K, 18 F, sup(131-135)I, and in the secondary circuit - due to 131 I and 133 I. The data obtained testify to the permanent leakage of water from the primary circuit to the secondary one averaging 19+-12 g/hr at the pressure gradient between the circuits approximately 7.0 MPa. With the probability of 1.2x10 -1 -1.6x10 -2 (eff. reactor day) -1 or 1.2x10 -8 -1.6x10 -9 (kWxhr) -1 the leakage may for some time increase to 1 l/hr

  14. 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…

  15. Specificities of reactor coolant pumps units with lead and lead-bismuth coolant

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Anotonenkov, M.A.; Bokov, P.A.; Baranova, V.S.; Kustov, M.S.

    2009-01-01

    The analysis results of impact of lead and lead-bismuth coolants specific properties on the coolants flow features in flow channels of the main and auxiliary circulating pumps are presented. Impossibility of cavitation initiation in flow channels of vane pumps pumping lead and lead-bismuth coolants was demonstrated. The experimental research results of discontinuity of heavy liquid metal coolant column were presented and conditions of gas cavitation initiation in coolant flow were discussed. Invalidity of traditional calculation methods of water and sodium coolants circulation pumps calculations for lead and lead-bismuth coolants circulation pumps was substantiated [ru

  16. Responses to Small Break Loss of Coolant Accidents for SMART

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Kim, Hee C.; Chang, Moon H.; Zee, Sung Q.; Kim, Si-Hwan; Lee, Un-Chul

    2004-01-01

    The SMART NSSS adopts the design characteristics of containing most of the primary circuit components, such as the reactor core, main coolant pumps (MCPs), steam generators (SGs), and N 2 gas pressurizer (PZR) in a single leak-tight Reactor Pressure Vessel (RPV) with a relatively large ratio of the primary coolant inventory to the core power compared to the conventional loop-type PWR. Due to these design characteristics, the SMART can fundamentally eliminate the possibility of Large Break Loss of Coolant Accidents (LBLOCAs), improve the natural circulation capability, and assure a sufficient time to mitigate the possibility of core uncover. Also, SMART adopts inherent safety improving features and passive engineered safety systems such as the substantially large negative moderator temperature coefficients, passive residual heat removal system, emergency core cooling system, and a steel-made leak-tight Safeguard Vessel (SV) housing the RPV. This paper presents the results of the safety analyses using a MARS/SMR code for the instantaneous guillotine ruptures of the major pipelines penetrating the RPV. The analysis results, employing conservative initial/boundary conditions and assumptions, show that the safety systems of the SMART basic design adequately remove the core decay heat without causing core uncover for all the cases of the Small Break Loss of Coolant Accidents (SBLOCAs). The sensitivity study results with variable SV conditions show that the reduced SV net free volume can shorten the time for reaching the thermal and mechanical equilibrium condition between the RPV and SV. Under these boundary conditions, the primary system inventory loss can be minimized and the core remains covered for a longer period of time without any makeup of the coolant. (authors)

  17. Simulation of steam explosion in stratified melt-coolant configuration

    International Nuclear Information System (INIS)

    Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

    2016-01-01

    Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To

  18. Trace organics in AGR coolants

    International Nuclear Information System (INIS)

    Smith, R.; Green, L.O.; Johnson, P.A.V.

    1980-01-01

    Several analytical techniques have been employed in previous studies of the stable organic compounds arising from the radiolysis of methane/carbon monoxide/carbon dioxide coolants. The majority of this early information was collected from the Windscale AGR prototype. Analyses were also carried out on the liquors obtained from the WAGR humidryers. Three classes of compound were found in the liquors; aliphatic acids in the aqueous phase and methyl ketones and aromatic hydrocarbons in the oily phase. Acetic acid was found to be the predominant carboxylic acid. This paper outlines the major findings from a recent analytical survey of coolants taken over a wide range of dose rate, pressure, temperature and composition, from materials testing reactor facilities, WAGR and CAGR. (author)

  19. NaV1.6a is required for normal activation of motor circuits normally excited by tactile stimulation

    Science.gov (United States)

    Low, Sean E.; Zhou, Weibin; Choong, Xinling; Saint-Amant, Louis; Sprague, Shawn M.; Hirata, Hiromi; Cui, Wilson W.; Hume, Richard I.; Kuwada, John Y.

    2010-01-01

    A screen for zebrafish motor mutants identified two non-complementing alleles of a recessive mutation that were named non-active (navmi89 and navmi130). nav embryos displayed diminished spontaneous and touch-evoked escape behaviors during the first three days of development. Genetic mapping identified the gene encoding NaV1.6a (scn8aa) as a potential candidate for nav. Subsequent cloning of scn8aa from the two alleles of nav uncovered two missense mutations in NaV1.6a that eliminated channel activity when assayed heterologously. Furthermore the injection of RNA encoding wild type scn8aa rescued the nav mutant phenotype indicating that scn8aa was the causative gene of nav. In vivo electrophysiological analysis of the touch-evoked escape circuit indicated that voltage-dependent inward current was decreased in mechanosensory neurons in mutants, but they were able to fire action potentials. Furthermore tactile stimulation of mutants activated some neurons downstream of mechanosensory neurons but failed to activate the swim locomotor circuit in accord with the behavioral response of initial escape contractions but no swimming. Thus mutant mechanosensory neurons appeared to respond to tactile stimulation but failed to initiate swimming. Interestingly fictive swimming could be initiated pharmacologically suggesting that a swim circuit was present in mutants. These results suggested that NaV1.6a was required for touch-induced activation of the swim locomotor network. PMID:20225246

  20. Radionuclide deposits on heat transfer surfaces in a circumt with dissociating N2O4 coolant

    International Nuclear Information System (INIS)

    Dolgov, V.M.; Katanaev, A.O.; Komissarov, F.D.

    1984-01-01

    Radionuclides deposits on heat transfer surfaces of a circuit with dissociating coolant are studied. The areas of preferential deposition of 54 Mn, 51 Cr, 134 Cs and their distribution along the heating and cooling surfaces are determined. The comparison of the obtained data on the nuclide and chemical compositions of the deposits in the areas of N 2 O 4 coolant heating and cooling shows that 54 Mn, 51 Cr, 134 Cs deposit preferentially on heat transfer surfaces in the area of the coolant heating. Fixed and movable deposits consists of the structural material oxides. The quantity of radionuclides in the deposits on the surfaces of heat transfer tubes in the area of cooling decreases with the coolant temperature drop

  1. 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.

  2. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  3. Upgradation of design features of primary coolant pumps of Indian 220 MWe PHWR

    International Nuclear Information System (INIS)

    Sharma, S.S.; Mhetre, S.G.; Manna, M.M.

    1994-01-01

    Evolution in the design features of Primary Coolant Pump (PCP) had started in fifties for catering to stringent specification requirements of reactor coolant systems of larger capacity reactors of various kinds. Primary coolant pumps of PWR and PHWR are employed for circulating radioactive, pressurized hot water in a circuit consisting of reactor (heat source) and steam generator (heat sink). As primary coolant pump capacity decides the station capacity, larger capacity primary coolant pumps have been evolved. Since primary coolant pump pressure containing parts are part of Primary Heat Transport system envelope, the parts are designed, manufactured, inspected and tested in accordance with the applicable system guidelines. Flywheel is mounted on the motor shaft for increasing mass moment of inertia of pump motor rotor to meet the coast down requirements of reactor cooling system under Class-IV electrical power supply failure. Due to limited accessibility of the PCP (PCP installed in shut down accessible area), quick maintenance, condition monitoring, reliable shaft seal system/bearing system aspects have been of great concern to reactor owners and pump manufacturers. In this paper upgradation of design features of RAPS, MAPS and NAPS primary coolant pumps have been covered. (author). 4 figs., 1 tab

  4. The chemistry of the X-7 (organic) loop coolant part I, May 1960 to April 1965

    International Nuclear Information System (INIS)

    Smee, J.L.

    1966-01-01

    The report describes in detail the X-7 coolant chemistry from the start of loop operation in May 1960 to April 1965. During this period the coolant was Santowax OM containing a nominal 30% high boilers or high molecular weight decomposition products. During the first few months of operation it became apparent that there wa.s a serious problem in the fouling of fuel element heat transfer surfaces. This was overcome by continuous purification of the coolant by Attapulgus clay and filters. Since clay purification has been in use, the fouling rate has been less than 0.2 μg.cm -2 .h -1 (10 μm per year), the target value for successful operation of an organic cooled power reactor. Control of the fouling promoter chlorine has been accomplished by completely excluding it from the vicinity of the loop. Any which does get into the coolant is removed by a bed of Mg ribbon and Pd pellets. Since such a bed has been in use, the Cl content of the coolant has been less than 3 ppm. Also given in this report are: (a) a brief history of the loop since its inception in 1959. (b) the effect of the clay column on the coolant chemistry. (c) a complete description of the current purification, degas and make-up circuits, (d) a summary of the coolant chemistry during all fuel irradiations. (author)

  5. 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.)

  6. Installations having pressurised fluid circuits

    International Nuclear Information System (INIS)

    Rigg, S.; Grant, J.

    1977-01-01

    Reference is made to nuclear installations having pressurised coolant flow circuits. Breaches in such circuits may quickly result in much damage to the plant. Devices such as non-return valves, orifice plates, and automatically operated shut-off valves have been provided to prevent or reduce fluid flow through a breached pipe line, but such devices have several disadvantages; they may present large restrictions to normal flow of coolant, and may depend on the operation of ancillary equipment, with consequent delay in bringing them into operation in an emergency. Other expedients that have been adopted to prevent or reduce reverse flow through an upstream breach comprise various forms of hydraulic counter flow brakes. The arrangement described has at least one variable fluid brake comprising a fluidic device connected into a duct in the pressurised circuit, the device having an inlet, an outlet, a vortex chamber between the inlet and outlet, a control jet for introducing fluid into the vortex chamber, connections communicating the inlet and the outlet into one part of the circuit and the control jet into another region at a complementary pressure so that, in the event of a breach in the circuit in one region, fluid passes from the other region to enter the vortex chamber to stimulate pressure to create a flow restricting vortex in the chamber that reduces flow through the breach. The system finds particular application to stream generating pressure tube reactors, such as the steam generating heavy water reactor at UKAEA, Winfrith. (U.K.)

  7. Radioactivity analysis of KAMINI reactor coolant from regulatory perspectives

    International Nuclear Information System (INIS)

    Srinivasan, T.K.; Sulthan, Bajeer; Sarangapani, R.; Jose, M.T.; Venkatraman, B.; Thilagam, L.

    2016-01-01

    KAMINI (a 30kWt) research reactor is operated for neutron radiography of fuel subassemblies and pyro devices and activation analysis of various samples. The reactor is fueled by 233 U and DM water is used as the coolant. During reactor operation, fission product noble gasses (FPNGs) such as 85m Kr, 87 Kr, 88 Kr, 135 Xe, 135m Xe and 138 Xe are detected in the coolant water. In order to detect clad failure, the water is sampled during reactor operation at regular intervals as per the technical specifications. In the present work, analysis of measured activities in coolant samples collected during reactor operation at 25 kWt are presented and compared with computed values obtained using ORIGEN (Isotope Generation) code

  8. 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

  9. Nutritional State-Dependent Ghrelin Activation of Vasopressin Neurons via Retrograde Trans-Neuronal–Glial Stimulation of Excitatory GABA Circuits

    Science.gov (United States)

    Haam, Juhee; Halmos, Katalin C.; Di, Shi

    2014-01-01

    Behavioral and physiological coupling between energy balance and fluid homeostasis is critical for survival. The orexigenic hormone ghrelin has been shown to stimulate the secretion of the osmoregulatory hormone vasopressin (VP), linking nutritional status to the control of blood osmolality, although the mechanism of this systemic crosstalk is unknown. Here, we show using electrophysiological recordings and calcium imaging in rat brain slices that ghrelin stimulates VP neurons in the hypothalamic paraventricular nucleus (PVN) in a nutritional state-dependent manner by activating an excitatory GABAergic synaptic input via a retrograde neuronal–glial circuit. In slices from fasted rats, ghrelin activation of a postsynaptic ghrelin receptor, the growth hormone secretagogue receptor type 1a (GHS-R1a), in VP neurons caused the dendritic release of VP, which stimulated astrocytes to release the gliotransmitter adenosine triphosphate (ATP). ATP activation of P2X receptors excited presynaptic GABA neurons to increase GABA release, which was excitatory to the VP neurons. This trans-neuronal–glial retrograde circuit activated by ghrelin provides an alternative means of stimulation of VP release and represents a novel mechanism of neuronal control by local neuronal–glial circuits. It also provides a potential cellular mechanism for the physiological integration of energy and fluid homeostasis. PMID:24790191

  10. 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.

  11. Development of quench detection/protection system based on active power method for superconducting magnet by using capacitor circuit

    International Nuclear Information System (INIS)

    Nanato, N.; Otsuka, T.; Hesaka, S.; Murase, S.

    2013-01-01

    Highlights: ► The authors have presented an active power method for quench detection. ► A method for improving its characteristics using a capacitor circuit was proposed. ► Quench detection/protection test for a Bi2223 superconducting coil was carried out. ► The proposed method was more useful than the conventional one. -- Abstract: When a quench occurs in a superconducting magnet, excessive joule heating in normal region may damage the magnet. It is necessary to detect the quench as soon as possible and discharge magnetic energy stored in the magnet. The authors have presented a quench detection/protection system based on an active power method which detects the quench regardless of a self-inductive and mutual-inductive voltages and electromagnetic noise. In the conventional active power method, the inductive voltages are removed by cancel coils. In this paper, the authors propose a method to cancel an inductive voltage using a capacitor circuit. The quench detection/protection system becomes more precise and smaller than the conventional system through the capacitor circuit

  12. 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.

  13. 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.

  14. Activity build-up in the primary circuit of pressurized water reactors

    International Nuclear Information System (INIS)

    Sachse, G.; Mittag, I.

    1986-01-01

    Based upon international literature, the following topics are reviewed: research and development efforts; release, transport, and deposition of radioactive corrosion products under primary circuit conditions; experimental results in test and technical systems; possibilities of controlling radiation fields in nuclear power plants by water-chemical measures, decontamination, and high-temperature filtration. (author)

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

    OpenAIRE

    Lefèvre, Grégory; Živković, Ljiljana S.; Jaubertie, Anne

    2012-01-01

    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 suspec...

  16. Design criteria of primary coolant chemistry in SMART-P

    International Nuclear Information System (INIS)

    Choi, Byung Seon; Kim, Ah Young; Kim, Seong Hoon; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    SMART-P differs significantly from commercially designed PWRs. Materials inventories used in SMART-P differ from that at PWRs. All surfaces of the primary circuit with the primary coolant are either made from or plated with stainless steel. The material of steam generator (SG) is also different from that of the standard material of the commercially operating PWRs: titanium alloy for the steam generator tubes. Also, SMART-P primary coolant technology differs from that in PWRs: ammonia is used as a pH raising agent and hydrogen formed due to radiolytic processes is kept in specific range by ammonia dosing. Nevertheless, main objectives of the SMART-P primary coolant are the same as at PWRs: to assure primary system pressure boundary integrity, fuel cladding integrity and to minimize out-of-core radiation buildup. The objective of this work is to introduce the design criteria for the primary water chemistry for SMART-P from the viewpoint of the system characteristics and the chemical design concept

  17. 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

  18. 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

  19. Organization of Functional Long-Range Circuits Controlling the Activity of Serotonergic Neurons in the Dorsal Raphe Nucleus.

    Science.gov (United States)

    Zhou, Li; Liu, Ming-Zhe; Li, Qing; Deng, Juan; Mu, Di; Sun, Yan-Gang

    2017-03-21

    Serotonergic neurons play key roles in various biological processes. However, circuit mechanisms underlying tight control of serotonergic neurons remain largely unknown. Here, we systematically investigated the organization of long-range synaptic inputs to serotonergic neurons and GABAergic neurons in the dorsal raphe nucleus (DRN) of mice with a combination of viral tracing, slice electrophysiological, and optogenetic techniques. We found that DRN serotonergic neurons and GABAergic neurons receive largely comparable synaptic inputs from six major upstream brain areas. Upon further analysis of the fine functional circuit structures, we found both bilateral and ipsilateral patterns of topographic connectivity in the DRN for the axons from different inputs. Moreover, the upstream brain areas were found to bidirectionally control the activity of DRN serotonergic neurons by recruiting feedforward inhibition or via a push-pull mechanism. Our study provides a framework for further deciphering the functional roles of long-range circuits controlling the activity of serotonergic neurons in the DRN. Copyright © 2017 The Author(s). Published by Elsevier Inc. All rights reserved.

  20. 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

  1. Importance of ECP in the prediction of radiation fields in PWR and VVER primary circuits

    International Nuclear Information System (INIS)

    Urquidi-Macdonald, M.; Jacesko, S.L.; Macdonald, Digby D.; Salter-Williams, M.

    2002-01-01

    A model has been developed for predicting mass and activity transport in the primary coolant circuits of PWRs and VVERs with the objective of demonstrating and quantifying the importance of the electrochemical corrosion potential (ECP) in determining the impact of both processes on reactor operation. The model initially employs a radiolysis/mixed potential code to calculate the ECP at four locations (core, hot leg, steam generator, cold leg) and the ECP is then used to estimate the local magnetite solubility. The solubility is then averaged around the loop to yield the ''background'' solubility. Comparison of the background solubility with the local solubility determines whether precipitation or dissolution will occur at any given point in the circuit under any given set of conditions. It is further assumed that the concentration of 59 Co in the coolant is given by the isotopic fraction of this species compared with iron averaged over all materials and weighted by the respective wetted areas. Activation of 59 Co to 60 Co is assumed to occur in the coolant phase by fast, epithermal, and thermal neutron capture. The calculated activity is then used to train an artificial neural network (ANN) to establish relationships between activity at any given location and the operating properties of the reactor, including coolant pH, ECP, temperature, power level, etc. The model predicts that during shut down, magnetite (and hence 59 Co) migrates to the core, where it is irradiated and activated, particularly during subsequent start-up. During start-up, the magnetite (and hence 60 Co) migrates from the core to out-of-core surfaces where it establishes the radiation fields. (authors)

  2. Coolant inlet device for nuclear reactors

    International Nuclear Information System (INIS)

    Ando, Hiroshi; Abe, Yasuhiro; Iwabuchi, Toshihiko; Yamamoto, Kenji.

    1969-01-01

    Herein disclosed is a coolant inlet device for liquid-metal cooled reactors which employs a coolant distributor serving also as a supporting means for the reactor core. The distributor is mounted within the reactor vessel so as to slide horizontally on supporting lugs, and is further slidably connected via a junction pipe to a coolant inlet conduit protruding through the floor of the vessel. The distributor is adapted to uniformly disperse the highly pressured coolant over the reactor core so as to reduce the stresses sustained by the reactor vessel as well as the supporting lugs. Moreover, the slidable nature of the distributor allows thermal shock and excessive coolant pressures to be prevented or alleviated, factors which posed major difficulties in conventional coolant inlet devices. (Owens, K. J.)

  3. 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

  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. Activation in mesolimbic and visuospatial neural circuits elicited by smoking cues: evidence from functional magnetic resonance imaging.

    Science.gov (United States)

    Due, Deborah L; Huettel, Scott A; Hall, Warren G; Rubin, David C

    2002-06-01

    The authors sought to increase understanding of the brain mechanisms involved in cigarette addiction by identifying neural substrates modulated by visual smoking cues in nicotine-deprived smokers. Event-related functional magnetic resonance imaging (fMRI) was used to detect brain activation after exposure to smoking-related images in a group of nicotine-deprived smokers and a nonsmoking comparison group. Subjects viewed a pseudo-random sequence of smoking images, neutral nonsmoking images, and rare targets (photographs of animals). Subjects pressed a button whenever a rare target appeared. In smokers, the fMRI signal was greater after exposure to smoking-related images than after exposure to neutral images in mesolimbic dopamine reward circuits known to be activated by addictive drugs (right posterior amygdala, posterior hippocampus, ventral tegmental area, and medial thalamus) as well as in areas related to visuospatial attention (bilateral prefrontal and parietal cortex and right fusiform gyrus). In nonsmokers, no significant differences in fMRI signal following exposure to smoking-related and neutral images were detected. In most regions studied, both subject groups showed greater activation following presentation of rare target images than after exposure to neutral images. In nicotine-deprived smokers, both reward and attention circuits were activated by exposure to smoking-related images. Smoking cues are processed like rare targets in that they activate attentional regions. These cues are also processed like addictive drugs in that they activate mesolimbic reward regions.

  6. 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

  7. Active energy recovery clamping circuit to improve the performance of power converters

    Science.gov (United States)

    Whitaker, Bret; Barkley, Adam

    2017-05-09

    A regenerative clamping circuit for a power converter using clamping diodes to transfer charge to a clamping capacitor and a regenerative converter to transfer charge out of the clamping capacitor back to the power supply input connection. The regenerative converter uses a switch connected to the midpoint of a series connected inductor and capacitor. The ends of the inductor and capacitor series are connected across the terminals of the power supply to be in parallel with the power supply.

  8. 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

  9. 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...

  10. Low noise signal-to-noise ratio enhancing readout circuit for current-mediated active pixel sensors

    International Nuclear Information System (INIS)

    Ottaviani, Tony; Karim, Karim S.; Nathan, Arokia; Rowlands, John A.

    2006-01-01

    Diagnostic digital fluoroscopic applications continuously expose patients to low doses of x-ray radiation, posing a challenge to both the digital imaging pixel and readout electronics when amplifying small signal x-ray inputs. Traditional switch-based amorphous silicon imaging solutions, for instance, have produced poor signal-to-noise ratios (SNRs) at low exposure levels owing to noise sources from the pixel readout circuitry. Current-mediated amorphous silicon pixels are an improvement over conventional pixel amplifiers with an enhanced SNR across the same low-exposure range, but whose output also becomes nonlinear with increasing dosage. A low-noise SNR enhancing readout circuit has been developed that enhances the charge gain of the current-mediated active pixel sensor (C-APS). The solution takes advantage of the current-mediated approach, primarily integrating the signal input at the desired frequency necessary for large-area imaging, while adding minimal noise to the signal readout. Experimental data indicates that the readout circuit can detect pixel outputs over a large bandwidth suitable for real-time digital diagnostic x-ray fluoroscopy. Results from hardware testing indicate that the minimum achievable C-APS output current that can be discerned at the digital fluoroscopic output from the enhanced SNR readout circuit is 0.341 nA. The results serve to highlight the applicability of amorphous silicon current-mediated pixel amplifiers for large-area flat panel x-ray imagers

  11. A flexible organic active matrix circuit fabricated using novel organic thin film transistors and organic light-emitting diodes

    KAUST Repository

    Gutiérrez-Heredia, Gerardo

    2010-10-04

    We present an active matrix circuit fabricated on plastic (polyethylene naphthalene, PEN) and glass substrates using organic thin film transistors and organic capacitors to control organic light-emitting diodes (OLEDs). The basic circuit is fabricated using two pentacene-based transistors and a capacitor using a novel aluminum oxide/parylene stack (Al2O3/ parylene) as the dielectric for both the transistor and the capacitor. We report that our circuit can deliver up to 15 μA to each OLED pixel. To achieve 200 cd m-2 of brightness a 10 μA current is needed; therefore, our approach can initially deliver 1.5× the required current to drive a single pixel. In contrast to parylene-only devices, the Al2O 3/parylene stack does not fail after stressing at a field of 1.7 MV cm-1 for >10 000 s, whereas \\'parylene only\\' devices show breakdown at approximately 1000 s. Details of the integration scheme are presented. © 2010 IOP Publishing Ltd.

  12. A flexible organic active matrix circuit fabricated using novel organic thin film transistors and organic light-emitting diodes

    KAUST Repository

    Gutié rrez-Heredia, Gerardo; Gonzá lez, Luis A.; Alshareef, Husam N.; Gnade, Bruce E.; Quevedo-Ló pez, Manuel Angel Quevedo

    2010-01-01

    We present an active matrix circuit fabricated on plastic (polyethylene naphthalene, PEN) and glass substrates using organic thin film transistors and organic capacitors to control organic light-emitting diodes (OLEDs). The basic circuit is fabricated using two pentacene-based transistors and a capacitor using a novel aluminum oxide/parylene stack (Al2O3/ parylene) as the dielectric for both the transistor and the capacitor. We report that our circuit can deliver up to 15 μA to each OLED pixel. To achieve 200 cd m-2 of brightness a 10 μA current is needed; therefore, our approach can initially deliver 1.5× the required current to drive a single pixel. In contrast to parylene-only devices, the Al2O 3/parylene stack does not fail after stressing at a field of 1.7 MV cm-1 for >10 000 s, whereas 'parylene only' devices show breakdown at approximately 1000 s. Details of the integration scheme are presented. © 2010 IOP Publishing Ltd.

  13. 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

  14. 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

  15. 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.)

  16. 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.

  17. Passive and active RF-microwave circuits course and exercises with solutions

    CERN Document Server

    Jarry, Pierre

    2015-01-01

    Microwave and radiofrequency (RF) circuits play an important role in communication systems. Due to the proliferation of radar, satellite, and mobile wireless systems, there is a need for design methods that can satisfy the ever increasing demand for accuracy, reliability, and fast development times. This book explores the principal elements for receiving and emitting signals between Earth stations, satellites, and RF (mobile phones) in four parts; the theory and realization of couplers, computation and realization of microwave and RF filters, amplifiers and microwave and RF oscillators. Pas

  18. 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)

  19. Organic coolant in Winnipeg riverbed sediments

    International Nuclear Information System (INIS)

    Guthrie, J.E.; Acres, O.E.

    1979-03-01

    Between January and May 1977 a prolonged leak of organic coolant occurred from the Whiteshell Nuclear Research Establishment's nuclear reactor, and a minimum of 1450 kg of coolant entered the Winnipeg River and was deposited on the riverbed. The level of radioactivity associated with this coolant was low, contributing less than 0.2 μGy (0.02 mrad) a year to the natural background gamma radiation field from the riverbed. The concentration of coolant in the water samples never exceeded 0.02 mg/L, the lower limit of detection. The mortality of crayfish, held in cages where the riverbed was covered with the largest deposits of coolant, was not significantly different from that in the control cages upstream of the outfall. No evidence of fish kill was found. (author)

  20. ECCS control circuit

    International Nuclear Information System (INIS)

    Sato, Takashi.

    1986-01-01

    Purpose: To afford a sufficient margin to pressure vibrations upon starting of an automatic depressurization system by dispersing pressure vibration in suppression water due to the opening action of an automatic releaf valve in the automatic depressurization system thereby reducing the dynamic load exerted to the surface of the suppression walls. Constitution: Upon occurrence of loss of coolant accidents, an automatic releaf valve for automatic depressurization is opened to deliver the steams in the pressure vessel into the suppression pool. Since a plurality of automatic releaf valves have usually been disposed, if they are opened simultaneously, excess dynamic loads are exerted due to the pressure vibrations to the wall surface of the suppression pool. In this invention, a control circuit is disposed such that the opening timing for each of the automatic releaf valves is deviated upon occurrence of a driving signal for the automatic depressurization system to thereby disperse the pressure vibrations in the suppression water. (Kamimura, M.)

  1. 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.

  2. 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)

  3. Components of the primary circuit of LWRs

    International Nuclear Information System (INIS)

    1980-01-01

    This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives. (orig.) [de

  4. Corrosion product behaviour in the primary circuit of the KNK nuclear reactor facility

    International Nuclear Information System (INIS)

    Stamm, H.H.; Stade, K.Ch.

    1976-01-01

    During nuclear operation of the KNK facility from 1972 until September 1974 the composition and behaviour of radionuclides occuring in the primary circuit were investigated. Besides traces of 140 Ba/ 140 La, no fission product activity was detectable in the KNK primary circuit. The fuel element purification from sodium deposits (prior to transport to the reprocessing plant) did not yield any indication of a fuel element failure during KNK-I operation. The activity inventory of the primary loop was exclusively made up of activated corrosion products and 22 Na. The main activity was due to 65 Zn, followed by 54 Mn, 22 Na, sup(110m)Ag, 182 Ta, 60 Co and 124 Sb. It was found that the sorption of 65 Zn and 54 Mn on crucibles made from nickel was condiserably higher than on vessels made from other materials. This observation was confirmed both in tests with material samples from the primary circuit and for disks of gate valves of the primary circuit. sup(110m)Ag did hardly exhibit any sorption effects and had been dissolved largely homogeneously in the hot primary coolant. In the first primary cold trap which was removed from the circuit after some 20,000 hours of operation, only 65 Zn and 54 Mn were detected in addition to traces of 60 Co and 182 Ta. (author)

  5. Nanofluid as coolant for grinding process: An overview

    Science.gov (United States)

    Kananathan, J.; Samykano, M.; Sudhakar, K.; Subramaniam, S. R.; Selavamani, S. K.; Manoj Kumar, Nallapaneni; Keng, Ngui Wai; Kadirgama, K.; Hamzah, W. A. W.; Harun, W. S. W.

    2018-04-01

    This paper reviews the recent progress and applications of nanoparticles in lubricants as a coolant (cutting fluid) for grinding process. The role of grinding machining in manufacturing and the importance of lubrication fluids during material removal are discussed. In grinding process, coolants are used to improve the surface finish, wheel wear, flush the chips and to reduce the work-piece thermal deformation. The conventional cooling technique, i.e., flood cooling delivers a large amount of fluid and mist which hazardous to the environment and humans. Industries are actively looking for possible ways to reduce the volume of coolants used in metal removing operations due to the economical and ecological impacts. Thus as an alternative, an advanced cooling technique known as Minimum Quantity Lubrication (MQL) has been introduced to the enhance the surface finish, minimize the cost, to reduce the environmental impacts and to reduce the metal cutting fluid consumptions. Nanofluid is a new-fangled class of fluids engineered by dispersing nanometre-size solid particles into base fluids such as water, lubrication oils to further improve the properties of the lubricant or coolant. In addition to advanced cooling technique review, this paper also reviews the application of various nanoparticles and their performance in grinding operations. The performance of nanoparticles related to the cutting forces, surface finish, tool wear, and temperature at the cutting zone are briefly reviewed. The study reveals that the excellent properties of the nanofluid can be beneficial in cooling and lubricating application in the manufacturing process.

  6. 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

  7. 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

  8. On steady-state concentrations of ammonia and molecular hydrogen in the primary circuit of the WWER-1000 reactors

    International Nuclear Information System (INIS)

    Arkhipov, O.P.; Bugaenko, V.L.; Kamakchi, S.A.

    1997-01-01

    It is shown that the MORAVA-N2 software package describes well the coolant state in the primary circuit of an actual reactor facility with the WWER-1000 during on-load operation. It permits using the package for analysis of process perturbation effect on the coolant composition. Specific feature of ammonia radiation chemistry in the primary circuit of a reactor facility with the WWER-1000, assuring the rates hydrogen concentration in the coolant with ammonia concentration variation in the coolant within wide limits, when reactor operates on power, can be mentioned by way of example, the fact being ascertained in this study

  9. 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

  10. Behavioral synthesis of asynchronous circuits

    DEFF Research Database (Denmark)

    Nielsen, Sune Fallgaard

    2005-01-01

    This thesis presents a method for behavioral synthesis of asynchronous circuits, which aims at providing a synthesis flow which uses and tranfers methods from synchronous circuits to asynchronous circuits. We move the synchronous behavioral synthesis abstraction into the asynchronous handshake...... is idle. This reduces unnecessary switching activity in the individual functional units and therefore the energy consumption of the entire circuit. A collection of behavioral synthesis algorithms have been developed allowing the designer to perform time and power constrained design space exploration...

  11. Effect of irradiation on corrosion of low-activation austenite Cr-Mn steel in technological liquid mediums of nuclear power plant

    International Nuclear Information System (INIS)

    Demina, E.V.; Prusakova, M.D.; Vinogradova, N.A.; Orlova, G.D.; Nechaev, A.F.; Doil'nitsyn, V.A.

    2008-01-01

    Effect of γ-radiation on corrosion rate in cold-worked and annealed low-activation austenite 12Cr-20Mn steel has been studied. Corrosion tests were carried out in water solutions which simulate the coolant medium in the primary coolant circuit of WWER power reactor and in the circuit of multiple forced circulation of RBMK-1000 reactor as well as an aquatic environment in cooling pond for spent fuel. The worst radiation effect was observed in the cooling pond environment where the value of corrosion rate is increased by tens or hundreds times

  12. 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.

  13. Organic coolants and their applications to fusion reactors

    International Nuclear Information System (INIS)

    Gierszewski, P.; Hollies, B.

    1986-08-01

    Organic coolants offer a unique set of characteristics for fusion applications. Their advantages include high-temperature (670 K or 400 degrees C) but low-pressure (2 MPa) operation, limited reactivity with lithium and lithium-lead, reduced corrosion and activation, good heat-transfer capabilities, no magnetohydrodynamic (MHD) effects, and an operating temperature range that extends to room temperature. The major disadvantages are decomposition and flammability. However, organic coolants have been extensively studied in Canada, including nineteen years with an operating 60-MW organic-cooled reactor. Proper attention to design and coolant chemistry controlled these potential problems to acceptable levels. This experience provides an extensive data base for design under fusion conditions. The organic fluid characteristics are described in sufficient detail to allow fusion system designers to evaluate organic coolants for specific applications. To illustrate and assess the potential applications, analyses are presented for organic-cooled blankets, first walls, high heat flux components and thermal power cycles. Designs are identified that take advantage of organic coolant features, yet have fluid decomposition related costs that are a small fraction of the overall cost of electricity. For example, organic-cooled first walls make lithium/ferritic steel blankets possible in high-field, high-surface-heat-flux tokamaks, and organic-cooled limiters (up to about 8 MW/m 2 surface heating) are a safer alternative to water cooling for liquid metal blanket concept. Organics can also be used in intermediate heat exchanger loops to provide efficient heat transfer with low reactivity and a large tritium barrier. 55 refs

  14. 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

  15. New cooling system of the FRG-1 two barrier system of the primary coolant cycle

    International Nuclear Information System (INIS)

    Knop, W.; Schreiner, P.

    2003-01-01

    The GKSS research center operates the swimming pool reactor FRG-1 with a thermal power of 5 MW as national neutron source for neutron scattering experiments and sample irradiation as well. Before changing the primary coolant cycle consisted of the reactor core and the closed piping including pumps, heat exchanger and delay tank. The closed cooling circuit was located underneath the reactor pool, in the so-called radioactive cellar. This piping system served secondary coolant system. Due to the location of the primary coolant cycle below the operation pool a postulated 2-F line break and simultaneous failure of the pool slide gate valve could lead to a falling dry of the total reactor core. the new primary coolant system was built in the beginning 2002 in a partitioned cell all within the radioactive cellar, so that the reactor core remains with water with the assumed incident. Due to the new two barrier-inclusion of the primary circuit only the melting of two fuel plates (from total 252 fuel plates) has to be taken into account. This measure and the core compactness in 2000 with a neutron flux gain of a factor of 2 makes the FRG-1 ready for the next 15 years of reactor operation. (author)

  16. Theoretical design and analysis of wideband active hard electromagnetic surfaces using non-Foster circuit loaded anisotropic metasurfaces

    Science.gov (United States)

    Li, Yunbo; Li, Aobo; Sievenpiper, Daniel

    2018-02-01

    The electromagnetic (EM) hard surface which can both support transverse electric and transverse magnetic surface wave modes has the important ability to reduce the EM blockage of metallic obstacles. We propose a method to design an electrically thin hard surface with wide bandwidth by loading with non-Foster elements. The wideband hard surface composed of an anisotropic impedance coating can be considered as a kind of active metasurface. We develop a method to determine the values of the loading non-Foster circuit which can minimize the dispersion of the unit cells. For this method, we derive accurate values for the loading non-Foster elements through theoretical analysis. We also determine the fundamental limitations on the bandwidth due to stability requirements. To verify our theoretical design, we simulate the transmission performance between the two ports on opposite sides of a metallic rhombus-shaped obstacle coated with the non-Foster based metasurface. The simulated results show that the blockage has been largely reduced over a broad bandwidth from 0.2 GHz to 1.5 GHz. Finally, we provide a discussion on how the resistive part of the non-Foster circuit can affect the performance of the wideband hard surface coating.

  17. Benchmark of AC and DC Active Power Decoupling Circuits for Second-Order Harmonic Mitigation in Kilowatt-Scale Single-Phase Inverters

    DEFF Research Database (Denmark)

    Qin, Zian; Tang, Yi; Loh, Poh Chiang

    2016-01-01

    efficiency and high power density is identified and comprehensively studied, and the commercially available film capacitors, the circuit topologies, and the control strategies adopted for active power decoupling are all taken into account. Then, an adaptive decoupling voltage control method is proposed...... to further improve the performance of dc decoupling in terms of efficiency and reliability. The feasibility and superiority of the identified solution for active power decoupling together with the proposed adaptive decoupling voltage control method are finally verified by both the simulation and experimental......This paper presents the benchmark study of ac and dc active power decoupling circuits for second order harmonic mitigation in kW scale single-phase inverters. First of all, a brief comparison of recently reported active power decoupling circuits is given, and the best solution that can achieve high...

  18. Electronic circuit encyclopedia 2

    International Nuclear Information System (INIS)

    Park, Sun Ho

    1992-10-01

    This book is composed of 15 chapters, which are amplification of weak signal and measurement circuit audio control and power amplification circuit, data transmission and wireless system, forwarding and isolation, signal converting circuit, counter and comparator, discriminator circuit, oscillation circuit and synthesizer, digital and circuit on computer image processing circuit, sensor drive circuit temperature sensor circuit, magnetic control and application circuit, motor driver circuit, measuring instrument and check tool and power control and stability circuit.

  19. Electronic circuit encyclopedia 2

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sun Ho

    1992-10-15

    This book is composed of 15 chapters, which are amplification of weak signal and measurement circuit audio control and power amplification circuit, data transmission and wireless system, forwarding and isolation, signal converting circuit, counter and comparator, discriminator circuit, oscillation circuit and synthesizer, digital and circuit on computer image processing circuit, sensor drive circuit temperature sensor circuit, magnetic control and application circuit, motor driver circuit, measuring instrument and check tool and power control and stability circuit.

  20. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    Energy Technology Data Exchange (ETDEWEB)

    Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

    2014-12-15

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

  1. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    International Nuclear Information System (INIS)

    Kryk, Holger; Hoffmann, Wolfgang; Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan

    2014-01-01

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products

  2. 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

  3. Coolant processing device for nuclear reactor

    International Nuclear Information System (INIS)

    Kizawa, Hideo; Funakoshi, Toshio; Izumoji, Yoshiaki

    1981-01-01

    Purpose: To reduce an entire facility cost by concentrating and isolating tritium accumulated in coolants, removing the tritium out of the system, and returning hydrogen gas generated at a reactor accident to a recombiner in a closed loop by the switching of a valve. Constitution: Coolant from a reactor cooling system processed by a chemical volume control system facility (CVCS) and coolant drain from various devices processed by a liquid waste disposing system facility (LWDS) are fed to a tritium isolating facility, in which they are isolated into concentrated tritium water and dilute tritium water. The concentrated tritium water is removed out of the system and stored. The dilute tritium water is reused as supply water for coolant. If an accident occurs to cause hydrogen to be generated, a closed loop is formed between the containment vessel and the recombiner, the hydrogen is recombined with oxygen in the air of the closed loop to be thus returned to water. (Kamimura, M.)

  4. Fatigue management considering LWR coolant environments

    International Nuclear Information System (INIS)

    Park, Heung Bae; Jin, Tae eun

    2000-01-01

    Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

  5. Selection of nuclear reactor coolant materials

    International Nuclear Information System (INIS)

    Shi Lisheng; Wang Bairong

    2012-01-01

    Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

  6. Standardized sampling system for reactor coolants

    International Nuclear Information System (INIS)

    Divine, J.R.; Munson, L.F.; Nelson, J.L.; McDowell, R.L.; Jankowski, M.W.

    1982-09-01

    A three-pronged approach was developed to reach the objectives of acceptable coolant sampling, assessment of occupational exposure from corrosion products, and model development for the transport and buildup of corrosion products. Emphasis is on sampler design

  7. 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

  8. 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

  9. Coolant clean up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tajima, Fumio; Iwami, Hiroshi.

    1981-01-01

    Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

  10. Evaluation of steam as a potential coolant for nonbreeding blanket designs

    International Nuclear Information System (INIS)

    Stevens, H.C.; Misra, B.; Youngdahl, C.K.

    1978-01-01

    A steam-cooled nonbreeding blanket design has been developed as an evolution of the Argonne Experimental Power Reactor (EPR) studies. This blanket concept complete with maintenance considerations is to function at temperatures up to 650 0 C utilizing nickel-based alloys such as Inconel 625. Thermo-mechanical analyses were carried out in conjunction with thermal hydraulic analysis to determine coolant chennel arrangements that permit delivery of superheated steam at 500 0 C directly to a modern fossil plant-type turbine. A dual-cycle system combining a pressurized water circuit coupled with a superheated steam circuit can produce turbine plant conversion efficiencies approaching 41.5%

  11. 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

  12. 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.)

  13. 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)

  14. Neural circuits in the brain that are activated when mitigating criminal sentences.

    Science.gov (United States)

    Yamada, Makiko; Camerer, Colin F; Fujie, Saori; Kato, Motoichiro; Matsuda, Tetsuya; Takano, Harumasa; Ito, Hiroshi; Suhara, Tetsuya; Takahashi, Hidehiko

    2012-03-27

    In sentencing guilty defendants, jurors and judges weigh 'mitigating circumstances', which create sympathy for a defendant. Here we use functional magnetic resonance imaging to measure neural activity in ordinary citizens who are potential jurors, as they decide on mitigation of punishment for murder. We found that sympathy activated regions associated with mentalising and moral conflict (dorsomedial prefrontal cortex, precuneus and temporo-parietal junction). Sentencing also activated precuneus and anterior cingulate cortex, suggesting that mitigation is based on negative affective responses to murder, sympathy for mitigating circumstances and cognitive control to choose numerical punishments. Individual differences on the inclination to mitigate, the sentence reduction per unit of judged sympathy, correlated with activity in the right middle insula, an area known to represent interoception of visceral states. These results could help the legal system understand how potential jurors actually decide, and contribute to growing knowledge about whether emotion and cognition are integrated sensibly in difficult judgments.

  15. The wiring of developing sensory circuits - from patterned spontaneous activity to mechanisms of synaptic plasticity

    Directory of Open Access Journals (Sweden)

    Alexandra Helen Leighton

    2016-09-01

    Full Text Available In order to accurately process incoming sensory stimuli, neurons must be organized into functional networks, with both genetic and environmental factors influencing the precise arrangement of connections between cells. Teasing apart the relative contributions of molecular guidance cues, spontaneous activity and visual experience during this maturation is on-going. During development of the sensory system, the first, rough organization of connections is created by molecular factors. These connections are then modulated by the intrinsically generated activity of neurons, even before the senses have become operational. Spontaneous waves of depolarisations sweep across the nervous system, placing them in a prime position to strengthen correct connections and weaken others, shaping synapses into a useful network. A large body of work now supports the idea that, rather than being a mere side-effect of the system, spontaneous activity actually contains information which readies the nervous system so that, as soon as the senses become active, sensory information can be utilized by the animal. An example is the neonatal mouse. As soon as the eyelids first open, neurons in the cortex respond to visual information without the animal having previously encountered structured sensory input (Cang et al., 2005a; Ko et al., 2013; Rochefort et al., 2011; Zhang et al., 2012. In vivo imaging techniques have advanced considerably, allowing observation of the natural activity in the brain of living animals down to the level of the individual synapse. New (optogenetic methods make it possible to subtly modulate the spatio-temporal properties of activity, aiding our understanding of how these characteristics relate to the function of spontaneous activity. Such experiments have had a huge impact on our knowledge by permitting direct testing of ideas about the plasticity mechanisms at play in the intact system, opening up a provocative range of fresh questions. Here, we

  16. Activation of adenosine low-affinity A3 receptors inhibits the enteric short interplexus neural circuit triggered by histamine.

    Science.gov (United States)

    Bozarov, Andrey; Wang, Yu-Zhong; Yu, Jun Ge; Wunderlich, Jacqueline; Hassanain, Hamdy H; Alhaj, Mazin; Cooke, Helen J; Grants, Iveta; Ren, Tianhua; Christofi, Fievos L

    2009-12-01

    We tested the novel hypothesis that endogenous adenosine (eADO) activates low-affinity A3 receptors in a model of neurogenic diarrhea in the guinea pig colon. Dimaprit activation of H2 receptors was used to trigger a cyclic coordinated response of contraction and Cl(-) secretion. Contraction-relaxation was monitored by sonomicrometry (via intracrystal distance) simultaneously with short-circuit current (I(sc), Cl(-) secretion). The short interplexus reflex coordinated response was attenuated or abolished by antagonists at H2 (cimetidine), 5-hydroxytryptamine 4 receptor (RS39604), neurokinin-1 receptor (GR82334), or nicotinic (mecamylamine) receptors. The A1 agonist 2-chloro-N(6)-cyclopentyladenosine (CCPA) abolished coordinated responses, and A1 antagonists could restore normal responses. A1-selective antagonists alone [8-cyclopentyltheophylline (CPT), 1,3-dipropyl-8-(2-amino-4-chlorophenyl)xanthine (PACPX), or 8-cyclopentyl-N(3)-[3-(4-(fluorosulfonyl)benzoyloxy)propyl]-xanthine (FSCPX)] caused a concentration-dependent augmentation of crypt cell secretion or contraction and acted at nanomolar concentrations. The A3 agonist N(6)-(3-iodobenzyl)-adenosine-5'-N-methyluronamide (IB-MECA) abolished coordinated responses and the A3 antagonist 3-ethyl-5-benzyl-2-methyl-4-phenylethynyl-6-phenyl-1,4-(+/-)-dihydropyridine-3,5-dicarboxylate (MRS1191) could restore and further augment responses. The IB-MECA effect was resistant to knockdown of adenosine A1 receptor with the irreversible antagonist FSCPX; the IC(50) for IB-MECA was 0.8 microM. MRS1191 alone could augment or unmask coordinated responses to dimaprit, and IB-MECA suppressed them. MRS1191 augmented distension-evoked reflex I(sc) responses. Adenosine deaminase mimicked actions of adenosine receptor antagonists. A3 receptor immunoreactivity was differentially expressed in enteric neurons of different parts of colon. After tetrodotoxin, IB-MECA caused circular muscle relaxation. The data support the novel concept that

  17. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

    The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

  18. In Vivo Visualization of Active Polysynaptic Circuits With Longitudinal Manganese-Enhanced MRI (MEMRI

    Directory of Open Access Journals (Sweden)

    Suellen Almeida-Corrêa

    2018-05-01

    Full Text Available Manganese-enhanced magnetic resonance imaging (MEMRI is a powerful tool for in vivo non-invasive whole-brain mapping of neuronal activity. Mn2+ enters active neurons via voltage-gated calcium channels and increases local contrast in T1-weighted images. Given the property of Mn2+ of axonal transport, this technique can also be used for tract tracing after local administration of the contrast agent. However, MEMRI is still not widely employed in basic research due to the lack of a complete description of the Mn2+ dynamics in the brain. Here, we sought to investigate how the activity state of neurons modulates interneuronal Mn2+ transport. To this end, we injected mice with low dose MnCl2 2. (i.p., 20 mg/kg; repeatedly for 8 days followed by two MEMRI scans at an interval of 1 week without further MnCl2 injections. We assessed changes in T1 contrast intensity before (scan 1 and after (scan 2 partial sensory deprivation (unilateral whisker trimming, while keeping the animals in a sensory enriched environment. After correcting for the general decay in Mn2+ content, whole brain analysis revealed a single cluster with higher signal in scan 1 compared to scan 2: the left barrel cortex corresponding to the right untrimmed whiskers. In the inverse contrast (scan 2 > scan 1, a number of brain structures, including many efferents of the left barrel cortex were observed. These results suggest that continuous neuronal activity elicited by ongoing sensory stimulation accelerates Mn2+ transport from the uptake site to its projection terminals, while the blockage of sensory-input and the resulting decrease in neuronal activity attenuates Mn2+ transport. The description of this critical property of Mn2+ dynamics in the brain allows a better understanding of MEMRI functional mechanisms, which will lead to more carefully designed experiments and clearer interpretation of the results.

  19. Alterations in Neuronal Activity in Basal Ganglia-Thalamocortical Circuits in the Parkinsonian State

    Directory of Open Access Journals (Sweden)

    Adriana eGalvan

    2015-02-01

    Full Text Available In patients with Parkinson’s disease and in animal models of this disorder, neurons in the basal ganglia and related regions in thalamus and cortex show changes that can be recorded by using electrophysiologic single-cell recording techniques, including altered firing rates and patterns, pathologic oscillatory activity and increased inter-neuronal synchronization. In addition, changes in synaptic potentials or in the joint spiking activities of populations of neurons can be monitored as alterations in local field potentials, electroencephalograms or electrocorticograms. Most of the mentioned electrophysiologic changes are probably related to the degeneration of diencephalic dopaminergic neurons, leading to dopamine loss in the striatum and other basal ganglia nuclei, although degeneration of non-dopaminergic cell groups may also have a role. The altered electrical activity of the basal ganglia and associated nuclei may contribute to some of the motor signs of the disease. We here review the current knowledge of the electrophysiologic changes at the single cell level, the level of local populations of neural elements, and the level of the entire basal ganglia-thalamocortical network in parkinsonism, and discuss the possible use of this information to optimize treatment approaches to Parkinson’s disease, such as deep brain stimulation therapy.

  20. Alterations in neuronal activity in basal ganglia-thalamocortical circuits in the parkinsonian state

    Science.gov (United States)

    Galvan, Adriana; Devergnas, Annaelle; Wichmann, Thomas

    2015-01-01

    In patients with Parkinson’s disease and in animal models of this disorder, neurons in the basal ganglia and related regions in thalamus and cortex show changes that can be recorded by using electrophysiologic single-cell recording techniques, including altered firing rates and patterns, pathologic oscillatory activity and increased inter-neuronal synchronization. In addition, changes in synaptic potentials or in the joint spiking activities of populations of neurons can be monitored as alterations in local field potentials (LFPs), electroencephalograms (EEGs) or electrocorticograms (ECoGs). Most of the mentioned electrophysiologic changes are probably related to the degeneration of diencephalic dopaminergic neurons, leading to dopamine loss in the striatum and other basal ganglia nuclei, although degeneration of non-dopaminergic cell groups may also have a role. The altered electrical activity of the basal ganglia and associated nuclei may contribute to some of the motor signs of the disease. We here review the current knowledge of the electrophysiologic changes at the single cell level, the level of local populations of neural elements, and the level of the entire basal ganglia-thalamocortical network in parkinsonism, and discuss the possible use of this information to optimize treatment approaches to Parkinson’s disease, such as deep brain stimulation (DBS) therapy. PMID:25698937

  1. Brain's reward circuits mediate itch relief. a functional MRI study of active scratching.

    Directory of Open Access Journals (Sweden)

    Alexandru D P Papoiu

    Full Text Available Previous brain imaging studies investigating the brain processing of scratching used an exogenous intervention mimicking scratching, performed not by the subjects themselves, but delivered by an investigator. In real life, scratching is a conscious, voluntary, controlled motor response to itching, which is directed to the perceived site of distress. In this study we aimed to visualize in real-time by brain imaging the core mechanisms of the itch-scratch cycle when scratching was performed by subjects themselves. Secondly, we aimed to assess the correlations between brain patterns of activation and psychophysical ratings of itch relief or pleasurability of scratching. We also compared the patterns of brain activity evoked by self-scratching vs. passive scratching. We used a robust tridimensional Arterial Spin Labeling fMRI technique that is less sensitive to motion artifacts: 3D gradient echo and spin echo (GRASE--Propeller. Active scratching was accompanied by a higher pleasurability and induced a more pronounced deactivation of the anterior cingulate cortex and insula, in comparison with passive scratching. A significant involvement of the reward system including the ventral tegmentum of the midbrain, coupled with a mechanism deactivating the periaqueductal gray matter (PAG, suggests that itch modulation operates in reverse to the mechanism known to suppress pain. Our findings not only confirm a role for the central networks processing reward in the pleasurable aspects of scratching, but also suggest they play a role in mediating itch relief.

  2. 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.

  3. Iron crud supply device to reactor coolant

    International Nuclear Information System (INIS)

    Baba, Takao.

    1993-01-01

    In a device for supplying iron cruds into reactor coolants in a BWR type power plant, a system in which feed water containing iron cruds is supplied to the reactor coolants after once passing through an ion exchange resin is disposed. As a result, iron cruds having characteristics similar with those of naturally occurring iron cruds in the plant are obtained and they react with ionic radioactivity, to form composite oxides. Then, iron cruds having high performance of being secured to the surface of a fuel cladding tube can be supplied to the reactor coolants, thereby enabling to greatly reduce the density of reactor water ionic radioactivity. In its turn, dose rate on the surface of pipelines can be reduced, thereby enabling to reduce operators' radiation exposure dose in the plant. Further, contamination of a condensate desalting device due to iron cruds can be prevented, and further, the density of the iron cruds supplied can easily be controlled. (N.H.)

  4. Limits to fuel/coolant mixing

    International Nuclear Information System (INIS)

    Corradini, M.L.; Moses, G.A.

    1985-01-01

    The vapor explosion process involves the mixing of fuel with coolant prior to the explosion. A number of analysts have identified 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 sim plified approach is suggested to estimate the upper limit on the amount of fuel/coolant mixing pos sible. The approach uses concepts first advanced by Fauske in a different way. The results indicat that water depth is an important parameter as well as the mixing length scale D /SUB mix/ , and for large values of D /SUB mix/ the fuel mass mixed is limited to <7% of the core mass

  5. Integrated circuit-based electrochemical sensor for spatially resolved detection of redox-active metabolites in biofilms.

    Science.gov (United States)

    Bellin, Daniel L; Sakhtah, Hassan; Rosenstein, Jacob K; Levine, Peter M; Thimot, Jordan; Emmett, Kevin; Dietrich, Lars E P; Shepard, Kenneth L

    2014-01-01

    Despite advances in monitoring spatiotemporal expression patterns of genes and proteins with fluorescent probes, direct detection of metabolites and small molecules remains challenging. A technique for spatially resolved detection of small molecules would benefit the study of redox-active metabolites that are produced by microbial biofilms and can affect their development. Here we present an integrated circuit-based electrochemical sensing platform featuring an array of working electrodes and parallel potentiostat channels. 'Images' over a 3.25 × 0.9 mm(2) area can be captured with a diffusion-limited spatial resolution of 750 μm. We demonstrate that square wave voltammetry can be used to detect, identify and quantify (for concentrations as low as 2.6 μM) four distinct redox-active metabolites called phenazines. We characterize phenazine production in both wild-type and mutant Pseudomonas aeruginosa PA14 colony biofilms, and find correlations with fluorescent reporter imaging of phenazine biosynthetic gene expression.

  6. Refurbishment of the IEAR1 primary coolant system piping supports

    International Nuclear Information System (INIS)

    Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

    2015-01-01

    A partial replacement of the IEA-R1 piping system was concluded in 2014. This paper presents the study and the structural analysis of the IEA-R1 primary circuit piping supports, considering all the changes involved in the replacement. The IEA-R1 is a nuclear reactor for research purposes designed by Babcox-Willcox that is operated by IPEN since 1957. The reactor life management and modernization program is being conducted for the last two decades and already resulted in a series of changes, especially on the reactor coolant system. This set of components, divided in primary and secondary circuit, is responsible for the circulation of water into the core to remove heat. In the ageing management program that includes regular inspection, some degradation was observed in the primary piping system. As result, the renewing of the piping system was conducted in 2014. Moreover the poor condition of some original piping supports gave rise to the refurbishment of all piping supports. The aim of the present work is to review the design of the primary system piping supports taking into account the current conditions after the changes and refurbishment. (author)

  7. 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)

  8. A Novel LTPS-TFT Pixel Circuit to Compensate the Electronic Degradation for Active-Matrix Organic Light-Emitting Diode Displays

    Directory of Open Access Journals (Sweden)

    Ching-Lin Fan

    2013-01-01

    Full Text Available A novel pixel driving circuit for active-matrix organic light-emitting diode (AMOLED displays with low-temperature polycrystalline-silicon thin-film transistors (LTPS-TFTs is studied. The proposed compensation pixel circuit is driven by voltage programming scheme, which is composed of five TFTs and one capacitor, and has been certified to provide uniform output current by the Automatic Integrated Circuit Modeling Simulation Program with Integrated Circuit Emphasis (AIM-SPICE simulator. The results of simulation show excellent performance, such as the low average error rate of OLED current variation (<0.5% and the low average nonuniformity of OLED current variation (<0.8% while the shift of threshold voltage of the driving poly-Si TFT and the OLED are both in the worst case ( V for TFT and  V for OLED. The proposed pixel circuit shows high immunity to the threshold voltage deviation of both the driving poly-Si TFT and the OLED.

  9. 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

  10. Comparative design study of FR plants with various coolants. 1. Studies on Na coolant FR, Pb-Bi coolant FR, gas coolant FR

    International Nuclear Information System (INIS)

    Konomura, Mamoru; Shimakawa, Yoshio; Hori, Toru; Kawasaki, Nobuchika; Enuma, Yasuhiro; Kida, Masanori; Kasai, Shigeo; Ichimiya, Masakazu

    2001-01-01

    In Phase I of the Feasibility Studies on the Commercialized Fast Reactor (FR) Cycle System, plant designs on FR were performed with various coolants. This report describes the plant designs on FR with sodium, lead-bismuth, CO 2 gas and He gas coolants. A construction cost of 0.2 million yen/kWe was set up as a design goal. The result is as follows: The sodium reactor has a capability to obtain the goal, and lead-bismuth and gas reactors may satisfy the goal with further improvements. (author)

  11. On-Line Coolant Chemistry Analysis

    International Nuclear Information System (INIS)

    LM Bachman

    2006-01-01

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

  12. 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

  13. Instrumentation and test gear circuits manual

    CERN Document Server

    Marston, R M

    2013-01-01

    Instrumentation and Test Gear Circuits Manual provides diagrams, graphs, tables, and discussions of several types of practical circuits. The practical circuits covered in this book include attenuators, bridges, scope trace doublers, timebases, and digital frequency meters. Chapter 1 discusses the basic instrumentation and test gear principles. Chapter 2 deals with the design of passive attenuators, and Chapter 3 with passive and active filter circuits. The subsequent chapters tackle 'bridge' circuits, analogue and digital metering techniques and circuitry, signal and waveform generation, and p

  14. Activity-Based Anorexia Alters the Expression of BDNF Transcripts in the Mesocorticolimbic Reward Circuit.

    Science.gov (United States)

    Ho, Emily V; Klenotich, Stephanie J; McMurray, Matthew S; Dulawa, Stephanie C

    2016-01-01

    Anorexia nervosa (AN) is a complex eating disorder with severe dysregulation of appetitive behavior. The activity-based anorexia (ABA) paradigm is an animal model in which rodents exposed to both running wheels and scheduled feeding develop aspects of AN including paradoxical hypophagia, dramatic weight loss, and hyperactivity, while animals exposed to only one condition maintain normal body weight. Brain-derived neurotrophic factor (BDNF), an activity-dependent modulator of neuronal plasticity, is reduced in the serum of AN patients, and is a known regulator of feeding and weight maintenance. We assessed the effects of scheduled feeding, running wheel access, or both on the expression of BDNF transcripts within the mesocorticolimbic pathway. We also assessed the expression of neuronal cell adhesion molecule 1 (NCAM1) to explore the specificity of effects on BDNF within the mesocorticolimbic pathway. Scheduled feeding increased the levels of both transcripts in the hippocampus (HPC), increased NCAM1 mRNA expression in the ventral tegmental area (VTA), and decreased BDNF mRNA levels in the medial prefrontal cortex (mPFC). In addition, wheel running increased BDNF mRNA expression in the VTA. No changes in either transcript were observed in the nucleus accumbens (NAc). Furthermore, no changes in either transcript were induced by the combined scheduled feeding and wheel access condition. These data indicate that scheduled feeding or wheel running alter BDNF and NCAM1 expression levels in specific regions of the mesocorticolimbic pathway. These findings contribute to our current knowledge of the molecular alterations induced by ABA and may help elucidate possible mechanisms of AN pathology.

  15. 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.

  16. Principal working group 3 on primary circuit integrity

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-12-31

    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.

  17. 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

  18. Controllable circuit

    DEFF Research Database (Denmark)

    2010-01-01

    A switch-mode power circuit comprises a controllable element and a control unit. The controllable element is configured to control a current in response to a control signal supplied to the controllable element. The control unit is connected to the controllable element and provides the control...

  19. Benchmark of AC and DC active power decoupling circuits for second-order harmonic mitigation in kW-scale single-phase inverters

    DEFF Research Database (Denmark)

    Qin, Zian; Tang, Yi; Loh, Poh Chiang

    2015-01-01

    studied, where the commercially available film capacitors, circuit topologies, and control strategies for active power decoupling are all taken into account. Then, an adaptive decoupling voltage control method is proposed to further improve the performance of dc decoupling in terms of efficiency...... and reliability. The feasibility and superiority of the identified solution for active power decoupling together with the proposed adaptive decoupling voltage control method are finally verified by both the experimental results obtained on a 2 kW single-phase inverter.......This paper presents the benchmark study of ac and dc active power decoupling circuits for second-order harmonic mitigation in kW-scale single-phase inverters. First of all, the best solutions of active power decoupling to achieve high efficiency and power density are identified and comprehensively...

  20. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  1. 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.

  2. 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)

  3. 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.)

  4. Fission Product Releases from a Core into a Coolant of a Prismatic 350-MWth HTR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Min; Jo, C. K. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    A prismatic 350-MW{sub th} high temperature reactor (HTR) is a means to generate electricity and process heat for hydrogen production. The HTR will be operated for an extended fuel burnup of more than 150 GWd/MTU. Korea Atomic Energy Research Institute (KAERI) is performing a point design for the HTR which is a pre-conceptual design for the analysis and assessment of engineering feasibility of the reactor. In a prismatic HTR, metallic and gaseous fission products (FPs) are produced in the fuel, moved through fuel materials, and released into a primary coolant. The FPs released into the coolant are deposited on the various helium-wetted surfaces in the primary circuit, or they are sorbed on particulate matters in the primary coolant. The deposited or sorbed FPs are released into the environment through the leakage or venting of the primary coolant. It is necessary to rigorously estimate such radioactivity releases into the environment for securing the health and safety of the occupational personnel and the public. This study treats the FP releases from a core into a coolant of a prismatic 350-MW{sub th} HTR. These results can be utilized as input data for the estimation of FP migration from a coolant into the environment. The analysis of fission product release within a prismatic 350-MW{sub th} HTR has been done. It was assumed that the HTR was operated at constant temperature and power for 1500 EFPDs. - The final burnup is 152 GWd/tHM at packing fraction of 25 %, and the final fast fluence is about 8 X 10{sup 21} n/cm{sup 2}, E{sub n} > 0.1 MeV. - The temperatures at the compact center and at the center of a kernel located at the compact center are 884 and 893 .deg. C, respectively, when the packing fraction is 25 % and the coolant temperature is 850 .deg. C. - Xenon is the most radioactive fission product in a coolant of a prismatic HTR when there are broken TRISOs and fuel component contaminated with heavy metals. For metallic fission products, the radioactivity

  5. Activation of mesocorticolimbic reward circuits for assessment of relief of ongoing pain: a potential biomarker of efficacy.

    Science.gov (United States)

    Xie, Jennifer Y; Qu, Chaoling; Patwardhan, Amol; Ossipov, Michael H; Navratilova, Edita; Becerra, Lino; Borsook, David; Porreca, Frank

    2014-08-01

    Preclinical assessment of pain has increasingly explored operant methods that may allow behavioral assessment of ongoing pain. In animals with incisional injury, peripheral nerve block produces conditioned place preference (CPP) and activates the mesolimbic dopaminergic reward pathway. We hypothesized that activation of this circuit could serve as a neurochemical output measure of relief of ongoing pain. Medications commonly used clinically, including gabapentin and nonsteroidal anti-inflammatory drugs (NSAIDs), were evaluated in models of post-surgical (1 day after incision) or neuropathic (14 days after spinal nerve ligation [SNL]) pain to determine whether the clinical efficacy profile of these drugs in these pain conditions was reflected by extracellular dopamine (DA) release in the nucleus accumbens (NAc) shell. Microdialysis was performed in awake rats. Basal DA levels were not significantly different between experimental groups, and no significant treatment effects were seen in sham-operated animals. Consistent with clinical observation, spinal clonidine produced CPP and produced a dose-related increase in net NAc DA release in SNL rats. Gabapentin, commonly used to treat neuropathic pain, produced increased NAc DA in rats with SNL but not in animals with incisional, injury. In contrast, ketorolac or naproxen produced increased NAc DA in animals with incisional but not neuropathic pain. Increased extracellular NAc DA release was consistent with CPP and was observed selectively with treatments commonly used clinically for post-surgical or neuropathic pain. Evaluation of NAc DA efflux in animal pain models may represent an objective neurochemical assay that may serve as a biomarker of efficacy for novel pain-relieving mechanisms. Copyright © 2014 International Association for the Study of Pain. Published by Elsevier B.V. All rights reserved.

  6. Bandwidth of reactor internals vibration resonance with coolant pressure oscillations

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Novikov, K.S.; Galivec, E.Yu.

    2009-01-01

    In a few decades a significant increase in a part of an electricity development on the NPP will require NPP to be operated in non full capacity modes and increase in operation time in transitive modes. Operating in such conditions as compared to the operation on a constant mode will lead to the increase in cyclic dynamical loading. In water cooled water moderated reactors these loading are realized as low-cyclic and high-cyclic loadings. High-cyclic loadings increases are caused by a raised vibration in non stationary modes of operation. It is known, that in some modes of a non full capacity reactor high-cyclic dynamic loadings can increase. It is obvious, that the development of management technologies is necessary for the life time management operation. In the context of this problem one of the main tasks are revealing and the prevention of the conditions of the occurrence of the operation leading to the resonant interaction of the coolant fluctuations and the equipment, reactor vessel (RV), fuel assemblies (FA) and reactor internals (RI) vibration. To prevent the appearance of the conditions for resonance interaction between the fluid flow and the equipments, it is necessary to provide the different frequencies for the self oscillations in the separated elements of the circulating system and also in the parts of the system formed by the comprising of these elements. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of coolant outside of which there is no resonant interaction. The presented work is devoted to finding the solution of this problem. There are results of theoretical an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. The accordance of results had been calculated with had been measured are satisfied for practical purposes. These

  7. 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...

  8. Monitoring of primary circuit and reactor of NPP A-1

    International Nuclear Information System (INIS)

    Prazska, M.; Majersky, M.; Rezbarik, J.; Sekely, S.; Vozarik, P.; Walthery, R.; Stuller, P.

    2005-01-01

    Nuclear Power Plant A-1 in Jaslovske Bohunice was commissioned in 1972. Heavy water moderated, carbon dioxide cooled channel type reactor was shut down after two accidents in 1977. During more serious second accident, the reduced coolant flow caused local overheating of the fuel and consequent damage/melting of the fuel channel. Both accidents had led to the damage of several fuel assemblies with extensive local damage of fuel claddings. As a consequence, the main cooling circuit was significantly contaminated by fission products and long-life alpha nuclides. The detailed monitoring of dose rates, smearable contamination and sampling of contamination was performed. Extended monitoring in reacto vessel, primary circuit pipes, turbo-compressors, steam generators, main valves, gas tanks and also heavy water system with collectors, coolers, distilling and purification station, pumps and valves was done. Appropriate devices and procedures for the monitoring and examination of the installations were prepared and applied. Obtained results will serve for the future planning of the decontamination and decommissioning works. The 3-D model of the reactor that had been developed as part of this Project proved invaluable for orientation, visualisation, planning and analysis of results. Dose rates were measured in the technological channels from the reactor hall floor to the bottom of the hot gas chamber in decrements of 1 m and 0.5 m. The highest absolute values of dose rates were found in channels located in the middle of the reactor (up to 1900 mGy/h in the active zone region). It is estimated that the total contaminated area of primary circuit equipment (pipework, steam generators and turbo-compressors) is some 48 000 m 2 . It follows that the total gamma contamination is of the order of 10 14 to 10 15 Bq and total alpha contamination 10 11 to 10 13 Bq. The total amount of deposits in the gas circuit is about 14.3 tons. (authors)

  9. Corrosion products in the primary circuits of PWRs

    International Nuclear Information System (INIS)

    Darras, R.

    1983-01-01

    The characteristics of PWR primary circuits are recalled, particularly the chemical specifications of the medium and the various materials used (austenitic steel, nickel alloys, cobalt-based alloys and zirconium alloys). The behaviour of these materials as regards general corrosion in nominal and transient conditions is then outlined briefly, special emphasis being laid on the effect of the determining parameters on the quantity of corrosion products formed. The release of the latter into the primary coolant is caused by two main processes: solubilization and erosion. Particular attention was given therefore to the laws governing the solubility of the oxides involved, especially as a function of temperature and pH. Erosion, or release in the form of solid particles, is relatively severe during transient events. As these corrosion products are then carried through all circuits, they cause deposits to form in favourable places on the walls as a result either of precipitation of soluble species or of sedimentation followed by consolidation of suspended particles. The presence of corrosion products in the primary circuits creates a particular impact since they become radioactive as they pass through the core and especially when they remain in it in the form of deposits; as a result, the products are capable of contaminating the entire system. Finally, although long-term reliability is obviously an essential condition for materials developed, attention must also be given to problems associated with a build-up of corrosion products in the cooling circuits and efforts made to minimize them. To that end, a number of precautions are recommended, and various remedies can be applied: selecting materials which are not readily activated, keeping structures clean, purifying fluids properly, restricting solubilization and precipitation, and perhaps, periodic decontamination. (author)

  10. Behaviour of fission products in PWR primary coolant and defected fuel rods evaluation

    International Nuclear Information System (INIS)

    Bourgeois, P.; Stora, J.P.

    1979-01-01

    The activity surveillance of the PWR primary coolant by γ spectometry gives some informations on fuel failures. The activity of different nuclides e.g. Xenons, Kryptons, Iodines, can be correlated with the number of the defected fuel rods. Therefore the precharacterization with eventually a prelocalization of the related fuel assemblies direct the sipping-test and allows a saving of time during refueling. A model is proposed to calculate the number of the defected rods from the activity measurements of the primary coolant. A semi-empirical model of the release of the fission products has been built from the activity measurements of the primary coolant in a 900 MWe PWR. This model allows to calculate the number of the defected rods and also a typical parameter of the mean damage. Fission product release is described by three stages: release from uranium dioxide, transport across the gas gap and behaviour in the primary coolant. The model of release from the oxide considers a diffusion process in the grains with trapping. The release then occurs either directly to free surfaces or with a delay due to a transit into closed porosity of the oxide. The amount released is the same for iodine and rare gas. With the gas gap transit is associated a transport time and a probability of trapping for the iodines. In the primary coolant the purification and the radioactive decay are considered. (orig.)

  11. Automatic sweep circuit

    International Nuclear Information System (INIS)

    Keefe, D.J.

    1980-01-01

    An automatically sweeping circuit for searching for an evoked response in an output signal in time with respect to a trigger input is described. Digital counters are used to activate a detector at precise intervals, and monitoring is repeated for statistical accuracy. If the response is not found then a different time window is examined until the signal is found

  12. Automatic sweep circuit

    Science.gov (United States)

    Keefe, Donald J.

    1980-01-01

    An automatically sweeping circuit for searching for an evoked response in an output signal in time with respect to a trigger input. Digital counters are used to activate a detector at precise intervals, and monitoring is repeated for statistical accuracy. If the response is not found then a different time window is examined until the signal is found.

  13. Graphite beds for coolant filtration at high temperature

    International Nuclear Information System (INIS)

    Heathcock, R.E.; Lacy, C.S.

    1978-01-01

    High temperature filtration will be provided for new Ontario Hydro CANDU heat transport systems. Filtration has been shown to effectively reduce the concentration of circulating corrosion products in our heat transport systems, hence, minimizing the processes of activity transport. This paper will present one option we have for this application; Deep Bed Granular Graphite Filters. The filter system is described by discussing pertinent aspects of its development programme. The compatibility of the filter and the heat transport coolant are demonstrated by results from loop tests, both out- and in-reactor, and by subsequent results from a large filter installation in the NPD NGS heat transport system. (author)

  14. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)

    2007-07-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.

  15. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N.

    2007-01-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes

  16. Gas turbine with two circuits and intermediate fuel conversion process

    International Nuclear Information System (INIS)

    Bachl, H.

    1978-01-01

    The combination of a fuel conversion process with a thermal process saves coolant and subsequent separation plant, in order to achieve the greatest possible use of the mechanical or electrical energy. The waste heat of a thermal circuit is taken to an endothermal chemical fuel conversion process arranged before a second circuit. The heat remaining after removal of the heat required for the chemical process is taken to a second thermal circuit. The reaction products of the chemical process which condense out during expansion in the second thermal process are selectively separated from the remaining gas mixture in the individual turbine stages. (HGOE) [de

  17. Evaluation of primary coolant pH operation methods for the domestic PWRs

    International Nuclear Information System (INIS)

    Paek, Seung Woo; Na, Jung Won; Kim, Yong Eak; Bae, Jae Heum

    1992-01-01

    Radioactive nuclides deposited on out-of-core surface after the radiation in the core by the transport of corrosion products (CRUD) through the primary coolant system in PWR which is the major plant type in Korea, are leading sources of radiation exposure to plant maintenance personnel. Thus, the optimal chemistry operation method is required for the reduction of radiation exposure by the corrosion products. This study analysed the actual water chemistry operation data of four operating domestic PWRs. And in order to evaluate the coolant chemistry operation data, a computer code which can calculate the activity buildup in the various chemistry conditions of PWR coolant was employed. Through the analysis of comparison between the activity buildup of actual water chemistry operation mode and that of assumed Elevated Li operation mode calculated by the computer code, it was found that the out-of-core radioactivity can be reduced by diminishing the deposition of corrosion products on the core in case that the Elevated Li operation mode is applied to the coolant chemistry operation of PWR. And the higher coolant pH operation was shown to have the advantage of the reduction of out-of-core activity buildup if the integrity of system structural materials and fuel cladding is guaranteed. (Author)

  18. Revised Mark 22 coolant temperature coefficients

    International Nuclear Information System (INIS)

    Graves, W.E.

    1987-01-01

    Coolant temperature coefficients for the Mark 22 charge published previously are non-conservative because of the neglect of a significant mechanism which has a positive contribution to reactivity. Even after correcting for this effect, dynamic tests made on a Mark VIB charge in the early 60's suggest the results are still non-conservative. This memorandum takes both of these sources of information into account in making a best estimate of the prompt (coolant plus metal) temperature coefficient. Although no safety issues arise from this work (the overall temperature coefficient still strongly contributes to reactor stability), it is obviously desirable to use best estimates for prompt coefficients in limits and other calculations

  19. Freeform Deposition Method for Coolant Channel Closeout

    Science.gov (United States)

    Gradl, Paul R. (Inventor); Reynolds, David Christopher (Inventor); Walker, Bryant H. (Inventor)

    2017-01-01

    A method is provided for fabricating a coolant channel closeout jacket on a structure having coolant channels formed in an outer surface thereof. A line of tangency relative to the outer surface is defined for each point on the outer surface. Linear rows of a metal feedstock are directed towards and deposited on the outer surface of the structure as a beam of weld energy is directed to the metal feedstock so-deposited. A first angle between the metal feedstock so-directed and the line of tangency is maintained in a range of 20-90.degree.. The beam is directed towards a portion of the linear rows such that less than 30% of the cross-sectional area of the beam impinges on a currently-deposited one of the linear rows. A second angle between the beam and the line of tangency is maintained in a range of 5-65 degrees.

  20. 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

  1. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  2. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  3. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  4. Enhancing resistance to burnout via coolant chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Tu, J. P.; Dinh, T. N.; Theofanous, T. G. [Univ. of California, Santa Barbara (United States)

    2003-07-01

    Boiling Crisis (BC) on horizontal, upwards-facing copper and steel surfaces under the influence of various coolant chemistries relevant to reactor containment waters is considered. In addition to Boric Acid (BA) and TriSodium Phosphate (TSP), pure De-Ionized Water (DIW) and Tap Water (TW) are included in experiments carried out in the BETA facility. The results are related to a companion paper on the large scale ULPU facility.

  5. Minimizing secondary coolant blowdown in HANARO

    International Nuclear Information System (INIS)

    Park, Y. C.; Woo, J. S.; Ryu, J. S.; Cho, Y. G.; Lim, N. Y.

    2000-01-01

    There is about 80m 3 /h loss of the secondary cooling water by evaporation, windage and blowdown during the operation of HANARO, 30MW research reactor. The evaporation and the windage is necessary loss to maintain the performance of cooling tower, but the blowdown is artificial lose to get rid of the foreign material and to maintain the quality of the secondary cooling water. Therefore, minimizing the blowdown loss was studied. It was confirmed, through the relation of the number of cycle and the loss rate of secondary coolant, that the number of cycle is saturated to 12 without blowdown because of the windage loss. When the secondary coolant is treated by high Ca-hardness treatment program (the number of cycle > 10) to maintain the number of cycle around 12 without blowdown, only the turbidity exceeds the limit. By adding filtering system it was confirmed, through the relation of turbidity and filtering rate of secondary cooling water, that the turbidity is reduced below the limit (5 deg.) by 2% of filtering rate without blowdown. And it was verified, through the performance test of back-flow filtering unit, that this unit gets rid of foreign material up to 95% of the back-flow and that the water can be reused as coolant. Therefore, the secondary cooling water can be treated by the high Ca-hardness program and filter system without blowdown

  6. 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.

  7. LOGIC CIRCUIT

    Science.gov (United States)

    Strong, G.H.; Faught, M.L.

    1963-12-24

    A device for safety rod counting in a nuclear reactor is described. A Wheatstone bridge circuit is adapted to prevent de-energizing the hopper coils of a ball backup system if safety rods, sufficient in total control effect, properly enter the reactor core to effect shut down. A plurality of resistances form one arm of the bridge, each resistance being associated with a particular safety rod and weighted in value according to the control effect of the particular safety rod. Switching means are used to switch each of the resistances in and out of the bridge circuit responsive to the presence of a particular safety rod in its effective position in the reactor core and responsive to the attainment of a predetermined velocity by a particular safety rod enroute to its effective position. The bridge is unbalanced in one direction during normal reactor operation prior to the generation of a scram signal and the switching means and resistances are adapted to unbalance the bridge in the opposite direction if the safety rods produce a predetermined amount of control effect in response to the scram signal. The bridge unbalance reversal is then utilized to prevent the actuation of the ball backup system, or, conversely, a failure of the safety rods to produce the predetermined effect produces no unbalance reversal and the ball backup system is actuated. (AEC)

  8. Components of the LWR primary circuit. Pt. 2

    International Nuclear Information System (INIS)

    1984-01-01

    This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 0 C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives. (orig.) [de

  9. Possibilities for the reduction of the activity build-up in the primary circuit of water-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Sachse, G.; Mittag, I.

    1985-01-01

    Basing upon the international literature in a review are refered: research and development efforts; release, transport and deposition of radioactive corrosion products under primary circuit conditions; experimental results in test and technical systems; and possibilities to control radiation fields in nuclear power plants by waterchemical measures, decontamination, and high temperature filtration. Relevant patents are summarized. (author)

  10. 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)

  11. Short- circuit tests of circuit breakers

    OpenAIRE

    Chorovský, P.

    2015-01-01

    This paper deals with short-circuit tests of low voltage electrical devices. In the first part of this paper, there are described basic types of short- circuit tests and their principles. Direct and indirect (synthetic) tests with more details are described in the second part. Each test and principles are explained separately. Oscilogram is obtained from short-circuit tests of circuit breakers at laboratory. The aim of this research work is to propose a test circuit for performing indirect test.

  12. 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.

  13. Circuit class therapy and 7-day-week therapy increase physiotherapy time, but not patient activity: early results from the CIRCIT trial.

    Science.gov (United States)

    English, Coralie; Bernhardt, Julie; Hillier, Susan

    2014-10-01

    The optimum model of physiotherapy service delivery for maximizing active task practice during rehabilitation after stroke is unknown. The purpose of the study was to examine the relative effectiveness of 2 alternative models of physiotherapy service delivery against a usual care control with regard to increasing patient activity. Substudy within a large 3-armed randomized controlled trial, which compared 3 different models of physiotherapy service delivery, was provided for 4 weeks during subacute, inpatient rehabilitation (n=283). The duration of all physiotherapy sessions was recorded. In addition, 32 participants were observed at 10-minute intervals for 1 weekday and 1 weekend day between 8:00 am and 4:30 pm. At each observation, we recorded physical activity, location, and people present. Participants receiving 7-day-week and circuit class therapy received an additional 3 hours and 22 hours of physiotherapy time, respectively, when compared with usual care. Participants were standing or walking for a median of 8.2% of observations. On weekdays, circuit class therapy participants spent more time in therapy-related activity (10.2% of observations) when compared with usual care participants (6.1% of observations). On weekends, 7-day therapy participants spent more time in therapy-related activity (4.2% of observations) when compared with both usual care and circuit class therapy participants (0% of observations for both groups). Activity levels outside of therapy sessions did not differ between groups. A greater dosage of physiotherapy time did not translate into meaningful increases in physical activity across the day. http://www.anzctr.org.au/. Unique identifier: ACTRN12610000096055. © 2014 American Heart Association, Inc.

  14. Formation and hydraulic effects of deposits in high temperature sodium coolant systems

    International Nuclear Information System (INIS)

    Yunker, W.

    1976-01-01

    Deposition of sodium impurities in the high temperature (600 0 C), high flow (Reynolds Number approximately equal to 8 x 10 4 ) regions of a sodium coolant circuit is being studied to determine its possible hydraulic effects. Increases in flow impedance (pressure drop/volume flow 2 ) of up to 30 percent have been detected in an annular flow sensor. The apparatus and preliminary results of these tests are presented. Continuing tests are to specifically identify the materials involved and the system conditions under which the formations occur

  15. 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.

  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. 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)

  18. Inspection of the Sizewll 'B' reactor coolant pump flywheels

    International Nuclear Information System (INIS)

    McNulty, A.L.; Cheshire, A.

    1992-01-01

    The Sizewell ''B'' safety case has categorised some primary circuit items as components for which failure is considered to be incredible. These Incredibility of Failure (IOF) components are particularly critical in their safety function, and specially stringent and all embracing provisions are made in their design, manufacture, inspection and operation. These provisions are such as to limit the probability of failure to levels which are so low that it does not have to be taken into account and no steps are necessary to control the consequences. The reactor coolant pump flywheel is considered to be an IOF component. Consequently there is a need for rigorous inspection during both manufacture and in service (ISI). The ISI requirement results in the need for an automated inspection. There is therefore a prerequisite to perform a Pre-Service Inspection (PSI) for baseline fingerprinting purposes. Furthermore there is a requirement that the inspection procedure, the inspection equipment and the operators are validated at the Inspection Validation Centre (IVC) of the AEA Technology laboratories at Risley. Development work is described. (author)

  19. DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

    Energy Technology Data Exchange (ETDEWEB)

    Genin, J.B. [DEN/DEC/S3C, CEA Cadarache, 13 - Saint-Paul-lez-Durance (France); Harrer, A. [EdF/SEPTEN, 69 - Villeurbanne (France); Musante, Y. [FRAMATOME-ANP, 69 - Lyon (France)

    2002-07-01

    DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

  20. DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

    International Nuclear Information System (INIS)

    Genin, J.B.; Harrer, A.; Musante, Y.

    2002-01-01

    DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

  1. Method for determining the optimum mode of operation of the chemical water regime in the water-steam-circuit of power plants

    International Nuclear Information System (INIS)

    Sommerfeldt, P.; Reisner, H.; Hartmann, G.; Kulicke, P.

    1988-01-01

    The method aims at increasing the lifetime of secondary coolant circuit components in nuclear power plants through the determination of the optimum mode of operation of the chemical water regime by help of radioisotopes

  2. Circuit parties.

    Science.gov (United States)

    Guzman, R

    2000-03-01

    Circuit parties are extended celebrations, lasting from a day to a week, primarily attended by gay and bisexual men in their thirties and forties. These large-scale dance parties move from city to city and draw thousands of participants. The risks for contracting HIV during these parties include recreational drug use and unsafe sex. Limited data exists on the level of risk at these parties, and participants are skeptical of outside help because of past criticism of these events. Health care and HIV advocates can promote risk-reduction strategies with the cooperation of party planners and can counsel individuals to personally reduce their own risk. To convey the message, HIV prevention workers should emphasize positive and community-centered aspects of the parties, such as taking care of friends and avoiding overdose.

  3. Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

    International Nuclear Information System (INIS)

    Asher, R.C.; Davies, D.; Jones, P.G.

    1976-12-01

    A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

  4. TRANP - a computer code for digital simulation of steady - state and transient behavior of a pressurizer water reactor primary circuit

    International Nuclear Information System (INIS)

    Chalhoub, E.S.

    1980-09-01

    A digital computer code TRANP was developed to simulate the steady-state and transient behavior of a pressurizer water reactor primary circuit. The development of this code was based on the combining of three codes already developed for the simulation of a PWR core, a pressurizer, a steam generator and a main coolant pump, representing the primary circuit components. (Author) [pt

  5. Reactor coolant purification system circulation pumps (CUW pumps)

    International Nuclear Information System (INIS)

    Tsutsui, Toshiaki

    1979-01-01

    Coolant purification equipments for BWRs have been improved, and the high pressure purifying system has become the main type. The quantity of purifying treatment also changed to 2% of the flow rate of reactor feed water. As for the circulation pumps, canned motor pumps are adopted recently, and the improvements of reliability and safety are attempted. The impurities carried in by reactor feed water and the corrosion products generated in reactors and auxiliary equipments are activated by neutron irradiation or affect heat transfer adversely, adhering to fuel claddings are core structures. Therefore, a part of reactor coolant is led to the purification equipments, and returned to reactors after the impurities are eliminated perfectly. At the time of starting and stopping reactors, excess reactor water and the contaminated water from reactors are transferred to main condenser hot wells or waste treatment systems. Thus the prescribed water quality is maintained. The operational modes of and the requirements for the CUW pumps, the construction and the features of the canned motor type CUW pumps are explained. Recently, a pump operated for 11 months without any maintenance has been disassembled and inspected, but the wear of bearings has not been observed, and the high reliability of the pump has been proved. (Kako, I.)

  6. Coolant purification and radiochemical transformations in the presence of actadecylamine

    International Nuclear Information System (INIS)

    Kopylov, A.S.; Svechina, N.N.

    1987-01-01

    Applicability of octadecylamine (ODA) as an additive for the surfactants improve NPP steam-water circuit operation, is studied. Conditions for ODA sorption from circuit waters with different sorbents, including cationite KU-2-8 and activated carbon, are found. Material sorption capacities depending on the initial ODA concentration, temperature and filtration rate are determined. Efficiency of ODA sorption with sorbents investigated is shown to be higher than 90%. The investigation results of radiolysis of ODA aqueous emulsions in the dose interval of gamma-radiation from 1 to 10 6 rad are presented. ODA radiolysis is shown to occur analogously to radiolytic ammonia oxidation in the aqueous solution

  7. Commutation circuit for an HVDC circuit breaker

    Science.gov (United States)

    Premerlani, William J.

    1981-01-01

    A commutation circuit for a high voltage DC circuit breaker incorporates a resistor capacitor combination and a charging circuit connected to the main breaker, such that a commutating capacitor is discharged in opposition to the load current to force the current in an arc after breaker opening to zero to facilitate arc interruption. In a particular embodiment, a normally open commutating circuit is connected across the contacts of a main DC circuit breaker to absorb the inductive system energy trapped by breaker opening and to limit recovery voltages to a level tolerable by the commutating circuit components.

  8. 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.

  9. EDF PWRs primary coolant purification strategies

    International Nuclear Information System (INIS)

    Gressier, Frederic; Mascarenhas, Darren; Taunier, Stephane; Le-Calvar, Marc; Bretelle, Jean-Luc; Ranchoux, Gilles

    2012-09-01

    In order to achieve a good physico-chemical quality of the primary coolant fluid, the primary water is continuously treated by the Chemical and Volume Control System (CVCS). This system is composed of a treatment chain containing filters and ion-exchange resins. In the EDF design, an upstream filter is placed before the resin so as to prevent it from being saturated with insoluble particles. Then, the fluid passes through several resin beds (up to 3 depending on the configuration) and again through a downstream filter that prevents resin fines dissemination into the reactor coolant. Much work has been conducted in the last 5 years on the homogenisation of products and usage on French EDF NPP primary coolant treatment, while taking into account the compromise between source term reduction, liquid and solid waste, and buying and disposal costs. Two national markets have been created, and two operational documents for chemists on site have been published: a filtration guideline and an ion-exchange resin guideline. Both documents give general information about the products used, how are they characterized and selected for national market (technical requirements, standards and tests), how they should be used and what are the change-out criteria. They are also periodically updated based on feedback from sites. The positive impact on resin and filter lifetime (extension of some, limitation of others), homogenisation of products and usage will be presented. Moreover, EDF is constantly in the process of improving the current purification methods, as well as researching the use of existing and novel technologies. In this field, recent experiments on short loading of resin during reactor shutdown has been tested on site with success. In addition, work is done on silica free filters, filter consumption and filter chemical release. An overview of these optimization methods will be given. (authors)

  10. Calculation and analysis of neutron and radiation characteristics of lead coolants with isotopic tailoring for future nuclear power facilities

    Energy Technology Data Exchange (ETDEWEB)

    Blokhin, A.I.; Ivanov, A.P.; Korobeinikov, V.V.; Lunev, V.P.; Manokhin, V.N.; Khorasanov, G.L. [SSC RF A. I. Leypunsky Institute for Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation)

    2000-03-01

    A new type of safe fast reactor with lead coolant was proposed in Russia. The use of coolants with low moderating properties is one of the ways to get a hard neutron spectrum and an increase in the burning of Np-237, Am-243 and other miner actinides(MA) fissionable preferentially in the fast reactor. The stable lead isotope, Pb-208, is proposed as the one of such coolants. The neutron inelastic scattering cross-section of Pb-208 is 3.0-3.5 times less than the one of other lead isotopes. Calculation of the MA transmutation rates in the standard BN-type fast reactor with different coolants is performed by Monte-Carlo method using Code MMKFK. Six various models are simulated for the fast reactor blanket with different kinds of fuel and coolant. The fast reactor with natural-lead coolant practically does not differ from the reactor with sodium coolant relative to MA incineration. The use of Pb-208 as a coolant in the fast reactor results in increasing incineration of MA from 18 to 26% in comparison with a usual fast reactor. Calculation of induced radioactivity was performed using the FISPACT-3 inventory code, also. The results include total induced radioactivity and dose rate for initial material composition and selected long-lived radionuclides. The calculations show that the coolant consisting of lead isotope, Pb-206, or Pb-207, can be considered as the low-activation one because it does not practically contain long-lived toxic radionuclides. (M. Suetake)

  11. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  12. Microstructural characterization of primary coolant pipe steel

    International Nuclear Information System (INIS)

    Miller, M.K.; Bentley, J.

    1986-01-01

    Atom probe field-ion microscopy, analytical electron microscopy, and optical microscopy have been used to investigate the changes that occur in the microstructure of cast CF 8 primary coolant pipe stainless steel after long term thermal aging. The cast duplex microstructure consisted of austenite with 15% delta-ferrite. Investigation of the aged material revealed that the ferrite spinodally decomposed into a fine scaled network of α and α'. A fine G-phase precipitate was also observed in the ferrite. The observed degradation in mechanical properties is probably a consequence of the spinodal decomposition in the ferrite

  13. Calorimetric and reactor coolant system flow uncertainty

    International Nuclear Information System (INIS)

    Bates, L.; McLean, T.

    1991-01-01

    This paper describes a methodology for the quantification of errors associated with the determination of a feedwater flow, secondary power, and Reactor Coolant System (RCS) flow used at the Trojan Nuclear Plant to ensure compliance with regulatory requirements. The sources of error in Plant indications and process measurement are identified and tracked, using examples, through the mathematical processes necessary to calculate the uncertainty in the RCS flow measurement. An error of approximately 1.4 percent is calculated for secondary power. This error results, along with the consideration of other errors, in an uncertainty of approximately 3 percent in the RCS flow determination

  14. Application of liquid chromatography techniques to the measurement of soluble transition metals in PWR primary coolant

    International Nuclear Information System (INIS)

    Amey, M.D.H.; Brown, G.R.

    1987-01-01

    Two chromatographic techniques have been developed, and evaluated for the on-line analysis of soluble transition metals, particularly cobalt, in PWR primary coolant. Automatic operation and control, together with data processing and storage has been achieved by interfacing a Dionex ion chromatograph to a microprocessor control system. An absolute detection limit of 0.1 ng cobalt has been obtained which, with on-line sample preconcentration (100 ml), has enabled measurements to be made down to part-per-trillion levels (0.001 ppb). Application of the techniques to PWR coolant analysis was demonstrated by a programme of work on the Half Megawatt Loop at Winfrith. During this work some aspects of the behaviour of soluble metal species have been studied in both de-oxygenated and hydrogenated conditions. The effects of changes in coolant chemistry, operating temperature, and sample line flowrates on circulating impurity levels are reported, together with the dramatic effects observed when part of the circuit pipework was replaced with new stainless steel tubing. (author)

  15. Acute Effects of Three Different Circuit Weight Training Protocols on Blood Lactate, Heart Rate, and Rating of Perceived Exertion in Recreationally Active Women

    OpenAIRE

    Skidmore, Brook L.; Jones, Margaret T.; Blegen, Mark; Matthews, Tracey D.

    2012-01-01

    Interval and circuit weight training are popular training methods for maximizing time-efficiency, and are purported to deliver greater physiological benefits faster than traditional training methods. Adding interval training into a circuit weight-training workout may further enhance the benefits of circuit weight training by placing increased demands upon the cardiovascular system. Our purpose was to compare acute effects of three circuit weight training protocols 1) traditional circuit weigh...

  16. 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)

  17. 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.)

  18. 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.)

  19. In-Service Inspection system for coolant channels of Indian PHWRS - evolution and experience

    International Nuclear Information System (INIS)

    Puri, R.K.; Singh, M.

    2006-01-01

    In-Service Inspection (ISI) is the most important of all periodic monitoring and surveillance activities for assuring the structural integrity of coolant channels in the life extension and management of pressurized heavy water reactors (PHWR-CANDU). Indian PHWRs (220 MWe) are characterized by consists by 306 coolant channels in each unit. These channels have to be inspected for various parameters over the operating life of the reactor. ISI of coolant channels necessitated the indigenous development of an inspection system called BARCIS (BARC Channel Inspection System) at Bhabha Atomic Research Center. BARCIS consists of mainly three parts; drive and control unit, special sealing plug and an inspection head carrying various NDT sensors. Five such systems have been built and deployed at various power plants. The paper deals with the development of the BARCIS system for meeting the ISI requirements of coolant channels, development cycle of this system from its conception to evolution to the present state, challenges, data generated and experience gained (ISI of nearly 900 coolant channels has been completed). Prior to BARCIS, pressure tube gauging equipment for pre-service inspection of coolant tubes was developed in 1980. Moreover a tool for ISI of coolant channels in dry condition was developed in 1990. The paper also describes evolution of various contingency procedures and devices developed over the last one decade. Future plans taking into account technological advancement, changes in the scope of inspection due to design and operating experiences and plant layout will also be covered. The paper describes the efforts put in to develop drive and control mechanism to suit the different vault layouts. The drive mechanism is responsible for linear and rotary movement of the inspection head to carry out 100% volumetric inspection. Special emphasis has been laid on the safety devices required during the inspection activity. Special measures for heavy water retention in

  20. 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

  1. Effect of parameter variation of reactor coolant pump on loss of coolant accident consequence

    International Nuclear Information System (INIS)

    Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

    2015-01-01

    In this paper, the analyses were carried out on Ling'ao nuclear power station phase II to study the consequence of the loss of coolant accident when the homologous characteristic curves and free volumes of the reactor coolant pump changed. Two different pumps used in the analysis were 100D (employed on Ling'ao nuclear power station phase II) and ANDRITZ. The thermal characteristics in the large break LOCA accident were analyzed using CATHRE GB and CONPATE4, and the reactor coolant system hydraulics load during blow-clown phase of LOCA accident was analyzed using ATHIS and FORCET. The calculated results show that the homologous characteristic curves have great effect on the thermal characteristics of reactor core during the reflood phase of the large break LOCA accident. The maximum cladding surface temperatures are quite different when the pump's homologous characteristic curves change. On the other hand, the pump's free volume changing results in the variation of the LOCA rarefaction wave propagation, and therefore, the reactor coolant system hydraulic load in LOCA accident would be different. (authors)

  2. 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.

  3. Trigger circuit

    International Nuclear Information System (INIS)

    Verity, P.R.; Chaplain, M.D.; Turner, G.D.J.

    1984-01-01

    A monostable trigger circuit comprises transistors TR2 and TR3 arranged with their collectors and bases interconnected. The collector of the transistor TR2 is connected to the base of transistor TR3 via a capacitor C2 the main current path of a grounded base transistor TR1 and resistive means R2,R3. The collector of transistor TR3 is connected to the base of transistor TR2 via resistive means R6, R7. In the stable state all the transistors are OFF, the capacitor C2 is charged, and the output is LOW. A positive pulse input to the base of TR2 switches it ON, which in turn lowers the voltage at points A and B and so switches TR1 ON so that C2 can discharge via R2, R3, which in turn switches TR3 ON making the output high. Thus all three transistors are latched ON. When C2 has discharged sufficiently TR1 switches OFF, followed by TR3 (making the output low again) and TR2. The components C1, C3 and R4 serve to reduce noise, and the diode D1 is optional. (author)

  4. 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.

  5. Characterization of primary coolant purification system samples for assay of spent ion exchanger radionuclide inventor

    International Nuclear Information System (INIS)

    Sajin Prasad, S.; Pant, Amar; Sharma, Ranjit; Pal, Sanjit

    2018-01-01

    The primary coolant system water of a research reactor contains various fission and activation products and the water is circulated continuously through ion exchange resin cartridges, to reduce the radioactive ionic impurity present in it. The coolant purification system comprises of an ion exchange cooler, two micro filters, and a battery of six ion exchanger beds, associated valves, piping and instrumentation (Heavy water System Operating manual, 2014). The spent cartridge is finally disposed off as active solid waste which contains predominantly long lived fission and activation products. The heavy water coolant is also used to cool the structural assemblies after passing through primary heat exchanger and a metallic strainer, which accumulates the fission and activation products. When there is a reduction of coolant flow through these strainers, they are removed for cleaning and decontamination. This paper describes the characterization of ion exchange resin samples and liquid effluent generated during ultra sonic decontamination of strainer. The results obtained can be used as a methodology for the assay of the spent ion exchanger cartridges radionuclide inventory, during its disposal

  6. Optimization of a primary circuit of the nuclear power plant from the vibration point of view

    International Nuclear Information System (INIS)

    Dupal, J.; Zeman, V.

    2003-01-01

    The primary circuit of the nuclear power plant (NPP) as a dynamical vibrating system can be disturbed by various excitation including earthquake or pressure pulsation generated by main circulation pumps (MCP). Especially, unpleasant pulsation vibration growth can be caused by the small differences of revolutions between main circulation pumps of the individual coolant loops. This growth corresponds to the well known beats. The paper deals with an approach to the improving and optimization of dynamical properties of the whole primary circuit system including the reactor and coolant loops under pressure pulsation. (author)

  7. Temperature monitoring and leak detection in sodium circuits of FBR using Raman distributed fiber optic sensor

    International Nuclear Information System (INIS)

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

    2013-01-01

    This paper discusses the fiber optic temperature sensor based leak detection in the coolant circuits of fast breeder reactor. These sensors measure the temperature based on spontaneous Raman scattering principle and is not influenced by the electromagnetic interference. Various experiments were conducted to evaluate the performance of the fiber optic sensor based leak detection using Raman distributed Temperature Sensor (RDTS). This paper also deals with the details of fiber optic sensor type leak detector layout for the coolant circuit of FBR, performance requirement of leak detection system, description of the test facility, experimental procedure and test results of various experiments conducted. (author)

  8. Natural circulation in reactor coolant system

    International Nuclear Information System (INIS)

    Han, J.T.

    1987-01-01

    Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

  9. CFD analyses of coolant channel flowfields

    Science.gov (United States)

    Yagley, Jennifer A.; Feng, Jinzhang; Merkle, Charles L.

    1993-01-01

    The flowfield characteristics in rocket engine coolant channels are analyzed by means of a numerical model. The channels are characterized by large length to diameter ratios, high Reynolds numbers, and asymmetrical heating. At representative flow conditions, the channel length is approximately twice the hydraulic entrance length so that fully developed conditions would be reached for a constant property fluid. For the supercritical hydrogen that is used as the coolant, the strong property variations create significant secondary flows in the cross-plane which have a major influence on the flow and the resulting heat transfer. Comparison of constant and variable property solutions show substantial differences. In addition, the property variations prevent fully developed flow. The density variation accelerates the fluid in the channels increasing the pressure drop without an accompanying increase in heat flux. Analyses of the inlet configuration suggest that side entry from a manifold can affect the development of the velocity profile because of vortices generated as the flow enters the channel. Current work is focused on studying the effects of channel bifurcation on the flow field and the heat transfer characteristics.

  10. Requirements of coolants in nuclear reactors

    International Nuclear Information System (INIS)

    Abass, O. A. M.

    2014-11-01

    This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

  11. 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

  12. 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.)

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

    Science.gov (United States)

    Hunsbedt, Anstein N.

    1988-01-01

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

  14. 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.

  15. 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.)

  16. 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.

  17. Automated surveillance of reactor coolant pump performance

    International Nuclear Information System (INIS)

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

    1992-01-01

    An artificial intelligence based expert system has been developed for continuous surveillance and diagnosis of centrifugal-type reactor coolant pump (RCP) performance and operability. The expert system continuously monitors digitized signals from a variety of physical variables (speed, vibration level, motor power, discharge pressure) associated with RCP performance for annunciation of the incipience or onset of off-normal operation. The system employs an extremely sensitive pattern-recognition technique, the sequential probability ratio test (SPRT) for rapid identification of pump operability degradation. The sequential statistical analysis of the signal noise has been shown to provide the theoretically shortest sampling time to detect disturbances and thus has the potential of providing incipient fault detection information to operators sufficiently early to avoid forced plant shutdowns. The sensitivity and response time of the expert system are analyzed in this paper using monte carlo simulation techniques

  18. 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.

  19. Reactor coolant flow measurements at Point Lepreau

    International Nuclear Information System (INIS)

    Brenciaglia, G.; Gurevich, Y.; Liu, G.

    1996-01-01

    The CROSSFLOW ultrasonic flow measurement system manufactured by AMAG is fully proven as reliable and accurate when applied to large piping in defined geometries for such applications as feedwater flows measurement. Its application to direct reactor coolant flow (RCF) measurements - both individual channel flows and bulk flows such as pump suction flow - has been well established through recent work by AMAG at Point Lepreau, with application to other reactor types (eg. PWR) imminent. At Point Lepreau, Measurements have been demonstrated at full power; improvements to consistently meet ±1% accuracy are in progress. The development and recent customization of CROSSFLOW to RCF measurement at Point Lepreau are described in this paper; typical measurement results are included. (author)

  20. 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

  1. 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.

  2. Channel type reactors with supercritical water coolant. Russian experience

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Gabaraev, B.A.

    2003-01-01

    Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

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

    Science.gov (United States)

    Hutter, Ernest

    1986-01-01

    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.

  4. Fuel coolant interaction experiment by direct electrical heating method

    International Nuclear Information System (INIS)

    Takeda, Tsuneo; Hirano, Kenmei

    1979-01-01

    In the PCM (Power Cooling Mismatch) experiments, the FCI (Fuel Coolant Interaction) test is one of necessary tests in order to predict various phenomena that occur during PCM in the core. A direct electrical heating method is used for the FCI tests for fuel pellet temperature of over 1000 0 C. Therefore, preheating is required before initiating the direct electrical heating. The fuel pin used in the FCI tests is typical LWR fuel element, which is surrounded by coolant water. It is undersirable to heat up the coolant water during preheating of the fuel pin. Therefore, a zirconia (ZrO 2 ) pellet which is similar to a UO 2 pellet in physical and chemical properties is used. Electric property (electric conductivity) of ZrO 2 is particularly suitable for direct electrical heating as in the case of UO 2 . In this experiment, ZrO 2 pellet (melting point 2500 0 C) melting was achieved by use of both preheating and direct electrical heating. Temperature changes of coolant and fuel surface, as well as the pressure change of coolant water, were measured. The molten fuel interacted with the coolant and generated shock waves. A portion of this molten fuel fragmented into small particles during this interaction. The peak pressure of the observed shock wave was about 35 bars. The damaged fuel pin was photographed after disassembly. This report shows the measured coolant pressure changes and the coolant temperature changes, as well as photographs of damaged fuel pin and fuel fragments. (author)

  5. 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

  6. 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

  7. Current limiter circuit system

    Science.gov (United States)

    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.

  8. Stress Analysis of Fuel Rod under Axial Coolant Flow

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

    2010-05-15

    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

  9. 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.)

  10. 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

  11. Device for preventing coolant in a reactor from being lost

    International Nuclear Information System (INIS)

    Maruyama, Hiromi; Matsumoto, Tomoyuki.

    1975-01-01

    Object: To prevent all of coolant from being lost from the core at the time of failure in rupture of pipe in a recirculation system to cool the core with the coolant remained within the reactor. Structure: A valve, which will be closed when a water level of the coolant within the core is in a level less than a predetermined level, is provided on a recirculating water outlet nozzle in a pressure vessel to thereby prevent the coolant from being lost when the pipe is broken, thus cooling the core by means of reduced-pressure boiling of coolant remained within the core and boiling due to heat, and restraining core reactivity by means of void produced at that time. (Kamimura, M.)

  12. The effect of coolant quantity on local fuel–coolant interactions in a molten pool

    International Nuclear Information System (INIS)

    Cheng, Songbai; Matsuba, Ken-ichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Tohru; Tobita, Yoshiharu

    2015-01-01

    Highlights: • We investigate local fuel–coolant interactions in a molten pool. • As water volume increases, limited pressurization and mechanical energy observed. • Only a part of water is evaporated and responsible for the pressurization. - Abstract: Studies on local fuel–coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). Motivated by providing some evidence for understanding this interaction, in this study several experimental tests, with comparatively larger difference in coolant volumes, were conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are evaluated and compared. It is found that as water quantity increases, a limited pressure-buildup and the resultant mechanical energy release are observable. The performed analyses also suggest that only a part of water is probably vaporized during local FCIs and responsible for the pressurization and mechanical energy release, especially for those cases with much larger water volumes

  13. Study of the accumulation and distribution of the radioactivity in the cooling circuit of the BOR-60 reactor

    International Nuclear Information System (INIS)

    Kizin, V.D.; Konyashov, V.V.; Lisitsyn, E.S.; Polyakov, V.I.; Chechetkin, Yu.V.

    1976-04-01

    The results of measurements of the radioactivity of the coolant and the deposits in the primary circuit of the BOR-60 reactor during its five years of operation are discussed. The values calculated for the exposure dose rate from the piping system and the contribution of the γ-radiation from the corrosion and fission product nuclides are given. The efficiency of coolant draining from the pipes in reducing the dose rate is estimated. (orig.) [de

  14. 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.

  15. Simulation of a loss of primary coolant accident due to a large break in Angra 2 Nuclear Power Plant with RELAP5/MOD3.2.2G code

    International Nuclear Information System (INIS)

    Sabunddjian, Gaiane; Andrade, Delvonei Alves de

    2003-01-01

    This work presents the simulation, with RELAP5/MOD.3.2.2G code, of the postulate accident with loss of coolant in the primary circuit for large break (LBLOCA), which is described in Chapter 15 of the Final Safety Analysis Report of Angra 2 FSAR. The accident consists basically of the total break of the cold leg (Loop 20) of Angra 2 Plant. The rupture area considered is 4418 cm 2 , which represents 100% of the primary circuit pipe flow area. The Emergency Core Cooling System (ECCS) efficiency is also verified for this accident. In this simulation, failure and repair criteria are adopted for the ECCS components, in order to verify the system operation, in carrying out its function as expected by the project to preserve the integrity of the reactor core and to guarantee its cooling. LBLOCA accidents are characterized by a fast blowdown in the primary circuit to values that the low pressure injection system is activated and then, followed by the water injection by the accumulators. The thermal-hydraulic processes inherent to the accident phenomenon, such as hot leg vaporization and consequently core vaporization causing an inappropriate flow distribution in the reactor core, can lead to a reduction in the core liquid level, until the ECCS is capable to reflood it. It is important to point out that the results do not represent an independent calculation for the licensing process, but a calculation to give support to the qualification process of Angra 2 transient basic nodalization (author)

  16. Evaluation of alternate secondary (and tertiary) coolants for the molten-salt breeder reactor

    International Nuclear Information System (INIS)

    Kelmers, A.D.; Baes, C.F.; Bettis, E.S.; Brynestad, J.; Cantor, S.; Engel, J.R.; Grimes, W.R.; McCoy, H.E.; Meyer, A.S.

    1976-04-01

    The three most promising coolant selections for an MSBR have been identified and evaluated in detail from the many coolants considered for application either as a secondary coolant in 1000-MW(e) MSBR configurations using only one coolant, or as secondary and tertiary coolants in an MSBR dual coolant configuration employing two different coolants. These are, as single secondary coolants: (1) a ternary sodium--lithium--beryllium fluoride melt; (2) the sodium fluoroborate--sodium fluoride eutectic melt, the present reference design secondary coolant. In the case of the dual coolant configuration, the preferred system is molten lithium--beryllium fluoride (Li 2 BeF 4 ) as the secondary coolant and helium gas as the tertiary coolant

  17. Recommended reactor coolant water chemistry requirements for WWER-1000 units with 235U higher enriched fuel

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2011-01-01

    The last decade worldwide experience of PWRs and WWERs confirms the trends for the improvement of the nuclear power industry electricity production through the implementation of high burn-up or high fuel duty, which are usually accompanied with the usage of UO 2 fuel with higher content of 235 U - 4.0% - 4.5% (5.0%). It was concluded that the onset of sub-cooled nucleate boiling (SNB) on the fuel cladding surfaces and the initial excess reactivity of the core are the primary and basic factors accompanying the implementation of uranium fuel with higher 235 U content, aiming extended fuel cycles and higher burn-up of the fuel in Pressurized Water Reactors. As main consequences of the presence of these factors the modifications of chemical / electrochemical environments of nuclear fuel cladding- and reactor coolant system- surfaces are evaluated. These conclusions are the reason for: 1) The determination of the choices of the type of fuel cladding materials in respect with their enough corrosion resistance to the specific fuel cladding environment, created by the presence of SNB; 2) The development and implementation of primary circuit water chemistry guidelines ensuring the necessary low corrosion rates of primary circuit materials and limitation of cladding deposition and out-of-core radioactivity buildup; 3) Implementation of additional neutron absorbers which allow enough decrease of the initial concentration of H 3 BO 3 in coolant, so that its neutralization will be possible with the permitted alkalising agent concentrations. In this paper the specific features of WWER-1000 units in Bulgarian Nuclear Power Plant; use of 235 U higher enriched fuel in the WWER-1000 reactors in the Kozloduy NPP; coolant water chemistry and radiochemistry plant data during the power operation period of the Kozloduy NPP Unit 5, 15 th fuel cycle; evaluation of the approaches and results by the conversion of the WWER-1000 Units at the Kozloduy NPP to the uranium fuel with 4.3% 235 U as

  18. Dynamic pulse difference circuit

    International Nuclear Information System (INIS)

    Erickson, G.L.

    1978-01-01

    A digital electronic circuit of especial use for subtracting background activity pulses in gamma spectrometry is disclosed which comprises an up-down counter connected to count up with signal-channel pulses and to count down with background-channel pulses. A detector responsive to the count position of the up-down counter provides a signal when the up-down counter has completed one scaling sequence cycle of counts in the up direction. In an alternate embodiment, a detector responsive to the count position of the up-down counter provides a signal upon overflow of the counter

  19. Electronics circuits and systems

    CERN Document Server

    Bishop, Owen

    2007-01-01

    The material in Electronics - Circuits and Systems is a truly up-to-date textbook, with coverage carefully matched to the electronics units of the 2007 BTEC National Engineering and the latest AS and A Level specifications in Electronics from AQA, OCR and WJEC. The material has been organized with a logical learning progression, making it ideal for a wide range of pre-degree courses in electronics. The approach is student-centred and includes: numerous examples and activities; web research topics; Self Test features, highlighted key facts, formulae and definitions. Each chapter ends with a set

  20. Electronics circuits and systems

    CERN Document Server

    Bishop, Owen

    2011-01-01

    The material in Electronics - Circuits and Systems is a truly up-to-date textbook, with coverage carefully matched to the electronics units of the 2007 BTEC National Engineering and the latest AS and A Level specifications in Electronics from AQA, OCR and WJEC. The material has been organized with a logical learning progression, making it ideal for a wide range of pre-degree courses in electronics. The approach is student-centred and includes: numerous examples and activities; web research topics; Self Test features, highlighted key facts, formulae and definitions. Ea

  1. 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

  2. Development of fast-burn combustion with elevated coolant temperatures for natural gas engines. Final report, May 1985-May 1990

    Energy Technology Data Exchange (ETDEWEB)

    Bruch, K.L.; Dennis, J.W.

    1990-09-01

    The overall objective of the work was to improve the state of the art in the gas fired spark ignited engine for use in a cogeneration system. Four characteristics were enhanced for cogeneration, namely, Low Pressure Gas Induction, Improved Shaft Thermal Efficiency, Low NOx Emissions, and Increased Jacket Coolant Temperature. Using Taguchi methods and statistical design of experiment methodologies, an engine design evolved that exhibited: The ability to run satisfactorily on supply gas pressure as low as 1.5 psig (goal: 1 psig); A brake specific fuel consumption as low as 6950 Btu/hp-hr (36.6% thermal efficiency) at 2 gm/hp-hr NOx (goal: 7000 acceptable, 6800 excellent with NOx no more than 2 gm/hp-hr); A jacket water coolant system (with oil cooler on the same circuit) temperature of 225 F (goal); and The ability to burn gas with Methane Number as low as 67 (goal).

  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. 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

  5. The point of practical use for the transistor circuit

    International Nuclear Information System (INIS)

    1996-01-01

    This is comprised of eight chapters and goes as follows; what is transistor? the first step for use of transistor such as connection between power and signal source, static characteristic of transistor and equivalent circuit of transistor, design of easy small-signal amplifier circuit, design for amplification of electric power and countermeasure for prevention of trouble, transistor concerned interface, transistor circuit around micro computer, transistor in active use of FET and power circuit and transistor. It has an appendix on transistor and design of bias of FET circuits like small signal transistor circuit and FET circuit.

  6. 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

  7. Device for preventing coolant outflow in a reactor

    International Nuclear Information System (INIS)

    Nemoto, Kiyomitsu; Mochizuki, Keiichi.

    1975-01-01

    Object: To prevent outflow of coolant from a reactor vessel even in an occurrence of leaking trouble at a low position in a primary cooling system or the like in the reactor vessel. Structure: An inlet at the foremost end of a coolant inlet pipe inserted into a reactor vessel is arranged at a level lower than a core, and a check valve is positioned at a level higher than the core in a rising portion of the inlet. In normal condition, the check valve is pushed up by discharge pressure of a main circulating pump and remains closed, and hence, producing no flow loss of coolant, sodium. However, when a trouble such as rupture occurs at the lower position in the primary cooling system, the attractive force for allowing the coolant to back-flow outside the reactor vessel and the load force of the coolant within the reactor vessel cause the check valve to actuate, as a consequence of which a liquid level of the coolant downwardly moves to the position of the check valve to intake the cover gases into a gas intake, thereby cutting off a flow passage of the coolant to stop outflow thereof. (Kamimura, M.)

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Lister, D. [University of New Brunswick, Fredericton, NB (Canada). Dept. of Chemical Engineering; Lang, L.C. [Atomic Energy of Canada Ltd., Chalk River Lab., ON (Canada)

    2002-07-01

    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

  10. 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

  11. 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

  12. 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

  13. 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.)

  14. Characterization of reactor coolant by XRF

    Energy Technology Data Exchange (ETDEWEB)

    Legreid, G.; Beverskog, B. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  15. Characterization of reactor coolant by XRF

    International Nuclear Information System (INIS)

    Legreid, G.; Beverskog, B.

    2002-01-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  16. 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

  17. Speed control device for coolant recycling pump

    International Nuclear Information System (INIS)

    Kageyama, Takao.

    1992-01-01

    The present invention intends to increase a margin relative of the oscillations of neutron fluxes when the temperature of feedwater is lowered in a compulsory recycling type BWR reactor. That is, when the operation point represented by a reactor thermal power and a reactor core inlet flow rate is in a state approximate to an oscillation limit of the reactor power, the device of the present invention controls the recycling pump speed in the increasing direction depending on the lowering range of the feedwater temperature from a stationary state. With such a constitution, even if the reactor power is in the operation region near the oscillation limit in the BWR type reactor and a feedwater heating loss is caused, the speed of the coolant recycling pump is increased by 10% at the maximum depending on the extent of the reduction of the feedwater temperature, so that the oscillation of the reactor power can be prevented from lasting for a long period of time even if a reactivity external disturbance should occur in the reactor. (I.S.)

  18. Reactor coolant pump monitoring and diagnostic system

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  19. Reactor Coolant Temperature Measurement using Ultrasonic Technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, JaeCheon [KEPCO International Nuclear graduate School, Ulsan (Korea, Republic of); Seo, YongSun; Bechue, Nicholas [Krohne Messtechnik GmbH, Duisburg (Germany)

    2016-10-15

    In NPP, the primary piping temperature is detected by four redundant RTDs (Resistance Temperature Detectors) installed 90 degrees apart on the RCS (Reactor Coolant System) piping circumferentially. Such outputs however, if applied to I and C systems would not give balanced results. The discrepancy can be explained by either thermal stratification or improper arrangement of thermo-wells and RTDs. This phenomenon has become more pronounced in the hot-leg piping than in the cold-leg. Normally, the temperature difference among channels is in the range of 1°F in Korean nuclear power Plants. Consequently, a more accurate pipe average temperate measurement technique is required. Ultrasonic methods can be used to measure average temperatures with relatively higher accuracy than RTDs because the sound wave propagation in the RCS pipe is proportional to the average temperature around pipe area. The inaccuracy of RCS temperature measurement worsens the safety margin for both DNBR and LPD. The possibility of this discrepancy has been reported with thermal stratification effect. Proposed RCS temperature measurement system based on ultrasonic technology offers a countermeasure to cope with thermal stratification effect on hot-leg piping that has been an unresolved issue in NPPs. By introducing ultrasonic technology, the average internal piping temperature can be measured with high accuracy. The inaccuracy can be decreased less than ±1℉ by this method.

  20. 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)

  1. Mapping of olfactory memory circuits: region-specific c-fos activation after odor-reward associative learning or after its retrieval.

    Science.gov (United States)

    Tronel, Sophie; Sara, Susan J

    2002-01-01

    Although there is growing knowledge about intracellular mechanisms underlying neuronal plasticity and memory consolidation and reconsolidation after retrieval, information concerning the interaction among brain areas during formation and retrieval of memory is relatively sparse and fragmented. Addressing this question requires simultaneous monitoring of activity in multiple brain regions during learning, the post-acquisition consolidation period, and retrieval and subsequent reconsolidation. Immunoreaction to the immediate early gene c-fos is a powerful tool to mark neuronal activation of specific populations of neurons. Using this method, we are able to report, for the first time, post-training activation of a network of closely related brain regions, particularly in the frontal cortex and the basolateral amygdala (BLA), that is specific to the learning of an odor-reward association. On the other hand, retrieval of a well-established associative memory trace does not seem to differentially activate the same regions. The amygdala, in particular, is not engaged after retrieval, whereas the lateral habenula (LHab) shows strong activation that is restricted to animals having previously learned the association. Although intracellular mechanisms may be similar during consolidation and reconsolidation, this study indicates that different brain circuits are involved in the two processes, at least with respect to a rapidly learned olfactory task.

  2. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

    1995-10-01

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

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

    International Nuclear Information System (INIS)

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

    1984-11-01

    A review of 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 occurred 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 the compliance with the criteria defined in the operating technical specifications

  4. 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

  5. Evaluation of filtration and distillation methods for recycling automotive coolant

    International Nuclear Information System (INIS)

    Randall, P.M.; Gavaskar, A.R.

    1992-01-01

    Government regulations and high waste disposal cost of spent automotive coolant have driven the vehicle maintenance industry to explore on-site recycling. The USEPA in cooperation with the New Jersey Department of Environmental Protection (NJDEP) and the New Jersey Department of Transportation (NJDOT) evaluated two commercially available technologies that have potential for reducing the volume of spent automotive coolant. The objective of this study was to evaluate the quality of the recycled coolant, the pollution prevention potential, and the economic feasibility of the technologies

  6. 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

  7. Application of the regulations on pressurized components or light water reactor primary coolant circuits

    International Nuclear Information System (INIS)

    Barthelemy, F.; Menjon, G.

    1977-01-01

    This paper describes the philosophy and the provisions of the Order of 26 February 1974 concerning application of the regulations on pressurized components for light water reactor steam supply systems. The aim is to show how these regulations which differ from other regulations on pressurized components and is more detailed on many points, is applied in practice in France in the various stages of the design, construction and operation of PWRs. (NEA) [fr

  8. 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.)

  9. PWR steam generator tubes. Corrosion in primary coolant circuit. Evolution of knowledge

    International Nuclear Information System (INIS)

    Pinard Legry, G.

    1986-12-01

    Cracks can occur in nickel rich austenitic alloys in pure water at 350 0 C after few months. Influence of composition, microstructure stresses, corrosive effect of the medium, hydrogen embrittlement and temperature dependence on stress corrosion of alloy 600 are studied. A model is presented for the mechanism of crack formation [fr

  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. 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)

  12. 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

  13. Identification of a new gene regulatory circuit involving B cell receptor activated signaling using a combined analysis of experimental, clinical and global gene expression data

    Science.gov (United States)

    Schrader, Alexandra; Meyer, Katharina; Walther, Neele; Stolz, Ailine; Feist, Maren; Hand, Elisabeth; von Bonin, Frederike; Evers, Maurits; Kohler, Christian; Shirneshan, Katayoon; Vockerodt, Martina; Klapper, Wolfram; Szczepanowski, Monika; Murray, Paul G.; Bastians, Holger; Trümper, Lorenz; Spang, Rainer; Kube, Dieter

    2016-01-01

    To discover new regulatory pathways in B lymphoma cells, we performed a combined analysis of experimental, clinical and global gene expression data. We identified a specific cluster of genes that was coherently expressed in primary lymphoma samples and suppressed by activation of the B cell receptor (BCR) through αIgM treatment of lymphoma cells in vitro. This gene cluster, which we called BCR.1, includes numerous cell cycle regulators. A reduced expression of BCR.1 genes after BCR activation was observed in different cell lines and also in CD10+ germinal center B cells. We found that BCR activation led to a delayed entry to and progression of mitosis and defects in metaphase. Cytogenetic changes were detected upon long-term αIgM treatment. Furthermore, an inverse correlation of BCR.1 genes with c-Myc co-regulated genes in distinct groups of lymphoma patients was observed. Finally, we showed that the BCR.1 index discriminates activated B cell-like and germinal centre B cell-like diffuse large B cell lymphoma supporting the functional relevance of this new regulatory circuit and the power of guided clustering for biomarker discovery. PMID:27166259

  14. 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.

  15. Load testing circuit

    DEFF Research Database (Denmark)

    2009-01-01

    A load testing circuit a circuit tests the load impedance of a load connected to an amplifier. The load impedance includes a first terminal and a second terminal, the load testing circuit comprising a signal generator providing a test signal of a defined bandwidth to the first terminal of the load...

  16. Short-circuit logic

    NARCIS (Netherlands)

    Bergstra, J.A.; Ponse, A.

    2010-01-01

    Short-circuit evaluation denotes the semantics of propositional connectives in which the second argument is only evaluated if the first argument does not suffice to determine the value of the expression. In programming, short-circuit evaluation is widely used. A short-circuit logic is a variant of

  17. 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

  18. 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

  19. Improving Coolant Effectiveness through Drill Design Optimization in Gundrilling

    Science.gov (United States)

    Woon, K. S.; Tnay, G. L.; Rahman, M.

    2018-05-01

    Effective coolant application is essential to prevent thermo-mechanical failures of gun drills. This paper presents a novel study that enhances coolant effectiveness in evacuating chips from the cutting zone using a computational fluid dynamic (CFD) method. Drag coefficients and transport behaviour over a wide range of Reynold numbers were first established through a series of vertical drop tests. With these, a CFD model was then developed and calibrated with a set of horizontal drilling tests. Using this CFD model, critical drill geometries that lead to poor chip evacuation including the nose grind contour, coolant hole configuration and shoulder dub-off angle in commercial gun drills are identified. From this study, a new design that consists a 20° inner edge, 15° outer edge, 0° shoulder dub-off and kidney-shaped coolant channel is proposed and experimentally proven to be more superior than all other commercial designs.

  20. 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)

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  2. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  3. Active component modeling for analog integrated circuit design. Model parametrization and implementation in the SPICE-PAC circuit simulator; Modelisation de composants actifs pour la CAO de circuits integres analogiques. Parametrage et implantation de modeles dans le simulateur SPICE-PAC

    Energy Technology Data Exchange (ETDEWEB)

    Marchal, Xavier

    1992-06-19

    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{sub DS} = f(V{sub DS}, V{sub GS}, v{sub 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 thesis

  4. Hyperchaotic circuit with damped harmonic oscillators

    DEFF Research Database (Denmark)

    Lindberg, Erik; Murali, K.; Tamasevicius, A.

    2001-01-01

    A simple fourth-order hyperchaotic circuit with damped harmonic oscillators is described. ANP3 and PSpice simulations including an eigenvalue study of the linearized Jacobian are presented together with a hardware implementation. The circuit contains two inductors with series resistance, two ideal...... capacitors and one nonlinear active conductor. The Lyapunov exponents are presented to confirm the hyperchaotic nature of the oscillations of the circuit. The nonlinear conductor is realized with a diode. A negative impedance converter and a linear resistor. The performance of the circuit is investigated...... by means of numerical integration of the appropriate differential equations....

  5. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Keough, R.F.

    1981-01-01

    Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

  6. Review on research of small break loss of coolant accident

    International Nuclear Information System (INIS)

    Bo Jinhai; Wang Fei

    1998-01-01

    The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

  7. Effect of tungsten-187 in primary coolant on dose rate build-up in Vandellos 2

    International Nuclear Information System (INIS)

    Fernandez Lillo, E.; Llovet, R.; Boronat, M.

    1994-01-01

    The present work proposes a relationship between the Cobalt-60 piping deposited activity and the relatively high levels of Tungsten-187 in the coolant of Vandellos 2. The conclusions of this work can be applicable to other plants, since it proposes a tool to estimate and quantify the contribution of stellite to the generation of Cobalt-60 and the radiation dose build-up. (authors). 7 figs., 6 refs

  8. A study of the large break loss-of-coolant accident in the Angra-1 nuclear power plant

    International Nuclear Information System (INIS)

    Borges, E.M.

    1984-01-01

    The simulation of the Angra-I nuclear power plant under the condition of large break loss of coolant accident is presented, the thermal-hydraulic analysis of the primary circuit during each phase of the acident and thermal analysis of the hottest fuel rod curing reflooding are shown. Computer codes RELAP4/MOD5 (options EM and FLOOD) and TOODEE 2 are used to perform these computations. Fuel rod peak temperatures reached during the simulation are below the permissible levels. However, during the reflooding phase; the maximum oxidation of the cladding exceeds the limit of 0.17 times the original cladding thickness. (Author) [pt

  9. 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

  10. The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

    International Nuclear Information System (INIS)

    Silin, V.A.; Voznesensky, V.A.; Afrov, A.M.

    1993-01-01

    Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology. (orig.)

  11. Direct potentiometric control of chloride-ion content in water coolant of nuclear reactors

    International Nuclear Information System (INIS)

    Moskvin, L.N.; Vilkov, N.Ya.; Krasnoperov, V.M.; Epimakhova, L.V.

    1979-01-01

    The work of automatic chloride measuring device designed for continuous determination of chloride-ion concentration in water coolants of nuclear power plants is investigated. A series of experiments have been performed to investigate a device with sensitive element in the form of potentiometric cell with two flowing porous metal silver electrodes (PSE), placed in series. A calibration circuit of chloride measuring devices and PSE is described. A comparison is made between the results obtained by means of automatic chloride measuring device and results of manual control of samples. A conclusion is drawn that automatic chloride measuring devices meet the requirements of nuclear power plants for methods and instruments of control of chloride-ions microconcentration. The development and implantation of automatic chloride-ion analizers will make the analytical control on nuclear power plants easier and make it possible to obtain more reliable information

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

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

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

  13. Reactor coolant system hydrostatic test and risk analysis for the first AP1000 unit

    International Nuclear Information System (INIS)

    Cao Hongjun; Yan Xiuping

    2013-01-01

    The cold hydrostatic test scheme of the primary coolant circuit, of the first AP1000 unit was described. Based on the up-stream design documents, standard specifications and design technical requirements, the select principle of test boundary was identified. The design requirements for water quality, pressure, temperature and temporary hydro-test pump were proposed. A reasonable argument for heating and pressurization rate, and cooling and depressurization rate was proposed. The possible problems and risks during the hydrostatic test were analyzed. This test scheme can provide guidance for the revisions and implementations of the follow-up test procedures. It is a good reference for hydrostatic tests of AP1000 units in the future in China. (authors)

  14. Loss-of-coolant accident for large pipe breaks in light water reactor plants

    International Nuclear Information System (INIS)

    Keusenhoff, J.

    1980-01-01

    The importance of loss-of-coolant accidents (LOCA) and their control for nuclear reactor safety is explained. Showing the cooling circuits and emergency core cooling systems (ECCS) of both, PWR and BWR, the possible break spectrum and the general sequence of events is discussed. The governing physical phenomena for the different LOCA phases are pointed out in more detail. Special emphasis is taken on rules, regulations and failure criteria for licensing purposes. Analysis methods and codes for both, evaluation and best-estimate model are compared under deterministic and probabilistic approach, respectively. Some insight in present integral and separate effect tests demonstrates the interdependency of analysis and experiment. Results of LOCA analysis and experiments show the present state of the art. (orig.)

  15. 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

  16. 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.

  17. Circuit design for RF transceivers

    CERN Document Server

    Leenaerts, Domine; Vaucher, Cicero S

    2007-01-01

    Second edition of this successful 2001 RF Circuit Design book, has been updated, latest technology reviews have been added as well as several actual case studies. Due to the authors being active in industry as well as academia, this should prove to be an essential guide on RF Transceiver Design for students and engineers.

  18. Relative ultrasound energy measurement circuit

    OpenAIRE

    Gustafsson, E.Martin I.; Johansson, Jonny; Delsing, Jerker

    2005-01-01

    A relative ultrasound energy estimation circuit has been designed in a standard 0.35-μm CMOS process, to be a part of a thumb size internet connected wireless ultrasound measurement system. This circuit measures the relative energy between received ultrasound pulses, and presents an output signal that is linear to the received energy. Post-layout simulations indicate 7 bit linearity for 500 mV input signals, 5 μsec startup and stop times, 2.6 mW power consumption during active state. The acti...

  19. 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...

  20. Nuclear reactor with coolant circulation pumps

    International Nuclear Information System (INIS)

    Peck, D.A.; Stolecki, W.E.

    1975-01-01

    Thermally induced movement of a pump or a heat exchanger in the primary circuit of a PWR is made possible by a suspension device. This device must however be, so rigid that it does not yield in cases of emergency. For this purpose, in the case of the pump a lower ring is provided carrying the pump by means of four columns. The columns are flexibly supported on the ring and a fixed constuction. Turned about 90% from these columns, two additional horizontal bars are flexibly mounted on the ring and on the motor housing of the pump as well as on the fixed construction. At the upper end of the motor housing, two shock absorbers are hinged in the same way. The joints are shaped as ball- and socket hinges. (DG) [de

  1. Safety and environmental impact of the dual coolant blanket concept. SEAL subtask 6.2, final report

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Dammel, F.; Gabel, K.; Jordan, T.; Schmuck, I.

    1996-03-01

    The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four concepts under development, namely two of the solid breeder type and two of the liquid breeder type. At the Forschungszentrum Karlsruhe one blanket concept of each line has been pursued so far with the so-called dual coolant type representing the liquid breeder line. In the dual coolant concept the breeder material (Pb-17Li) is circulated to external heat exchangers to carry away the bulk of the generated heat and to extract the tritium. Additionally, the heavily loaded first wall is cooled by high pressure helium gas. The safety and environmental impact of the dual coolant blanket concept has been assessed as part of the blanket concept selection excercise, a European concerted action, aiming at selecting the two most promising concepts for futher development. The topics investigated are: (a) Blanket materials and toxic materials inventory, (b) energy sources for mobilisation, (c) fault tolerance, (d) tritium and activation products release, and (e) waste generation and management. No insurmountable safety problems have been identified for the dual coolant blanket. The results of the assessment are described in this report. The information collected is also intended to serve as input to the EU 'Safety and Environmental Assessment of Fusion longterm Programme' (SEAL). The unresolved issues pertaining to the dual coolant blanket which would need further investigations in future programmes are outlined herein. (orig.) [de

  2. Conference on instrumental activation analysis IAA 86

    International Nuclear Information System (INIS)

    Vobecky, M.; Obrusnik, I.

    1986-04-01

    Thirty five papers were presented at the conference held in Klucenice, Czechoslovakia from May 4 to 8, 1986. The abstracts of all papers are printed in the proceedings. The conference discussed the following problem areas: the application of activation analysis in determining elements in ores, tectites, fungi, the thyroid, the primary circuit coolant, semiconductor materials; the application of nuclear reaction analysis in determining elements in rubber and coal; the application of tracer techniques in metallurgy; the description of alpha and gamma spectrometric systems and their testing; the use of microcomputers for data processing, and the description of programs. (J.P.)

  3. Increased parietal circuit-breaker activity in delta frequency band and abnormal delta/theta band connectivity in salience network in hyperacusis subjects.

    Directory of Open Access Journals (Sweden)

    Jae Joon Han

    Full Text Available Recent studies have suggested that hyperacusis, an abnormal hypersensitivity to ordinary environmental sounds, may be characterized by certain resting-state cortical oscillatory patterns, even with no sound stimulus. However, previous studies are limited in that most studied subjects with other comorbidities that may have affected cortical activity. In this regard, to assess ongoing cortical oscillatory activity in idiopathic hyperacusis patients with no comorbidities, we compared differences in resting-state cortical oscillatory patterns between five idiopathic hyperacusis subjects and five normal controls. The hyperacusis group demonstrated significantly higher electrical activity in the right auditory-related cortex for the gamma frequency band and left superior parietal lobule (SPL for the delta frequency band versus the control group. The hyperacusis group also showed significantly decreased functional connectivity between the left auditory cortex (AC and left orbitofrontal cortex (OFC, between the left AC and left subgenual anterior cingulate cortex (sgACC for the gamma band, and between the right insula and bilateral dorsal anterior cingulate cortex (dACC and between the left AC and left sgACC for the theta band versus the control group. The higher electrical activity in the SPL may indicate a readiness of "circuit-breaker" activity to shift attention to forthcoming sound stimuli. Also, because of the disrupted salience network, consisting of the dACC and insula, abnormally increased salience to all sound stimuli may emerge, as a consequence of decreased top-down control of the AC by the dACC and dysfunctional emotional weight attached to auditory stimuli by the OFC. Taken together, abnormally enhanced attention and salience to forthcoming sound stimuli may render hyperacusis subjects hyperresponsive to non-noxious auditory stimuli.

  4. Method for reducing power consumption in a state retaining circuit, state reaining circuit and electronic device.

    NARCIS (Netherlands)

    2006-01-01

    A method for reducing the power consumption in a state retaining circuit during a standby mode is disclosed comprising, in an active state, providing a regular power supply (VDD) and a standby power supply (VDD STANDBY) to the state retaining circuit; for a transition from an active state to a

  5. Tumor heterogeneity is an active process maintained by a mutant EGFR-induced cytokine circuit in glioblastoma.

    Science.gov (United States)

    Inda, Maria-del-Mar; Bonavia, Rudy; Mukasa, Akitake; Narita, Yoshitaka; Sah, Dinah W Y; Vandenberg, Scott; Brennan, Cameron; Johns, Terrance G; Bachoo, Robert; Hadwiger, Philipp; Tan, Pamela; Depinho, Ronald A; Cavenee, Webster; Furnari, Frank

    2010-08-15

    Human solid tumors frequently have pronounced heterogeneity of both neoplastic and normal cells on the histological, genetic, and gene expression levels. While current efforts are focused on understanding heterotypic interactions between tumor cells and surrounding normal cells, much less is known about the interactions between and among heterogeneous tumor cells within a neoplasm. In glioblastoma multiforme (GBM), epidermal growth factor receptor gene (EGFR) amplification and mutation (EGFRvIII/DeltaEGFR) are signature pathogenetic events that are invariably expressed in a heterogeneous manner. Strikingly, despite its greater biological activity than wild-type EGFR (wtEGFR), individual GBM tumors expressing both amplified receptors typically express wtEGFR in far greater abundance than the DeltaEGFR lesion. We hypothesized that the minor DeltaEGFR-expressing subpopulation enhances tumorigenicity of the entire tumor cell population, and thereby maintains heterogeneity of expression of the two receptor forms in different cells. Using mixtures of glioma cells as well as immortalized murine astrocytes, we demonstrate that a paracrine mechanism driven by DeltaEGFR is the primary means for recruiting wtEGFR-expressing cells into accelerated proliferation in vivo. We determined that human glioma tissues, glioma cell lines, glioma stem cells, and immortalized mouse Ink4a/Arf(-/-) astrocytes that express DeltaEGFR each also express IL-6 and/or leukemia inhibitory factor (LIF) cytokines. These cytokines activate gp130, which in turn activates wtEGFR in neighboring cells, leading to enhanced rates of tumor growth. Ablating IL-6, LIF, or gp130 uncouples this cellular cross-talk, and potently attenuates tumor growth enhancement. These findings support the view that a minor tumor cell population can potently drive accelerated growth of the entire tumor mass, and thereby actively maintain tumor cell heterogeneity within a tumor mass. Such interactions between genetically

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

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

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

  8. Study on primary coolant system depressurization effect factor in pressurized water reactor

    International Nuclear Information System (INIS)

    Ji Duan; Cao Xuewu

    2006-01-01

    The progression of high-pressure core melting severe accident induced by very small break loss of coolant accident plus the loss of main feed water and auxiliary feed water failure is studied, and the entry condition and modes of primary cooling system depressurization during the severe accident are also estimated. The results show that the temperature below 650 degree C is preferable depressurization input temperature allowing recovery of core cooling, and the available and effective way to depressurize reactor cooling system and to arrest very small break loss of coolant accident sequences is activating pressurizer relief valves initially, then restoring the auxiliary feedwater and opening the steam generator relief valves. It can adequately reduce the primary pressure and keep the capacity loop of long-term core cooling. (authors)

  9. The Analysis of Applying Different Coolants for Cooling Systems in the Office Building

    Directory of Open Access Journals (Sweden)

    Rasa Kanapienytė

    2011-12-01

    Full Text Available The paper analyzes air conditioning systems of different coolants on the basis of an example of a typical office building. Depending on the type of a coolant fan coil unit, active chilled beams, variable refrigerant volumes and air cooling systems were designed. The article suggests hydraulic and aerodynamic calculations and evaluates initial investments, energy expenditures and operating costs of the compared systems. Considering economic calculations, the pay-back time of the systems was assessed and the sensitivity analysis of electricity prices was carried out. The results of the conducted investigation show the most appropriate analysed system for office buildings taking into account the efficient use of electricity and initial investments.Article in Lithuanian

  10. A device for monitoring the coolant in a nuclear reactor tank

    International Nuclear Information System (INIS)

    Smith, R.D.

    1984-01-01

    The invention deals with a gamma thermometer where the gamma absorber (stainless steel) is in heat conducting connection with an external casing located in the coolant in a reactor tank. A heat sink for the gamma absorber heated by gamma irradiation from reactor fuel is thereby established. The most sensitive joint in the thermocouple of the gamma thermometer is mounted vertically above the other joint. A differential voltage with a certain polarity will be generated between the two joints during uniform cooling of the external casing. If the coolant falls to a level under the most sensitive joint, the resulting polarity change can be utilized for the activation of alarm systems. The same gamma thermometer may simultaneously be used as a sensor for measurement of local power distribution

  11. Dysregulated neuronal activity patterns implicate corticostriatal circuit dysfunction in multiple rodent models of Huntington’s disease

    Directory of Open Access Journals (Sweden)

    Benjamin R. Miller

    2011-05-01

    Full Text Available Huntington’s disease (HD is an autosomal dominant neurodegenerative disorder that targets the corticostriatal system and results in progressive deterioration of cognitive, emotional, and motor skills. Although cortical and striatal neurons are widely studied in animal models of HD, there is little information on neuronal function during expression of the HD behavioral phenotype. To address this knowledge gap, we used chronically implanted micro-wire bundles to record extracellular spikes and local field potentials (LFPs in truncated (R6/1 and R6/2 and full-length (knock-in, KI mouse models as well as in tgHD rats behaving in an open-field arena. Spike activity was recorded in the striatum of all models and in prefrontal cortex (PFC of R6/2 and KI mice, and in primary motor cortex (M1 of R6/2 mice. We also recorded LFP activity in R6/2 striatum. All HD models exhibited altered neuronal activity relative to wild-type (WT controls. Although there was no consistent effect on firing rate across models and brain areas, burst firing was reduced in striatum, PFC, and M1 of R6/2 mice, and in striatum of KI mice. Consistent with a decline in bursting, the interspike-interval coefficient of variation was reduced in all regions of all models, except PFC of KI mice and striatum of tgHD rats. Among simultaneously recorded neuron pairs, correlated firing was reduced in all brain regions of all models, while coincident bursting, which measures the temporal overlap between bursting pairs, was reduced in striatum of all models as well as in M1 of R6/2's. Preliminary analysis of striatal LFPs revealed aberrant behavior-related oscillations in the delta to theta range and in gamma activity. Collectively, our results indicate that disrupted corticostriatal processing occurs across multiple HD models despite differences in the severity of the behavioral phenotype. Efforts aimed at normalizing corticostriatal activity may hold the key to developing new HD

  12. 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

  13. 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

    inductor. With such a configuration, this leg can control the current going into the two output capacitors connected in series for power decoupling, and the other leg can control the line current according to active and reactive power requirement. The proposed topology does not require additional passive...... 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...

  14. Improved Active-Neutral-Point-Clamped (I-ANPC) Multilevel Converter: Fundamental Circuit Topology, Innovative Modulatioin Technique, and Experimental Validation

    DEFF Research Database (Denmark)

    Dargahi, Vahid; Corzine, Keith A.; Enslin, Johan H.

    2018-01-01

    For medium-voltage (MV) high-power industrial applications including HVDC and variable-speed motor drives, multilevel converters are deemed predominant topology. One of the promising derived-topologies from neutral-point-clamped (NPC) configuration is active NPC (ANPC) inverter that offers improved......-capacitor (FC) and four switches such as the insulated-gate bipolar transistors (IGBTs). The I-ANPC converter has considerable advantages over the classic multilevel inverters that makes it a preferable topology for MV applications. The substantial reduction in the number of cells in comparison with classic...

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

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  16. Activity incorporation into zinc doped PWR oxides

    International Nuclear Information System (INIS)

    Maekelae, Kari

    1998-01-01

    Activity incorporation into the oxide layers of PWR primary circuit constructional materials has been studied in Halden since 1993. The first zinc injection tests showed that zinc addition resulted in thinner oxide layers on new metal surfaces and reduced further incorporation of activity into already existing oxides. These tests were continued to find out the effects of previous zinc additions on the pickup of activity onto the surface oxides which were subsequently exposed to zinc-free coolant. The results showed that previous zinc addition will continue to reduce the rate of Co-60 build-up on out-of-core surfaces in subsequent exposure to zinc-free coolants. However, the previous Zn free test was performed for a relatively short period of time and the water chemistry programme was continued to find out the long term effects for extended periods without zinc. The activity incorporation into the stainless steel oxides started to increase as soon as zinc dosing to the coolant was stopped. The Co-60 concentration was lowest on all of the coupons which were first oxidised in Zn containing primary coolant. After the zinc injection period the thickness of the oxides increased, but activity in the oxide films did not increase at the same rate. This could indicate that zinc in the oxide blocks the adsorption sites for Co-60 incorporation. The Co-60 incorporation rate into the oxides on Inconel 600 seemed to be linear whether the oxide was pre-oxidised with or without Zn. The results indicate that zinc can either replace or prevent cobalt transport in the oxides. The results show that for zinc injection to be effective it should be carried out continuously. Furthermore the actual mechanism by which Zn inhibits the activity incorporation into the oxides is still not clear. Therefore, additional work has to follow with specified materials to verify the conclusions drawn in this work. (author)

  17. Components of the LWR primary circuit. Pt. 2. Komponenten des Primaerkreises von Leichtwasserreaktoren. T. 2

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400/sup 0/C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives.

  18. Components of the LWR primary circuit. Pt. 2. Design, construction and calculation. Draft

    International Nuclear Information System (INIS)

    1995-01-01

    This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 deg C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives. (orig.) [de

  19. [Shunt and short circuit].

    Science.gov (United States)

    Rangel-Abundis, Alberto

    2006-01-01

    Shunt and short circuit are antonyms. In French, the term shunt has been adopted to denote the alternative pathway of blood flow. However, in French, as well as in Spanish, the word short circuit (court-circuit and cortocircuito) is synonymous with shunt, giving rise to a linguistic and scientific inconsistency. Scientific because shunt and short circuit made reference to a phenomenon that occurs in the field of the physics. Because shunt and short circuit are antonyms, it is necessary to clarify that shunt is an alternative pathway of flow from a net of high resistance to a net of low resistance, maintaining the stream. Short circuit is the interruption of the flow, because a high resistance impeaches the flood. This concept is applied to electrical and cardiovascular physiology, as well as to the metabolic pathways.

  20. Primary coolant recycling device for FBR type reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru; Tokiwai, Moriyasu

    1998-01-01

    A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)