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Sample records for coolant pump motor

  1. Trends and experiences in reactor coolant pump motors

    Anon.

    1980-01-01

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

  2. Impedance calculations for power cables to primary coolant pump motors

    Hegerhorst, K.B.

    1977-01-01

    The LOFT primary system motor generator sets are located in Room B-239 and are connected to the primary coolant pumps by means of a power cable. The calculated average impedance of this cable is 0.005323 ohms per unit resistance and 0.006025 ohms per unit reactance based on 369.6 kVA and 480 volts. The report was written to show the development of power cable parameters that are to be used in the SICLOPS (Simulation of LOFT Reactor Coolant Loop Pumping System) digital computer program as written in LTR 1142-16 and also used in the pump coastdowns for the FSAR Analysis

  3. Reactor Coolant Pump Motor Maintenance Experience in Krsko NPP

    Vukovic, J.; Besirevic, A.; Boljat, Z.

    2016-01-01

    After thirty years of service as well as maintenance in Krsko NPP both original Reactor Coolant Pump (RCP) motors are remanufactured by original vendor Westinghouse and a new one was purchased. Design function of the RCP motor is to drive Reactor Coolant Pump and for coast-down feature during Design Basis Accident. This paper will give a view on maintenance issues of RCP motor during the thirty years of service and maintenance in Krsko NPP to be kept functionally operational. During the processes of remanufacturing inspection and disassembly it was made possible to get a deeper perspective in the motor condition and the wear or fatigue of the motor parts. Parameters like bearing & winding temperature, absolute and relative vibration greatly affect motor operation if not kept inside design margins. Rotational speed causes heat generation at the bearings which is then associated with oil temperatures and as a consequence bearing temperatures. That is why the most critical parts of the motor are the components of upper and lower bearing assembly. The condition of motor stator and rotor assembly technical characteristics shall be explained with respect to influence of demanding environmental conditions that the motor is exposed. Assessment shall be made how does the wear of critical RCP motor parts can influence reliable performance of the motor if not maintained in proper way. Information on upgrades that were done on RCP motor shall be shared: Oil Spillage Protection System (OSPS), Stator upgrades, Dynamic Port, etc. (author).

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

    Deng Shaowen

    1998-01-01

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

  5. Reactor coolant pump testing using motor current signatures analysis

    Burstein, N.; Bellamy, J.

    1996-12-01

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

  6. Condition monitoring of primary coolant pump-motor units of Indian PHWR

    Rshikesan, P.B.; Sharma, S.S.; Mhetre, S.G.

    1994-01-01

    As the primary coolant pump motor units are located in shut down accessible area, their start up, satisfactory operation and shut down are monitored from control room. As unavailability of one pump in standardised 220 MWe station reduces the station power to about 110 MWe, satisfactory operation of the pump is also important from economic considerations. All the critical parameters of pump shaft, mechanical seal, bearing system, motor winding and shaft displacement (vibrations) are monitored/recorded to ensure satisfactory operation of critical, capital intensive pump-motor units. (author). 2 tabs., 1 fig

  7. Reactor coolant pump motors manufacturing capability and references

    Baudin, Patyrick [AREVA NP, Paris (France)

    2008-04-15

    Flywheel: - Main inertia of the RCP rotor: - 2 disks, shrunk to the upper side of the shaft, driven in rotation by 3 keys. - Material: rolling A533 grade B class 1 low alloy steel plates - Major inertia of the RCP rotor (Allows a slow shut down of the RCP). - Centered by the runner collar in normal operating conditions. - Designed to withstand over-speed of 1.25 x nominal rotating speed. - Easy periodic ultrasonic inspection without disassembly of the flywheel and/or removal of the motor. Anti-reverse rotation device: Prevents reverse rotation of shaft-line when RCP is stopped and others running. 5 forged pawls assembled on the flywheel outside diameter. Ratchet plate with shock absorbers and springs. Operation: Pawls are maintained lifted by centrifugal effect since N > 150 rpm. During RCP shut-down, as N < 150 rpm pawls drop on the ratchet plate prevents reverse-rotation due to reverse torque. Inertia effects are limited by shock-absorbers. Double thrust bearing 'Kings bury' type designed to support loads of about 60 tons 8 babbit ted steel shoes with temperature sensors, equalizing pads distribute equal axial load on each shoe, designed to withstand normal, transient and incidental loading conditions. Viscosity pump ensure continuous oil lubrication and oil circulation to cooler. Instrumentation: shoes temperature (167 .deg. F max). High pressure oil pump provides an oil film between runner and shoes before and during RCP start-up and shut-down. Secondary function: oil spray into the upper guide bearing. Characteristics: minimum oil injection pressure 610 psi. Upper guide bearing 8 babbit ted steel shoes. Preloaded shoes to improve the vibratory behavior. Lubricated by oil. Oil capacity: {+-} 240 gallons. Magnetic core made of high silicon steel sheets, insulated on both sides with 'ALKOPHOS' Stacks of sheets are periodically spaced by vent spacers Winding made of rectangular section copper bars, insulated with mica tape Vacuum impregnation

  8. Reactor coolant pump motors manufacturing capability and references

    Baudin, Patyrick

    2008-01-01

    Flywheel: - Main inertia of the RCP rotor: - 2 disks, shrunk to the upper side of the shaft, driven in rotation by 3 keys. - Material: rolling A533 grade B class 1 low alloy steel plates - Major inertia of the RCP rotor (Allows a slow shut down of the RCP). - Centered by the runner collar in normal operating conditions. - Designed to withstand over-speed of 1.25 x nominal rotating speed. - Easy periodic ultrasonic inspection without disassembly of the flywheel and/or removal of the motor. Anti-reverse rotation device: Prevents reverse rotation of shaft-line when RCP is stopped and others running. 5 forged pawls assembled on the flywheel outside diameter. Ratchet plate with shock absorbers and springs. Operation: Pawls are maintained lifted by centrifugal effect since N > 150 rpm. During RCP shut-down, as N < 150 rpm pawls drop on the ratchet plate prevents reverse-rotation due to reverse torque. Inertia effects are limited by shock-absorbers. Double thrust bearing 'Kings bury' type designed to support loads of about 60 tons 8 babbit ted steel shoes with temperature sensors, equalizing pads distribute equal axial load on each shoe, designed to withstand normal, transient and incidental loading conditions. Viscosity pump ensure continuous oil lubrication and oil circulation to cooler. Instrumentation: shoes temperature (167 .deg. F max). High pressure oil pump provides an oil film between runner and shoes before and during RCP start-up and shut-down. Secondary function: oil spray into the upper guide bearing. Characteristics: minimum oil injection pressure 610 psi. Upper guide bearing 8 babbit ted steel shoes. Preloaded shoes to improve the vibratory behavior. Lubricated by oil. Oil capacity: ± 240 gallons. Magnetic core made of high silicon steel sheets, insulated on both sides with 'ALKOPHOS' Stacks of sheets are periodically spaced by vent spacers Winding made of rectangular section copper bars, insulated with mica tape Vacuum impregnation with epoxy resin End

  9. Study on the VFD (Variable Frequency Drive) for RCP (Reactor Coolant Pump) Motors of APR1400

    Park, Jung Ha; Robert, M. Field; Kim, Tae Ryong [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    Most industrial facilities are continually searching for ways to reduce energy costs while increasing or maintaining current production. In terms of electric motors, Variable Frequency Drive (VFD) units represent a critical opportunity for energy savings. Currently, VFDs are used on about ten (10) percent of industrial process motors, and this percentage is increasing every year. Properly applied VFDs have been documented to save as much as fifty percent of the energy consumed by certain industrial processes. Nuclear Power - Power plants in general and Nuclear Power Plants (NPPs) in particular are slow to adopt new technology. The nuclear power industry requires a nearly absolute demonstration through operating experience in other industries in which the new approach will result in a net improvement in plant reliability without any surprises. Only recently has the nuclear industry begun to adapt VFD units for large motors. Specifically, there are several examples in the Boiling Water Reactor (BWR) fleet of replacing Motor-Generator (M-G) sets with VFD units for Reactor Recirculation (RR) pump motor service. At one station, VFD units were introduced upstream of the Circulating Water (CWP) pump motors to address environmental issues. They units are taking advantage of VFD technology whose benefits include increased reliability, reduction in electrical house load, improved flow control, and reduced maintenance. RCP Application - In the case of new generation, it has been reported that the Westinghouse AP1000 will make use of VFD units for the Reactor Coolant Pump (RCP) motors.

  10. Reactor coolant pump transportation incident

    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

  11. Development of motors and drives for main coolant pump and CEDM

    Kang, Do Hyun; Ha, Hoi Doo; Park, Jung Woo; Koo, Dae Hyun; Chang, Ki Chan; Kim, Jong Moo; Kim, Won Ho; Rim, Geun Hie; Baek, Ju Won; Park, Doh Young; Hwang, Don Ha; Jeon, Jeong Woo [Korea Electrotechnology Research Institute, Changwon (Korea)

    1999-03-01

    A canned type 170kW induction motor for the main coolant pump (MCP) of the integral reactor SMART was designed to minimize the eddy current loss in the can and the volume of motor. In order to verify the design and analysis methodology, a canned type 30kW induction motor and an inverter were developed and tested. The motor was designed to have two poles with squirrel cage solid rotor and open slot stator. The motor driver was designed as VVVF inverter to operate both at 900(r.p.m) and at 3600(r.p.m). The calculated design values showed a good agreement with the experimental results. The measured efficiencies of the canned motor and the inverter were 70(%) and 96(%), respectively. A variable reluctance type linear pulse motor (LPM) with double air-gaps for the Control Element Drive Mechanism (CEDM) to lift 100kg was designed, analyzed, manufactured and tested. A converter and a test facility were manufactured to verity the dynamic performance of the LPM. The mover of the LPM was welded with magnetic material(SUS430) and non-magnetic material(SUS304) to get flux path between inner stator and outer stator. The measured thrust force was about 20(%) less than the designed thrust force. As for the rotary stepping motors for CEDM-II, which have transverse flux pattern, three design options were proposed with thrust force density of 8kN/m{sup 2}, 14kN/m{sup 2} and 52kN/m{sup 2} respectively. (author). 31 refs., 219 figs., 60 tabs.

  12. Decontamination of main coolant pumps

    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

  13. Vertical reactor coolant pump instabilities

    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

  14. Regulatory instrument review: Aging management of LWR cables, containment and basemat, reactor coolant pumps, and motor-operated valves

    Werry, E.V.; Somasundaram, S.

    1995-09-01

    The results of Stage 2 of the Regulatory Instrument Review are presented in this volume. Selected regulatory instruments, such as the Code of Federal Regulations (CFR), US Nuclear Regulatory Commission (NRC), Regulatory Guides, and ASME Codes, were investigated to determine the extent to which these regulations apply aging management to selected safety-related components in nuclear power plants. The Regulatory Instrument Review was funded by the NRC under the Nuclear Plant Aging Research (NPAR) program. Stage 2 of the review focused on four safety-related structures and components; namely, cables, containment and basemat, reactor coolant pumps, and motor-operated valves. The review suggests that the primary-emphasis of the regulatory instruments was on the design, construction, start-up, and operation of a nuclear power plant, and that aging issues were primarily addressed after an aging-related problem was recognized. This Stage 2 review confirms the results of the prior review; (see Regulatory Instrument Review: Management of Aging of LWR Major Safety-Related Components NUREG/CR-5490. The observations indicate that the regulations generally address management of age-related degradation indirectly. Specific age-related degradation phenomena frequently are dealt with in bulletins and notices or through generic issues, letters, etc. The major recommendation of this report, therefore, is that the regulatory instruments should more directly and explicitly address the aging phenomenon and the management of the age-related degradation process

  15. Reactor having coolant recycling pump

    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. Reactor coolant purification system circulation pumps (CUW pumps)

    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.)

  17. Mathematical model of the reactor coolant pump

    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)

  18. Automated surveillance of reactor coolant pump performance

    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

  19. Condition monitoring of main coolant pumps, Dhruva

    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

  20. Reactor coolant pump seals: improving their performance

    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

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

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

    2004-01-01

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

  2. Main coolant pump testing at Ontario Hydro

    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

  3. Reactor coolant pumps for nuclear reactors

    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

  4. Power supplyer for reactor coolant recycling pump

    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.)

  5. The operating reliability of the reactor coolant pump

    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

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

    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

  7. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

    2010-01-15

    According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

  8. Motor-pump unit provided with a lifting appliance of the motor

    Veronesi, Luciano; Francis, W.R.

    1978-01-01

    This invention relates to lifting appliances and particularly concerns a 'pump and motor set' or motor-pump unit fitted with a lifting appliance enabling the motor to be separated from the pump. In nuclear power stations the reactor discharges heat that is carried by the coolant to a distant point away from the reactor to generate steam and electricity conventionally. In order to cause the reactor coolant to flow through the system, coolant motor-pump units are provided in the cooling system. These units are generally of the vertical type with an electric motor fitted vertically on the pump by means of a cylindrical or conical structure called motor support [fr

  9. Qualification test of a main coolant pump for SMART pilot

    Park, Sang Jin; Yoon, Eui Soo; Oh, Hyong Woo

    2006-01-01

    SMART Pilot is a multipurpose small capacity integral type reactor. Main Coolant Pump (MCP) of SMART Pilot is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel and steam generator in the primary system. The reactor is designed to operate under condition of 310 .deg. C and 14.7 MPa. Thus MCP has to be tested under same operating condition as reactor design condition to verify its performance and safety. In present work, a test apparatus to simulate real operating situations of the reactor has been designed and constructed to test MCP. And then functional tests, performance tests, and endurance tests have been carried out upon a prototype MCP. Canned motor characteristics, homologous head/torque curves, coast-down curves, NPSH curves and life-time performance variations were obtained from the qualification test as well as hydraulic performance characteristics of MCP

  10. Reactor coolant pump for a nuclear reactor

    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

  11. Reactor coolant pump seal leakage monitoring

    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

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

    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

  13. Nuclear reactor with coolant circulation pumps

    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

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

    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

  15. Lubrication analysis of the thrust bearing in the main coolant pump of SMART

    Lee, J. S.; Park, J. S.; Kim, J. H.; Hur, H.; Kim, J. I.

    2001-01-01

    Thrust bearing and journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel and especially the MCP bearings are lubricated with water without external lubricating oil supply. Because axial load capacity of the thrust bearing can hardly meet requirement to acquire hydrodynamic or fluid film lubrication state, self-lubrication characteristics of silicon graphite meterials would be needed. Lubricational analysis method for thrust bearing for the main coolant pump of SMART is proposed, and lubricational characteristics of the bearing generated by solving the Reynolds equation are examined in this paper

  16. Lubrication analysis of the journal bearing in the main coolant pump of SMART

    Lee, J. S.; Park, J. S.; Kim, J. H.; Kim, J. I.; Jang, M. H.

    2000-01-01

    Special type journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel. The MCP bearings are lubricated with water without external lubricating oil supply. Long bearing with vertical grooves is designed with relatively large bearing clearance to accommodate the long shaft. Lubricational analysis method for journal bearing with vertical grooves in the main coolant pump of SMART is proposed, and lubricational characteristics of the bearings are examined in this paper

  17. Reactor coolant pump monitoring and diagnostic system

    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

  18. Speed control device for coolant recycling pump

    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.)

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

    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)

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

    Grandia, M.R.

    1976-03-01

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

  1. Feasibility study on the type of KALIMER coolant circulation pump

    Nam, H. Y.; Kim, Y. K.; Lee, Y. B.; Hwang, J. S.; Choi, S. K.

    1997-07-01

    The characteristics of mechanical pump and electromagnetic (EM) pump for liquid sodium coolant in a liquid metal reactor are compared and analysed as a design concept of KALIMER coolant pumps. The type of coolant circulation pump affects the selection of reactor type, economics, and reliability of reactor. Though the mechanical pump has much application experience and give satisfaction to the reliability of developed reactor type, the possibility of development is limited and its large weight and volume have a negative effect on the design of the economical liquid metal reactor. The large scale electromagnetic pump has not been verified yet, but it is expected to be demonstrated in time. Because the size of EM pump is small relative to the mechanical pump, the compact reactor design is possible. Therefore the selection of EM pump can be one of the methods to improve the economics. Since the shape of EM pump can be varied according to the arrangement of electromagnet coils, a new or unique reactor type can be developed easily in the process of KALIMER development. In the view point of economic LMR development, it is desirable to adopt the electromagnetic pump. (author). 50 refs., 11 tabs., 24 figs

  2. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  3. Reactor coolant pump shaft seal behavior during blackout conditions

    Mings, W.J.

    1985-01-01

    The United States Nuclear Regulatory Commission has classified the problem of reactor coolant pump seal failures as an unresolved safety issue. This decision was made in large part due to experimental results obtained from a research program developed to study shaft seal performance during station blackout and reported in this paper. Testing and analysis indicated a potential for pump seal failure under postulated blackout conditions leading to a loss of primary coolant with a concomitant danger of core uncovery. The work to date has not answered all the concerns regarding shaft seal failure but it has helped scope the problem and focus future research needed to completely resolve this issue

  4. Wet motor geroter fuel pump

    Wiernicki, M.V.

    1987-05-05

    This patent describes a wet motor gerotor fuel pump for pumping fuel from a fuel source to an internal combustion which consists of: gerotor pump means comprising an inner pump gear, an outer pump gear, and second tang means located on one of the inner and outer pump gears. The second tang means further extends in a second radial direction radially offset from the first radial direction and forms a driving connection with the first tang means such that the fuel pump pumps fuel from the fuel source into the narrow conduit inlet chamber, through the gerotor pump means past the electric motor means into the outlet housing means substantially along the flow axis to the internal combustion engine.

  5. Structural integrity analysis of reactor coolant pump flywheel(I)

    Kim, Young Jin

    1986-01-01

    A reactor coolant pump flywheel is an important machine element to provide the necessary rotational inertia in the event of loss of power to the pumps. This paper attempts to assess the influence of keyways on flywheel stresses and fracture behaviour in detail. The finite element method was used to determine stresses near keyways, including residual stresses, and to establish stress intensity factors for keyway cracks for use in fracture mechanics assessments. (Author)

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

    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)

  7. On-line monitoring of main coolant pump seals

    Stevens, D.M.; Spencer, J.W.; Morris, D.J.; Glass, S.W.; Sommerfield, G.A.; Harrison, D.

    1984-06-01

    The Babcock and Wilcox Company has developed and implemented a Reactor Coolant Pump Monitoring and Diagnostic System (RCPM and DS). The system has been installed at Toledo Edison Company's Davis-Besse Nuclear Power Station Unit 1. The RCPM and PS continuously monitors a number of indicators of pump performance and notifies the plant operator of out-of-tolerance conditions or pump performance trending toward out-of-tolerance conditions. Pump seal parameters being monitored include pump internal pressures, temperatures, and flow rates. Rotordynamic performanvce and plant operating conditions are also measured with a variety of dynamic sensors. This paper describes the implementation of the system and the results of on-line monitoring of four RC pumps

  8. Vibration monitoring/diagnostic techniques, as applied to reactor coolant pumps

    Sculthorpe, B.R.; Johnson, K.M.

    1986-01-01

    With the increased awareness of reactor coolant pump (RCP) cracked shafts, brought about by the catastrophic shaft failure at Crystal River number3, Florida Power and Light Company, in conjunction with Bently Nevada Corporation, undertook a test program at St. Lucie Nuclear Unit number2, to confirm the integrity of all four RCP pump shafts. Reactor coolant pumps play a major roll in the operation of nuclear-powered generation facilities. The time required to disassemble and physically inspect a single RCP shaft would be lengthy, monetarily costly to the utility and its customers, and cause possible unnecessary man-rem exposure to plant personnel. When properly applied, vibration instrumentation can increase unit availability/reliability, as well as provide enhanced diagnostic capability. This paper reviews monitoring benefits and diagnostic techniques applicable to RCPs/motor drives

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

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

    1990-01-01

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

  10. Reactor coolant pump type RUV for Westinghouse Electric Company LLC reactor AP1000 TM

    Baumgarten, S.; Brecht, B.; Bruhns, U.; Fehring, P.

    2010-01-01

    The RUV is a reactor coolant pump, specially designed for the Westinghouse Electric Company LLC AP1000 TM reactor. It is a hermetically sealed, wet winding motor pump. The RUV is a very compact, vertical pump/motor unit, designed to fit into the compartment next to the reactor pressure vessel. Each of the two steam generators has two pump casings welded to the channel head by the suction nozzle. The pump/motor unit consists of a pump part, where a semi-axial impeller/diffuser combination is mounted in a one-piece pump casing. Computational Fluid Dynamics methods combined with various hydraulic tests in a 1:2 scale hydraulic test assure full compliance with the specific customer requirements. A short and rigid shaft, supported by a radial bearing, connects the impeller with the high inertia flywheel. This flywheel consists of a one-piece forged stainless steel cylinder, with an option for several smaller heavy metal cylinders inside. The flywheel is located inside the thermal barrier, which forms part of the pressure boundary. A specific arrangement of cooling water circuits guarantees a homogeneous temperature distribution in and around the flywheel, minimizes the friction losses of the flywheel and protects the motor from hot coolant. The driving torque is transmitted by the motor shaft, which itself is supported by two radial bearings. A three-phase, high-voltage squirrel-cage induction motor generates the driving torque. Due to the wet winding concept it is possible to achieve positive effects regarding motor lifetime. The cooling water is forced through the stator windings and the gap between rotor and stator by an auxiliary impeller. Furthermore, this wet winding motor concept has higher efficiency as compared to a canned motor since there are no eddy current losses. As part of the design process and in addition to the hydraulic scale model, a complete half scale model pump was built. It was used to verify the calculations performed like coast

  11. Design of Reactor Coolant Pump Seal Online Monitoring System

    Ah, Sang Ha; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of); Lee, Song Kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2008-05-15

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation.

  12. Design of Reactor Coolant Pump Seal Online Monitoring System

    Ah, Sang Ha; Chang, Soon Heung; Lee, Song Kyu

    2008-01-01

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation

  13. Development of manufacturing technology and fabrication of prototype for main coolant pump

    Chung, Koon Seok; Han, C.K.; Chei, J.M.; Chung, K.S.; Youn, M.H.; Shin, S.A.; Choi, D.J.; Kim, H.C. [HALLA Industrial Co., Ltd., Pusan (Korea)

    1999-03-01

    This study presents the development of the manufacturing technology for the Main Coolant Pump of the SMART. This report contains the followings; (1) Select axial type pump for the MCP (2) MCP is drived by squirrel-cage induction motor that consisted canned motor type. (3) MCP shaft has three horizontal and one vertical support bearings. (4) Design of several part of the MCP (5) Manufacturing of the performance test motor (6) Design and manufacturing of the speed sensor (7) Procedures for three-axial and five-axial M.C.T., Tig welding and Electron Beam Welding were developed. (8) Conceptional design of the MCP test facility for the performance test under operating conditions. (9) Results of standard weld test specimens according to the ASME section IX. (author). 21 refs., 35 figs., 10 tabs.

  14. In-operation diagnostic system for reactor coolant pump

    Sugiyama, Mitsunobu; Hasegawa, Ichiro; Kitahara, Hiromichi; Shimamura, Kazuo; Yasuda, Chiaki; Ikeda, Yasuhiro; Kida, Yasuo.

    1996-01-01

    A reactor coolant pump (RCP) is one of the most important rotating machines in the primary loop nuclear power plants. To improve the reliability and of nuclear power plants, a new diagnostic system that enables early detection of RCP faults has been developed. This system is based on continuous monitoring of vibration and other process data. Vibration is an important indicator of mechanical faults providing information on physical phenomena such as changes in dynamic characteristics and excitation forces changes that signal failure or incipient failure. This new system features comparative vibration analysis and simulation to anticipate equipment failure. (author)

  15. Reactor coolant pump shaft seal stability during station blackout

    Rhodes, D.B.; Hill, R.C.; Wensel, R.G.

    1987-05-01

    Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries

  16. Reactor coolant pump shaft seal stability during station blackout

    Rhodes, D B; Hill, R C; Wensel, R G

    1987-05-01

    Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries.

  17. Integrated main coolant pumps for pressurized-water reactors

    Wieser, R.

    1975-01-01

    The efficiency of an integrated main coolant pump for PWR's is increased. For this purpose, the pump is installed eccentric relative to the vertical axis of the U-type steam generator in the three-section HP chamber in such a way that its impeller wheel and the shell of the latter penetrate into the outlet chamber. The axis of the pump lies in the vertical plane of symmetry of the outlet chamber of the steam generator. The suction tube is arranged in the outlet chamber. To allow it to be installed, it is manufactured out of several parts. The diffusor tube, which is also made of several components, is attached to the horizontal separation plate between the outlet chamber and the pressure chamber so as to penetrate into it. To improve the outflow conditions at the diffusor tube, a plowshare-shaped baffle shield is installed between the diffusor tube and the HP chamber. Moreover, in order to improve the outflow conditions from the pump and from the pressure chamber, the outflow opening of the pressure chamber is put into the cylindrical shell of the HP chamber. In this way, the tensioning anchor is located between the pump and the outlet opening. (DG/RF) [de

  18. Submersible canned motor mixer pump

    Guardiani, R.F.; Pollick, R.D.

    1997-01-01

    A mixer pump is described used in a waste tank for mobilizing high-level radioactive liquid waste having a column assembly containing power cables, a motor housing with electric motor means which includes a stator can of a stator assembly and a rotor can of a rotor assembly, and an impeller assembly with an impeller connected to a shaft of the rotor assembly. The column assembly locates the motor housing with the electric motor means adjacent to the impeller which creates an hydraulic head, and which forces the liquid waste into the motor housing to cool the electric motor means and to lubricate radial and thrust bearing assemblies. Hard-on-hard bearing surfaces of the bearing assemblies and a ring assembly between the impeller and electric motor means act to grind down large particles in the liquid waste flow. These larger particles are received in slots in the static bearing members of the radial bearing assemblies. Only solid waste particles smaller than the clearances in the system can pass there through, thereby resisting damage to and the interruption of the operation of the mixer pump. 10 figs

  19. Reactor coolant pump shaft seal behavior during station blackout

    Kittmer, C.A.; Wensel, R.G.; Rhodes, D.B.; Metcalfe, R.; Cotnam, B.M.; Gentili, H.; Mings, W.J.

    1985-04-01

    A testing program designed to provide fundamental information pertaining to the behavior of reactor coolant pump (RCP) shaft seals during a postulated nuclear power plant station blackout has been completed. One seal assembly, utilizing both hydrodynamic and hydrostatic types of seals, was modeled and tested. Extrusion tests were conducted to determine if seal materials could withstand predicted temperatures and pressures. A taper-face seal model was tested for seal stability under conditions when leaking water flashes to steam across the seal face. Test information was then used as the basis for a station blackout analysis. Test results indicate a potential problem with an elastomer material used for O-rings by a pump vendor; that vendor is considering a change in material specification. Test results also indicate a need for further research on the generic issue of RCP seal integrity and its possible consideration for designation as an unresolved safety issue

  20. Multi-state reliability for coolant pump based on dependent competitive failure model

    Shang Yanlong; Cai Qi; Zhao Xinwen; Chen Ling

    2013-01-01

    By taking into account the effect of degradation due to internal vibration and external shocks. and based on service environment and degradation mechanism of nuclear power plant coolant pump, a multi-state reliability model of coolant pump was proposed for the system that involves competitive failure process between shocks and degradation. Using this model, degradation state probability and system reliability were obtained under the consideration of internal vibration and external shocks for the degraded coolant pump. It provided an effective method to reliability analysis for coolant pump in nuclear power plant based on operating environment. The results can provide a decision making basis for design changing and maintenance optimization. (authors)

  1. Submersible canned motor transfer pump

    Guardiani, R.F.; Pollick, R.D.; Nyilas, C.P.; Denmeade, T.J.

    1997-01-01

    A transfer pump is described which is used in a waste tank for transferring high-level radioactive liquid waste from a waste tank and having a column assembly, a canned electric motor means, and an impeller assembly with an upper impeller and a lower impeller connected to a shaft of a rotor assembly. The column assembly locates a motor housing with the electric motor means adjacent to the impeller assembly which creates an hydraulic head, and which forces the liquid waste, into the motor housing to cool the electric motor means and to cool and/or lubricate the radial and thrust bearing assemblies. Hard-on-hard bearing surfaces of the bearing assemblies and a ring assembly between the upper impeller and electric motor means grind large particles in the liquid waste flow. Slots in the static bearing member of the radial bearing assemblies further grind down the solid waste particles so that only particles smaller than the clearances in the system can pass there through, thereby resisting damage to and the interruption of the operation of the transfer pump. The column assembly is modular so that sections can be easily assembled, disassembled and/or removed. A second embodiment employs a stator jacket which provides an alternate means for cooling the electric motor means and lubricating and/or cooling the bearing assemblies, and a third embodiment employs a variable level suction device which allows liquid waste to be drawn into the transfer pump from varying and discrete levels in the waste tank. 17 figs

  2. Station blackout with reactor coolant pump seal leakage

    Evinay, A.

    1993-01-01

    The U.S. Nuclear Regulatory Commission (NRC) amended its regulations in 10CFR50 with the addition of a new section, 50.63, open-quotes Loss of All Alternating Current Power.close quotes The objective of these requirements is to ensure that all nuclear plants have the capability to withstand a station blackout (SBO) and maintain adequate reactor core cooling and containment integrity for a specified period of time. The NRC also issued Regulatory Guide (RG) 1.155, open-quotes Station Blackout,close quotes to provide guidance for meeting the requirements of 10CFR50.63. Concurrent with RG-1.155, the Nuclear Utility Management and Resources Council (NUMARC) has developed NUMARC 87-00 to address SBO-coping duration and capabilities at light water reactors. Licensees are required to submit a topical report based on NUMARC 87-00 guidelines, to demonstrate compliance with the SBO rule. One of the key compliance criteria is the ability of the plant to maintain adequate reactor coolant system (RCS) inventory to ensure core cooling for the required coping duration, assuming a leak rate of 25 gal/min per reactor coolant pump (RCP) seal in addition to technical specification (TS) leak rate

  3. Leak rate analysis of the Westinghouse Reactor Coolant Pump

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

    1985-07-01

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

  4. Reactor coolant pump seal response to loss of cooling

    Graham, T.; Metcalfe, R.; Burchett, P.

    2000-01-01

    This paper describes the results of a test done to determine the performance of a reactor coolant pump seal for a water cooled nuclear reactor under loss of all cooling conditions. Under these conditions, seal faces can lose their liquid lubricating film and elastomers can rapidly degrade. Temperatures in the seal-cartridge tester reached 230 o C in three hours, at which time the tester was stopped and the temperature increased to 265 o C for a further five hours before cooling was restored. Seal leakage was 'normal' throughout the test. Parts sustained minor damage with no effect on seal integrity. Plant operators were shown to have ample margin beyond their 15 minute allowable reaction time. (author)

  5. Operation diagnostics of the reactor coolant pumps in the Jaslovske Bohunice nuclear power plant, CSSR

    Bahna, J.; Jaros, I.; Oksa, G.

    1990-01-01

    The state of the art of the materials basis, the diagnostics methods used, organization of data collection and processing, and some results of routine and specific investigations concerned with diagnosis of the reactor coolant pump in the Jaslovske Bohunice NPP V-1 are presented. Some information is given about the reactor coolant pump monitor developed in the VUJE. (author)

  6. High-inertia hermetically sealed main coolant pump for next generation passive nuclear power plants

    Kujawski, Joseph M.; Nair, Bala R.; Vijuk, Ronald P.

    2003-01-01

    The main coolant pump for the Westinghouse AP1000 advanced passive nuclear power plant represents a significant scale-up in power, flow capacity, and physical size from its predecessor designed for the smaller AP600 power plant. More importantly, the AP1000 pump incorporates several innovative features that contribute to improved efficiency, operational reliability, and plant safety. The features include an internals design which provides the highest hydraulic efficiency achieved in commercial nuclear power plant applications. Another feature is the use of a distributed inertial mass system in the rotating assembly to develop the high rotational inertia to meet the extended system flow coastdown requirement for core heat removal in the event of loss of power to the pumps. This advanced canned motor pump also incorporates the latest development in higher operating voltage, providing plant designers with the ability to eliminate plant transformers and operate directly on the site electrical bus in many cases. The salient features of the pump design and performance data are presented in this paper. (author)

  7. Conceptual design of main coolant pump for integral reactor SMART

    Park, Jin Seok; Kim, Jong In; Kim, Min Hwan [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    The conceptual design for MCP to be installed in the integral reactor SMART was carried out. Canned motor pump was adopted in the conceptual design of MCP. Three-dimensional modeling was performed to visualize the conceptual design of the MCP and to check interferences between the parts. The theoretical design procedure for the impeller was developed. The procedures for the flow field and structural analysis of impeller was also developed to assess the design validity and to verify its structural integrity. A computer program to analyze the dynamic characteristics of the rotor shaft of MCP was developed. The rotational speed sensor was designed and its performance test was conducted to verify the possibility of operation. A prototypes of the canned motor was manufactured and tested to confirm the validity of the design concept. The MCP design concept was also investigated for fabricability by establishing the manufacturing procedures. 41 refs., 96 figs., 10 tabs. (Author)

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

    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)

  9. Statistical analysis of the Ft. Calhoun reactor coolant pump system

    Heising, Carolyn D.

    1998-01-01

    In engineering science, statistical quality control techniques have traditionally been applied to control manufacturing processes. An application to commercial nuclear power plant maintenance and control is presented that can greatly improve plant safety. As a demonstration of such an approach to plant maintenance and control, a specific system is analyzed: the reactor coolant pumps (RCPs) of the Ft. Calhoun nuclear power plant. This research uses capability analysis, Shewhart X-bar, R-charts, canonical correlation methods, and design of experiments to analyze the process for the state of statistical control. The results obtained show that six out of ten parameters are under control specifications limits and four parameters are not in the state of statistical control. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators Such a system would provide operators with ample time to respond to possible emergency situations and thus improve plant safety and reliability. (author)

  10. Statistical analysis of the Ft. Calhoun reactor coolant pump system

    Patel, Bimal; Heising, C.D.

    1997-01-01

    In engineering science, statistical quality control techniques have traditionally been applied to control manufacturing processes. An application to commercial nuclear power plant maintenance and control is presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) of the Ft. Calhoun nuclear power plant. This research uses capability analysis, Shewhart X-bar, R charts, canonical correlation methods, and design of experiments to analyze the process for the state of statistical control. The results obtained show that six out of ten parameters are under control specification limits and four parameters are not in the state of statistical control. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with ample time to respond to possible emergency situations and thus improve plant safety and reliability. (Author)

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

    Kolaei, Alireza Rezania; Rosendahl, Lasse

    2012-01-01

    the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The threedimensional governing equations for the fluid flow and the heat......The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find...... heat sink configurations reduces the coolant pumping power in the system....

  12. Technical findings related to Generic Issue 23: Reactor coolant pump seal failure

    Ruger, C.J.; Luckas, W.J. Jr.

    1989-03-01

    Reactor coolant pumps contain mechanical seals to limit the leakage of pressurized coolant from the reactor coolant system to the containment. These seals have the potential to leak, and a few have degraded and even failed resulting in a small break loss of coolant accident (LOCA). As a result, ''Reactor Coolant Pump Seal Failure,'' Generic Issue 23 was established. This report summarizes the findings of a technical investigation generated as part of the program to resolve this issue. These technical findings address the various fact-finding issue tasks developed for the action plan associated with the generic issue, namely background information on seal failure, evaluation of seal cooling, and mechanical- and maintenance-induced failure mechanisms. 46 refs., 15 figs., 14 tabs

  13. Simulations and field tests of a reactor coolant pump emergency start-up by means of remote gas units

    Omahen, P.; Gubina, F.

    1992-01-01

    The problem of the reactor coolant pump start-up in case of emergency by means of remote gas power plant units was analyzed. In this paper a simulation model is developed which enabled a detailed simulation of the transient process occurring at the start-up. The start-up of the RCP motor set was simulated in case of available one and two gas units. The field tests were performed and the measured variable values complied well with the simulation results. Two gas units have been determined as a safe start-up scheme of the RCP motor set considering for safety reasons accepted busbars and motor protection settings. A derived model for deep rotor bars was experimentally confirmed as effective means for the RCP motor set start-up transient simulation. Start-up procedures have been designed and adopted to the safety procedures of the Nuclear Power Plant Krsko

  14. Using the motor to monitor pump conditions

    Casada, D. [Oak Ridge National Lab., TN (United States)

    1996-12-01

    When the load of a mechanical device being driven by a motor changes, whether in response to changes in the overall process or changes in the performance of the driven device, the motor inherently responds. For induction motors, the current amplitude and phase angle change as the shaft load changes. By examining the details of these changes in amplitude and phase, load fluctuations of the driven device can be observed. The usefulness of the motor as a transducer to improve the understanding of devices with high torque fluctuations, such as positive displacement compressors and motor-operated valves, has been recognized and demonstrated for a number of years. On such devices as these, the spectrum of the motor current amplitude, phase, or power normally has certain characteristic peaks associated with various load components, such as the piston stroke or gear tooth meshing frequencies. Comparison and trending of the amplitudes of these peaks has been shown to provide some indication of their mechanical condition. For most centrifugal pumps, the load fluctuations are normally low in torque amplitude, and as a result, the motor experiences a correspondingly lower level of load fluctuation. However, both laboratory and field test data have demonstrated that the motor does provide insight into some important pump performance conditions, such as hydraulic stability and pump-to-motor alignment. Comparisons of other dynamic signals, such as vibration and pressure pulsation, to motor data for centrifugal pumps are provided. The effects of inadequate suction head, misalignment, mechanical and hydraulic unbalance on these signals are presented.

  15. Using the motor to monitor pump conditions

    Casada, D.

    1996-01-01

    When the load of a mechanical device being driven by a motor changes, whether in response to changes in the overall process or changes in the performance of the driven device, the motor inherently responds. For induction motors, the current amplitude and phase angle change as the shaft load changes. By examining the details of these changes in amplitude and phase, load fluctuations of the driven device can be observed. The usefulness of the motor as a transducer to improve the understanding of devices with high torque fluctuations, such as positive displacement compressors and motor-operated valves, has been recognized and demonstrated for a number of years. On such devices as these, the spectrum of the motor current amplitude, phase, or power normally has certain characteristic peaks associated with various load components, such as the piston stroke or gear tooth meshing frequencies. Comparison and trending of the amplitudes of these peaks has been shown to provide some indication of their mechanical condition. For most centrifugal pumps, the load fluctuations are normally low in torque amplitude, and as a result, the motor experiences a correspondingly lower level of load fluctuation. However, both laboratory and field test data have demonstrated that the motor does provide insight into some important pump performance conditions, such as hydraulic stability and pump-to-motor alignment. Comparisons of other dynamic signals, such as vibration and pressure pulsation, to motor data for centrifugal pumps are provided. The effects of inadequate suction head, misalignment, mechanical and hydraulic unbalance on these signals are presented

  16. Two-phase coolant pump model of pressurized light water nuclear reactors

    Santos, G.A. dos; Freitas, R.L.

    1990-01-01

    The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The homologous curves set up the complete performance of the pump and are input for accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

  17. Main-coolant-pump shaft-seal guidelines. Volume 3. Specification guidelines. Final report

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

    1983-03-01

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

  18. Influence of building and supply conditions on coolant pumps and the various coolant pump designs for cooling towers

    Holzhueter, E.; Migod, A.; Siekmann, H.

    1977-01-01

    This contribution tries to present the various factors influencing the design of cooling tower pumps. As cooling tower pumps are very often designed as concrete speral casing pumps, the suction bend construction often offers itself. The running wheel of cooling tower pumps is usually of semi-axial design, whereby one has to differ between rigid, adjustable, and resetable running wheels. Finally, the type of cooling system and the nominal width are decisive for either the construction type of the spiral casing pump or the tubular type pump. Both methods are compared in a critical way. (orig.) [de

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

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

    1993-01-01

    An expert system for determining the operability of a specified pump is described comprising: a set of pumps of which the specified pump is a member; means for measuring physical parameters representative to the operations condition each pump of said set of pumps; means for acquiring data generated by said measuring means; an artificial-intelligence based inference engine coupled to said data acquiring means where said inference engine applies a sequential probability ratio test to statistically evaluate said acquired data to determine a status for the specified pump and its respective measuring means by continually monitoring and comparing changes in a specific operational parameter signal acquired from a plurality of measurement means; means for transferring said status generated by said interference engine to an output system

  20. Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

    Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

    2000-01-01

    A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

  1. Regulatory analysis for Generic Issue 23: Reactor coolant pump seal failure. Draft report for comment

    Shaukat, S K; Jackson, J E; Thatcher, D F

    1991-04-01

    This report presents the regulatory/backfit analysis for Generic Issue 23 (GI-23), 'Reactor Coolant Pump Seal Failure'. A backfit analysis in accordance with 10 CFR 50.109 is presented in Appendix E. The proposed resolution includes quality assurance provisions for reactor coolant pump seals, instrumentation and procedures for monitoring seal performance, and provisions for seal cooling during off-normal plant conditions involving loss of all seal cooling such as station blackout. Research, technical data, and other analyses supporting the resolution of this issue are summarized in the technical findings report (NUREG/CR-4948) and cost/benefit report (NUREG/CR-5167). (author)

  2. Experience on vibration analysis of primary coolant pumps in Cirus

    Ullas, O.P.; Tilara, Manoj; Kharpate, A.V.

    2002-01-01

    Full text: 40 MW (thermal) CIRUS research reactor has been in operation for over four decades. During the major portion of its life almost all the major mechanical equipment operated continuously in a healthy condition. Since 1988 ageing related breakdown has been noticed in some of the critical components, PCW pumps being one of them. Vibration measurement and analysis is carried out on a routine basis as a part of conditioning monitoring programme. Ageing degradation of various components of the pump has been detected by such a performance monitoring programme. Conditioning monitoring has been found to be quite useful for scheduling of maintenance work on pumps

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

    Meng Qinghu; Meng Qingfeng; Feng Wuwei

    2012-01-01

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

  4. RETRAN analysis of inter-system LOCA within the primary coolant pump

    Gangadharan, A.; Pratt, G.F.

    1992-01-01

    One example of an inter-system loss of coolant accident is the failure of the tubing within the primary coolant pump (PCP) thermal barrier heat exchanger. Such a failure would result in the entry of primary coolant into the component cooling water (CCW) system. The primary coolant flowrate through the break would rapidly pressurize the CCW system when the relief valves are too small. The piping in the CCW system at Palisades has a low pressure rating. Failures in this system outside the containment boundary could lead to primary coolant release to the atmosphere. RETRAN-02 was used to perform a simulation of the break in the PCP integral heat exchanger. The model included a detailed nodalization of the Byron-Jackson primary coolant pump internals leading up to the CCW system relief valves. Preliminary studies show the need for increased relief capacity in the CCW system. A case was run using a larger relief valve. Critical flow in the system upstream of the relief valves maintains the pressures in those volumes above the CCW design pressure. The pressures downstream from the relief valves and outside containment will be at or below the design pressure. This paper presents the results of the transient analysis

  5. Fault diagnosis of main coolant pump in the nuclear power station based on the principal component analysis

    Feng Junting; Xu Mi; Wang Guizeng

    2003-01-01

    The fault diagnosis method based on principal component analysis is studied. The fault character direction storeroom of fifteen parameters abnormity is built in the simulation for the main coolant pump of nuclear power station. The measuring data are analyzed, and the results show that it is feasible for the fault diagnosis system of main coolant pump in the nuclear power station

  6. Gear-shaft linkage, especially for nuclear reactor coolant pumps

    Delaunois, T.; Lefevre, R.

    1990-01-01

    The pump comprises: - inlet and outlet channels for the pumped fluid - a rotating shaft - a gear wheel mounted on the shaft by an axial locking nut which can support the axial hydraulic force - a thermal barrier above the gear wheel. A hydrostatic bearing fitted to the exterior surround of the gear wheel, the gear shaft linkage is made by at least a centering and locating device having a cylindrical span and an axial stop and another independent device which can take up the torque [fr

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

    Rezania, A.; Rosendahl, L. A.

    2012-06-01

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

  8. Vibration of safety injection pump motors

    Wattrelos, D.

    1996-12-01

    This paper covers a fault encountered in the safety injection pump motors of the French 900 MWe unit nuclear power stations. This fault was not revealed either during the low pressure safety injection and containment spray system pump qualification tests under accident conditions or during the special tests on a test bench carried out to attempt to replicate the fault and to identify ways of remedying it. This constitutes a potential common mode of failure of the safety injection system and the containment spray system pumps. The vibration phenomena illustrate the importance of carrying out tests in the plants under conditions as close as possible to those of actual accident situations.

  9. Vibration of safety injection pump motors

    Wattrelos, D.

    1997-01-01

    This paper covers a fault encountered in the safety injection pump motors of the French 900 MWe unit nuclear power stations. This fault was not revealed either during the low pressure safety injection and containment spray system pump qualification test under accident conditions or during the special tests on a test bench carried out to attempt to replicate the fault and to identify ways of remedying it. This constitutes a potential common mode of failure of the safety injection system and the containment spray system pumps. The vibration phenomena illustrate the importance of carrying out test in the plants under conditions as close as possible to those of actual accident. (author)

  10. Vibration of safety injection pump motors

    Wattrelos, D.

    1996-01-01

    This paper covers a fault encountered in the safety injection pump motors of the French 900 MWe unit nuclear power stations. This fault was not revealed either during the low pressure safety injection and containment spray system pump qualification tests under accident conditions or during the special tests on a test bench carried out to attempt to replicate the fault and to identify ways of remedying it. This constitutes a potential common mode of failure of the safety injection system and the containment spray system pumps. The vibration phenomena illustrate the importance of carrying out tests in the plants under conditions as close as possible to those of actual accident situations

  11. Literature survey, numerical examples, and recommended design studies for main-coolant pumps. Final report

    Allaire, P.E.; Barrett, L.E.

    1982-06-01

    This report presents an up-to-date literature survey, examples of calculations of seal forces or other pump properties, and recommendations for future work pertaining to primary coolant pumps and primary recirculating pumps in the nuclear power industry. Five main areas are covered: pump impeller forces, fluid annuli, bearings, seals, and rotor calculations. The main conclusion is that forces in pump impellers is perhaps the least well understood area, seals have had some good design work done on them recently, fluid annuli effects are being discussed in the literature, bearing designs are fairly well known, and rotor calculations have been discussed widely in the literature. It should be noted, however, that usually the literature in a given area is not applied to pumps in nuclear power stations. The most immediate need for a combined theoretical and experimental design capability exists in mechanical face seals

  12. Full sized tests on a french coolant pump under two-phase flow

    Huchard, J.C.; Bore, C.; Dueymes, E.

    1997-01-01

    The French Safety Authorities required EDF to demonstrate the ability of the new N4 main coolant pump to withstand two-phase flow conditions without damage. Therefore three full sized tests, simulating a bleeding flow on the primary system, were performed on a laboratory test loop under real operating conditions (temperature = 290 deg. C, pressure = 155 b, flowrate = 7 m 3 /s; electrical power = 7 MW). The maximum value of the mean void fraction reached 75 %. The outcome of the tests is very positive: the mechanical behaviour of the main coolant pump is good, even at high void fraction. The maximum vibration levels were below the limits fixed by the manufacturer. Correlations between the mechanical behaviour of the pump and the pressure pulsation in the test loop have been found. (authors)

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

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

    1993-01-01

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

  14. Reactor coolant pump seal issues and their applicability to new reactor designs

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

    1993-01-01

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

  15. Experimental investigation of thermoelectric power generation versus coolant pumping power in a microchannel heat sink

    Kolaei, Alireza Rezania; Rosendahl, Lasse; Andreasen, Søren Juhl

    2012-01-01

    The coolant heat sinks in thermoelectric generators (TEG) play an important role in order to power generation in the energy systems. This paper explores the effective pumping power required for the TEGs cooling at five temperature difference of the hot and cold sides of the TEG. In addition......, the temperature distribution and the pressure drop in sample microchannels are considered at four sample coolant flow rates. The heat sink contains twenty plate-fin microchannels with hydraulic diameter equal to 0.93 mm. The experimental results show that there is a unique flow rate that gives maximum net-power...

  16. Development of a reactor coolant pump monitoring and diagnostic system. Progress report, June 1982-July 1983

    Morris, D.J.; Sommerfield, G.A.

    1983-12-01

    The quality of operating data has been insufficient to allow proper evaluation of theoretical reactor coolant (RC) pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables: The rotordynamic behavior of the pump shaft and related components, the internal conditions and performance of the seals, and the plant or pump operating environment (controlled by the plant operator). Interrelationships between these areas will be developed during the data collection task, scheduled to begin in October 1983 (for a full fuel cycle at Davis-Besse). This report describes system software and hardware development, testing, and installation work performed during this period. Also described is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

  17. Transient flow characteristics of nuclear reactor coolant pump in recessive cavitation transition process

    Wang Xiuli; Yuan Shouqi; Zhu Rongsheng; Yu Zhijun

    2013-01-01

    The numerical simulation calculation of the transient flow characteristics of nuclear reactor coolant pump in the recessive cavitation transition process in the nuclear reactor coolant pump impeller passage is conducted by CFX, and the transient flow characteristics of nuclear reactor coolant pump in the transition process from reducing the inlet pressure at cavitation-born conditions to NPSHc condition is studied and analyzed. The flow field analysis shows that, in the recessive cavitation transition process, the speed diversification at the inlet is relative to the bubble increasing, and makes the speed near the blade entrance increase when the bubble phase region becomes larger. The bubble generation and collapse will affect the the speed fluctuation near the entrance. The vorticity close to the blade entrance gradually increasing is influenced by the bubble phase, and the collapse of bubble generated by cavitation will reduce the vorticity from the collapse to impeller outlet. Pump asymmetric structure causes the asymmetry of the flow, velocity and outlet pressure distribution within every impeller flow passage, which cause the asymmetry of the transient radial force. From the dimensionless t/T = 0.6, the bubble phase starts to have impact on the impeller transient radial force, and results in the irregular fluctuations. (authors)

  18. Operating experience feedback report: Experience with pump seals installed in reactor coolant pumps manufactured by Byron Jackson

    Bell, L.G.; O'Reilly, P.D.

    1992-09-01

    This report examines the reactor coolant pump (RCP) seal operating experience through August 1990 at plants with Byron Jackson (B-J) RCPs. ne operating experience examined in this analysis included a review of the practice of continuing operation with a degraded seal. Plants with B-J RCPs that have had relatively good experience with their RCP seals attribute this success to a combination of different factors, including: enhanced seal QA efforts, modified/new seal designs, improved maintenance procedures and training, attention to detail, improved seal operating procedures, knowledgeable personnel involved in seal maintenance and operation, reduction in frequency of transients that stress the seals, seal handling and installation equipment designed to the appropriate precision, and maintenance of a clean seal cooling water system. As more plants have implemented corrective measures such as these, the number of B-J RCP seal failures experienced has tended to decrease. This study included a review of the practice of continued operation with a degraded seal in the case of PWR plants with Byron Jackson reactor coolant pumps. Specific factors were identified which should be addressed in order to safety manage operation of a reactor coolant pump with indications of a degrading seal

  19. Reactor coolant pump service life evaluation for current life cycle optimization and license renewal

    Doroshuk, B.W.; Berto, D.S.; Robles, M.

    1990-01-01

    This paper reports that as part of the plant life cycle management and license renewal program, Baltimore Gas and Electric Company (BG and E) has completed a service life evaluation of their reactor coolant pumps, funded jointly by EPRI and performed by ABB Combustion Engineering Nuclear Power. Two of the goals of the BG and E plant life cycle management and license renewal program, and of this current evaluation, are to identify actions which would optimize current plant operation, and ensure that license renewal remains a viable option. The reactor coolant pumps (RCPs) at BG and E's Calvert Cliffs Units 1 and 2 are Byron Jackson pumps with a diffuser and a single suction. This pump design is also used in many other nuclear plants. The RCP service life evaluation assessed the effect of all plausible age-related degradation mechanisms (ARDMs) on the RCP components. Cyclic fatigue and thermal embrittlement were two ARDMs identified as having a high potential to limit the service life of the pump case. The pump case is a primary pressure boundary component. Hence, ensuring its continued structural integrity is important

  20. Independent modification on water lubrication loop of radial-axial bearing of Russian reactor coolant pump

    Gu Yingbin

    2012-01-01

    Water lubrication was used for radial-axial bearings of 1391M reactor coolant pumps at both units of Tianwan Nuclear Power Plant Phase I Project, which was the first trial on large commercial pressurized water reactors in the world. As a prototype, there were inherent deficiencies leading to a series of operational events. Jiangsu Nuclear Power Corporation conducted the independent innovative technical modification to cope with the defects, and succeeded in reducing heat removal rate of the radial-axial bearings of the reactor coolant pumps, mitigating or preventing the cavitation abrasion of the bearings and improving the cooling effects. This paper illustrates the reasons of the innovative modification, the design and implementation preparation of modification program, the implementation process and evaluation of modification effect, including detailed follow-up work program. (author)

  1. Secondary seal effects in hydrostatic non-contact seals for reactor coolant pump shaft

    Fujita, T.; Koga, T.; Tanoue, H.; Hirabayashi, H.

    1987-01-01

    The paper presents a seal flow analysis in a hydrostatic non-contact seal for a PWR coolant pump shaft. A description is given of the non-contact seal for the reactor coolant pump. Results are presented for a distortion analysis of the seal ring, along with the seal flow characteristics and the contact pressure profiles of the secondary seals. The results of the work confirm previously reported findings that the seal ring distortion is sensitive to the o-ring location (which was placed between the ceramic seal face and the seal ring retainer). The paper concludes that the seal flow characteristics and the tracking performance depend upon the dynamic properties of the secondary seal. (U.K.)

  2. Cleaning device for recycling pump motor cooling system in nuclear reactor

    Katayama, Kenjiro; Kondo, Takahisa; Shindo, Kenjiro; Akimoto, Jun.

    1996-01-01

    The cleaning device of the present invention comprises a cleaning water supply pump, a filter for filtering the cleaning water and a cap member for isolating the inside of a motor casing from the inside of a reactor pressure vessel. A motor in the motor casing and a pump in the reactor pressure vessel are removed, the cap member is attached to the upper end of the motor casing to isolate the inside of the motor casing from the inside of the reactor pressure vessel. If the cleaning water supply pump is operated in this state, the cleaning water flows from a returning pipeline for cooling water circulation, connected to the motor casing to supply pipelines through a heat exchange and is discharged. The discharged water passes through a filter and is sent again, as the cleaning water, to the cleaning water supply pump. With such procedures, the recycling pump motor cooling system in the BWR type reactor can be cleaned without disposing a cyclone separator and irrespective of presence or absence of reactor coolants in the reactor pressure vessel. (I.N.)

  3. Main-coolant-pump shaft-seal guidelines. Volume 1. Maintenance-manual guidelines. Final report

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

    1983-03-01

    This report presents a set of guidelines and a listing of information and data which should be included in maintenance manuals and procedures for Main Coolant Pump Shaft Seals. The noted guidelines and data listing are developed from EPRI sponsored nuclear plant seal operating experience studies. The maintenance oriented results of the most recent such study is summarized. The shaft seal and its auxiliary supporting systems are discussed from both technical and maintenance related viewpoints

  4. Main-coolant-pump shaft-seal reliability investigation. Interim report

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

    1982-09-01

    This report contains the results of a survey of reactor coolant pump shaft seal reliability. The survey sample is representatively large (approx. = 27% of total US commercial plant population) and includes the three industry seal suppliers (Bingham-Williamette, Byron Jackson, and Westinghouse). Operationally incurred/induced problems and seal redesign parameters are identified. Failure hypotheses in the form of fault trees have been developed to describe the failure mechanisms. Recommendations are made for seal reliability improvement

  5. Main-coolant-pump shaft-seal guidelines. Volume 2. Operational guidelines. Final report

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

    1983-03-01

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

  6. Prediction of Hydraulic Performance of a Scaled-Down Model of SMART Reactor Coolant Pump

    Kwon, Sun Guk; Park, Jin Seok; Yu, Je Yong; Lee, Won Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-08-15

    An analysis was conducted to predict the hydraulic performance of a reactor coolant pump (RCP) of SMART at the off-design as well as design points. In order to reduce the analysis time efficiently, a single passage containing an impeller and a diffuser was considered as the computational domain. A stage scheme was used to perform a circumferential averaging of the flux on the impeller-diffuser interface. The pressure difference between the inlet and outlet of the pump was determined and was used to compute the head, efficiency, and break horse power (BHP) of a scaled-down model under conditions of steady-state incompressible flow. The predicted curves of the hydraulic performance of an RCP were similar to the typical characteristic curves of a conventional mixed-flow pump. The complex internal fluid flow of a pump, including the internal recirculation loss due to reverse flow, was observed at a low flow rate.

  7. Development for LMR coolant technology - Development of a submersible-in-pool electromagnetic pump

    Hong, Sang Hee; Kim, Hee Reyoung; Lee, Sang Don; Seo, Chun Ho [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyungki University, Suwon (Korea, Republic of)

    1995-08-01

    The conceptual and detailed designs of an annular linear induction electromagnetic pump of small scale submersible-in-pool type are performed for the purpose of domestic development of the pumps used for the high-temperature natrium coolant transportation in liquid metal reactors. The pump drawings for and input power of 1,100 VA, an input frequency of 17 Hz, a maximum flowrate of 60 l/min and a maximum operation temperature of 600 deg C are obtained from the optimum design analyses by solving MHD and equivalent circuit equations. The characteristics of pump materials in the high temperature and neutron irradiation environment are reflected in designing the pump, and theoretical analyses for improving the pump performance and efficiency are tried through calculations of magnetic flux and temperature distributions inside the pump. The present project contributes to the further design of engineering proto-type electromagnetic pump with higher capacity and the development of liquid metal reactor with innovative simplicity. 44 refs., 4 tabs., 33 figs. (author)

  8. Development of LMR Coolant Technology - Development of a submersible-in-pool electromagnetic pump

    Hong, Sang Hi; Kim, Hee Reyoung; Lee, Sang Don; Seo, Joon Ho [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyoungki University, Suwon (Korea, Republic of)

    1997-07-15

    A submersible-in-pool type annular linear induction pumps of 60 l/min and 200 l/min, and 600 deg C has been designed with optimum geometrical and operating values found from MHD and circuit analyses reflecting the high-temperature characteristics of pump materials. Through the characteristics analyses inside the narrow flow channel of electromagnetic pump, the distribution of the time-varying flow field is calculated, and magnetic flux and force density are evaluated by end effects of linear induction electromagnetic pump and the instability analyses are carried out introducing one-dimensional linear perturbation. Testing the pump with the flow rate of 60 l/min in the suitably manufactured loop system shows a flow rate of 58 l/min at an input power of 1,377 VA with 60Hz. The design of a scaled-up pump is further taken into account LMR coolant system requiring increased capacity, and a basic analysis is carried out on the pump of 40,000 l/min for KALIMER. The present project contributes to the further design of engineering prototype electromagnetic pumps with higher capacity and to the development of liquid metal reactor with innovative simplicity. 89 refs., 8 tabs., 45 figs. (author)

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

    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...

  10. Always at the correct temperature. Thermal management with electric coolant pump; Immer richtig temperiert. Thermomanagement mit elektrischer Kuehlmittelpumpe

    Genster, A.; Stephan, W. [Pierburg GmbH, Neuss (Germany)

    2004-11-01

    Through the use of the electric coolant pump it has become possible for the first time to attain a cooling performance which is adapted precisely to the engine load and which is independent of engine speed. For cooling the new BMW six cylinder in-line Otto engine with an engine power rating of 190 kW, the electric coolant pump by Pierburg requires only 200 W of electrical power from the onboard electrical system. (orig.)

  11. The empirical intensity of PWR primary coolant pumps failure and repair

    Milivojevicj, S.; Riznicj, J.

    1988-01-01

    The wealth of operating experience concerning PWR type and nuclear reactors that has been regularly monitored and systematically processes since 1971, enabled an analysis of the PWR primary coolant pumps operation. Failure intensity α and repair intensity μ of the pump during its working life were calculated, as these values are necessary in order to determine the reliability and availability of the pump as the basis for analyzing its effect on the safety and efficiency of the nuclear power plant. The trend of failure intensity α follows the theoretically expected changes in α over time, and this is around 10 -5 in the majority of life-time. Repair intensity μ indicates a slow rise during life-time, i.e. its faster return to operation. (author).7 refs.; 5 figs

  12. Moment inertia pump analysis used in the Rsg-Gas primary coolant loop under lofa condition

    Sudarmono; Setiyanto; Dhandhang, P.; Dibyo, S.; Royadi

    1998-01-01

    The moment inertia of primary cooling system analysis under LOFA condition has been done. It is potentially one of limiting design constraints of the RSG-GAS safety because the coolant flow rate reduces very rapidly under LOFA condition due to the low inertia circulation pumps. If a loss of flow accident occurs, the mass flow will decrease rapidly and the heat transfer coefficient between cladding and coolant will also decreases. As a consequence the fuel and cladding temperature will increase. The whole core was represented by the 1/4 sector and divided into 19 subchannels and 40 axial nodes. In the present study, moment inertia of pump analysis for RSG-GAS reactor was performed with COBRA-IV-I subchannel code. As the DNB correlation, W-3 Correlation was selected for base case. The flow and power transients under pump trip accident were determined from experiments. The result above compared with the design data are 75 kg m 2 and 81 Kg m 2 respectively. The result shows that the RSG-GAS requires the inertia more than 75 kg m 2

  13. Wet motor gerotor fuel pump with self-aligning bearing

    Carleton, W.A.

    1987-02-24

    A wet motor gerotor fuel pump is described for pumping fuel from a fuel source to an internal combustion engine comprising: a pump case having one end, an opposite end and a flow axis therethrough, the pump case further comprising an inlet end bore at the one end adapted to communicate with the fuel source; an inlet chamber adjacent to the inlet end bore; a motor chamber located in the opposite end of the pump case; a pump chamber interposed the motor chamber and the inlet chamber; first means for sealing the pump case, the first means for sealing located at the opposite end of the pump case; inlet housing means mounted in the pump chamber, the inlet housing means comprising an annular hub protruding into the inlet chamber. The inlet housing means further comprises a gerotor cavity about a gerotor axis located parallel to and displaced a predetermined distance in an eccentric radial direction from the flow axis.

  14. Coast-down model based on rated parameters of reactor coolant pump

    Jiang Maohua; Zou Zhichao; Wang Pengfei; Ruan Xiaodong

    2014-01-01

    For a sudden loss of power in reactor coolant pump (RCP), a calculation model of rotor speed and flow characteristics based on rated parameters was studied. The derived model was verified by comparing with the power-off experimental data of 100D RCP. The results indicate that it can be used in preliminary design calculation and verification analysis. Then a design criterion of RCP was described based on the calculation model. The moment of inertia in AP1000 RCP was verified by this criterion. (authors)

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

    Peyrouty, P.

    1997-01-01

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

  16. Unbalance response and stability analyses of the rotor of SMART main coolant pump

    Park, J. H.; Park, J. S.; Kim, J. I.

    2001-01-01

    SMART main coolant pump(MCP) is being designed as a vertical type and the rotor is operated immersed in hot and high pressure water. Hydraulic forces which are taken place at journal bearings, impellers and gaps between rotor and housing are inherently highly nonlinear and have unstable characteristics. Furthermore, since vertical rotor rather than horizontal type has no dominant static bearing load such as one's weight, traveling of journal center in the clearance circle of the bearing as varying of rotational speed make change in rotor characteristics greatly. In this paper, MCP rotor dynamic characteristics are estimated relating in hydraulic forces at journal bearings and gaps

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

    Peyrouty, P.

    1996-12-01

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

  18. Analytical prediction on the pump-induced pulsating pressure in a reactor coolant pipe

    Lee, K.B.; Im, I.Y.; Lee, S.K.

    1992-01-01

    An analytical method is presented for predicting the amplitudes of pump-induced fluctuating pressures in a reactor coolant pipe using a linear transformation technique which reduces a homogeneous differential equation with non-homogeneous boundary conditions into a nonhomogeneous differential equation with homogeneous boundary conditions. At the end of the pipe, three types of boundary conditions are considered-open, closed and piston-spring supported. Numerical examples are given for a typical reactor. Comparisons of measured pressure amplitudes in the pipe with model prediction are shown to be in good agreement for the forcing frequencies. (author)

  19. Consequences in the pumps operation during a large loss of coolant accident

    Santos, G.A. dos; Sabundjian, G.

    1991-08-01

    The event of living on or turning off the operation of the Reactor Cooling Pumps - RCPs, in the case of a Loss of Coolant Accident - LOCA, has been a reason of a lot of studies after the Three Mile Island 2 accident. Thus, it was investigated a large break LOCA in the cold leg of Angra 1, with the RELAP4/MOD5 Code during the blowdown. The attained results indicated that the best performance of the core was in the case where the RCPs had been turned off in the beginning of the transient, when compared with different operation conditions of the RCPs. (author)

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

    Peyrouty, P.

    1996-01-01

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

  1. Heating great residential units with combustion-motor heat pumps

    Vossen, W

    1982-10-01

    Economic usage of combustion-motor heat pumps requires: reliable technology and delivery of the heat pump; design and operation. The heat pump must be integrated perfectly into the heating system. This contributions is based on a three-year operational experience with over 150 heat pumps used mainly in residential and administrative buildings (plus commercial buildings, swimming pools, sport centres etc.). These are heat pumps operating on the compression principle with natural gas, liquid gas, or fuel oil.

  2. Diapo, applying advanced AI methods to diagnosis of PWR reactor coolant pump

    Porcheron, M.; Ricard, B.

    1993-01-01

    Electricite de France has decided to increase the capabilities of its monitoring and diagnostic architecture with the development of an AI system for reactor coolant pump diagnostic support. This development is carried out with the cooperation of the equipment constructor Jeumont Schneider Industries. This diagnostic system will eventually be included in an integrated surveillance architecture. We present the architecture of the system and the basics of the knowledge model used. Main data for diagnosis are provided by sensor data issued by the pump monitoring system. Diagnostic reasoning is based on the cooperation of two main activities : a heuristic search among typical symptomatic situations that leads to the formulation of hypotheses and a ''deep'' causal analysis that consists in backtracking from identified situations up to initial faults or causes. This approach is well fitted to field expert reasoning, and provides powerful diagnostic capabilities that help to overcome conventional limitations of expert systems entirely based on heuristic knowledge. (authors). 9 figs., 11 refs

  3. Design, construction and testing of replacement nuclear coolant pump stators to meet today's equipment reliability expectations

    Fostier, L.; Howell, D.

    2005-01-01

    The reliability expectations of equipment and components in today's nuclear power plant are much greater than three or more decades ago when nuclear plants were first constructed due to economic impact of a failure. Very few components in a pressurized water reactor plant can have as much impact of the plants capacity factor as a catastrophic failure of a reactor coolant pump winding. This paper describes the maintenance approach taken by one North American utility in attempt to preclude such failures. The paper will discuss the challenges of the reactor coolant pump application and the enhancements made in the winding design and construction by the supplier to address failure mechanisms so as to better meet present reliability expectations in accordance with dedicated specifications. The paper will also present the in-process and final testing requirements and limits imposed in an attempt to ensure quality of the machine windings, along with selected test results from the stators that have been designed and constructed to these specifications to date. (author)

  4. Analysis on transient hydrodynamic characteristics of cavitation process for reactor coolant pump

    Wang Xiuli; Wang Peng; Yuan Shouqi; Zhu Rongsheng; Fu Qiang

    2014-01-01

    The reactor coolant pump hydrodynamic characteristics at different cavitation conditions were studied by using flow field analysis software ANSYS CFX, and the corresponding data were processed and analyzed by using Morlet wavelet transform and fast Fourier transform. The results show that gas content presents the law of exponential function with the pressure reduction or time increase. In the cavitation primary condition, the pulsation frequency of head for the reactor coolant pump is mainly low frequency, and the main frequency of pressure pulsation is still rotation frequency while the effect of the pressure pulsation caused by cavitation on main frequency is not obvious. With the development of cavitation, the pressure fluctuation induced by cavitation becomes more serious especially for the main frequency, secondary frequency and pulsating amplitude while the head pulsation frequency is given priority to low frequency pulse. Under serious cavitation condition, the head pulsation frequency is given priority to irregular changes of pulse high frequency, and also contains almost regular changes of low frequency. (authors)

  5. Analysis of Pressure Pulsation Induced by Rotor-Stator Interaction in Nuclear Reactor Coolant Pump

    Xu Zhang

    2017-01-01

    Full Text Available The internal flow of reactor coolant pump (RCP is much more complex than the flow of a general mixed-flow pump due to high temperature, high pressure, and large flow rate. The pressure pulsation that is induced by rotor-stator interaction (RSI has significant effects on the performance of pump; therefore, it is necessary to figure out the distribution and propagation characteristics of pressure pulsation in the pump. The study uses CFD method to calculate the behavior of the flow. Results show that the amplitudes of pressure pulsation get the maximum between the rotor and stator, and the dissipation rate of pressure pulsation in impellers passage is larger than that in guide vanes passage. The behavior is associated with the frequency of pressure wave in different regions. The flow rate distribution is influenced by the operating conditions. The study finds that, at nominal flow, the flow rate distribution in guide vanes is relatively uniform and the pressure pulsation amplitude is the smallest. Besides, the vortex shedding or backflow from the impeller blade exit has the same frequency as pressure pulsation but there are phase differences, and it has been confirmed that the absolute value of phase differences reflects the vorticity intensity.

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

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

    2015-01-01

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

  7. One-phase and two-phase homologous curves for coolant pumps of the pressurized light water nuclear reactors

    Santos, G.A. dos.

    1990-01-01

    The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The single-phase pump characteristics are an essential feature for operational transients studies, for example, the shut-down and start-up of pump. These parameters, in terms of the homologous curves, set up the complete performance of the pump and are input for transients and accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the single-phase and two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

  8. Development of a reactor-coolant-pump monitoring and diagnostic system. Semi-annual progress report, December 1981-May 1982

    Morris, D.J.; Gabler, H.C.

    1982-10-01

    Reactor coolant (RC) pump seal failures have resulted in excessive leakage of primary coolant into reactor containment buildings. In some cases, high levels of airborne activity and surface contamination following these failures have necessitated extensive cleanup efforts and personnel radiation exposure. Unpredictable pump seal performance has also caused forced outages and frequent maintenance. The quality of operating data has been insufficient to allow proper evaluation of theoretical RC pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables. This report describes system software and hardware development, testing, and installation work performed during the period of December 1981 through May 1982. Also described herein is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

  9. Canned motor pumps at Heavy Water Project, Baroda

    Batra, R.K.; Waishampayan, S.C.

    1981-01-01

    Pumps to be used in heavy water plants must be reliable and should require negligible maintenance, because most of them are totally unapproachable under normal circumstances. Canned motor pumps fulfil these requirements. Their design features are described briefly. The details of: (1) the pumps in the isotopic exchange tower and (2) pumps for liquid ammonia and catalyst are given. Problems faced during commissioning of such pumps in Baroda Heavy Water Project were bulging of rotors of tower pumps, bulging of stators, jamming and failure of bearings. Solution of these problems is described. (M.G.B.)

  10. Apparatus for servicing an eleongated suspended pump motor in an electric power plant with limited access

    Chavez, R.; Johnson, F.T.; Ekeroth, D.E.; Matusz, J.M.

    1994-01-01

    Elongated coolant pump motors suspended under steam generators within containment in a power plant with limited access space, are removed and replaced by an upright elongated maintenance cart with an elongated opening along one side in which the motor is received. Rollers support the cart for conveying the elongated motor in an upright position out from under the steam generator and onto an elevator. The elevator is lowered to transfer support of the cart and motor through trunnions to saddles straddling the elevator for rotation of the cart to a generally horizontal position. The elevator then raises the horizontally disposed cart carrying the motor to a higher floor where it is rolled off the elevator on rollers and out through the auxiliary equipment hatch. The cart includes a top V-shaped collar for supporting the motor, and a further lower support cradle operative when the cart is horizontal. Jacks support the motor during unbolting from the pump casing and lower it onto the cart. (Author)

  11. Effect of the Shrink Fit and Mechanical Tolerance on Reactor Coolant Pump Flywheel Integrity Evaluation

    Kim, Donghak [Korea KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Reactor coolant pump (RCP) flywheel should satisfy the RCP flywheel integrity criteria of the US NRC standard review plan (SRP) 5.4.1.1 and regulatory guide (RG) 1.14. Shrink-fit and rotational stresses should be calculated to evaluate the integrity. In this paper the effects of the shrink fit and mechanical tolerance on the RCP flywheel integrity evaluation are studied. The shrink fit should be determined by the joint release speed and the stresses in the flywheel will be increased by the shrink fit. The stress at the interface between the hub and the outer wheel shows the highest value. The effect of the mechanical tolerance should be considered for the stress evaluation. And the effect of the mechanical tolerance should be not considered to determine the joint release speed.

  12. Effect of the Shrink Fit and Mechanical Tolerance on Reactor Coolant Pump Flywheel Integrity Evaluation

    Kim, Donghak

    2015-01-01

    Reactor coolant pump (RCP) flywheel should satisfy the RCP flywheel integrity criteria of the US NRC standard review plan (SRP) 5.4.1.1 and regulatory guide (RG) 1.14. Shrink-fit and rotational stresses should be calculated to evaluate the integrity. In this paper the effects of the shrink fit and mechanical tolerance on the RCP flywheel integrity evaluation are studied. The shrink fit should be determined by the joint release speed and the stresses in the flywheel will be increased by the shrink fit. The stress at the interface between the hub and the outer wheel shows the highest value. The effect of the mechanical tolerance should be considered for the stress evaluation. And the effect of the mechanical tolerance should be not considered to determine the joint release speed

  13. Summary of failed reactor coolant pump rotating assembly experience at Crystal River Unit 3

    Hayner, G.O.; Clary, M.D.

    1992-01-01

    Four reactor coolant pump (RCP) rotating assemblies (shafts) have failed or have severely cracked during operation at the Crystal River Unit 3 (CR-3) Nuclear Power Plant. The two failed shafts removed from RCP-1A have been extensively examined. All of the RCP shafts (except the D shaft) were fabricated from UNS S66286 superalloy (Alloy A-286). The D shaft was fabricated from UNS S20910 (Alloy XM-19/Nitronic 50). Torsional strain gauge analysis was performed on the RCP-1A shaft during the 1990 refueling outage. This type of analysis has not been performed previously on an operating RCP. Several results were found including: (1) the primary components of alternating torsional stress during normal RCP operation are impeller vane pass and a sub-2X torsional resonance with maximum components of ∼±0.8 ksi; (2) a typical vane pass cycle is initiated by an abrupt unloading of the shaft followed by a reload past equilibrium and a damped return to equilibrium; (3) a higher (compared to normal four pump operation) alternating torsional stress range resulted from solo operation of RCP-1A at low temperature and pressure (normal startup conditions); (4) the 2/0 combination produced the highest mean torsional stresses and the lowest alternating stresses and (5) a startup of a secured RCP with three operating pumps results in significantly higher alternating stress than a cold startup. The root cause RCP failure mechanism appears to involve RCP startup sequence at CR-3, peculiarities that necessitate this sequence and complex shaft stresses just above or under the journal bearing. The 1986 impeller bolt failure is not considered to be a root cause effect. It was also determined that fatigue cracking has always been responsible for both shaft initiation and propagation mechanisms and cracking can occur independent of shaft material

  14. Induction Motors Most Efficient Operation Points in Pumped Storage Systems

    Busca-Forcos, Andreea; Marinescu, Corneliu; Busca, Cristian

    2015-01-01

    A clear focus is nowadays on developing and improving the energy storage technologies. Pumped storage is a well-established one, and is capable of enhancing the integration of renewable energy sources. Pumped storage has an efficiency between 70-80%, and each of its elements affects it. Increased...... efficiency is desired especially when operating with renewable energy systems, which present low energy conversion factor (up to 50% - performance coefficient for wind turbines, and efficiency up to 40% for photovoltaic systems). In this paper the most efficient operation points of the induction motors...... in pumped storage systems are established. The variable speed operation of the pumped storage systems and motor loading conditions for pump applications have been the key factors for achieving the purpose of the paper....

  15. Segmentation of turbo generator and reactor coolant pump vibratory patterns: a syntactic pattern recognition approach

    Tira, Z.

    1993-02-01

    This study was undertaken in the context of turbogenerator and reactor coolant pump vibration surveillance. Vibration meters are used to monitor equipment condition. An anomaly will modify the signal mean. At the present time, the expert system DIVA, developed to automate diagnosis, requests the operator to identify the nature of the pattern change thus indicated. In order to minimize operator intervention, we have to automate on the one hand classification and on the other hand, detection and segmentation of the patterns. The purpose of this study is to develop a new automatic system for the segmentation and classification of signals. The segmentation is based on syntactic pattern recognition. For the classification, a decision tree is used. The signals to process are the rms values of the vibrations measured on rotating machines. These signals are randomly sampled. All processing is automatic and no a priori statistical knowledge on the signals is required. The segmentation performances are assessed by tests on vibratory signals. (author). 31 figs

  16. Team training using full-scale reactor coolant pump seal mock-ups

    McDonald, T.J.; Hamill, R.W.

    1987-01-01

    The use of full-scale reactor coolant pump (RCP) seal mock-ups has greatly enhanced Northeast Utilities' ability to effectively utilize the team training approach to technical training. With the advent of the Institute of Nuclear Power Operations accreditation come a new emphasis and standards for the integrated training of plant engineering personnel, maintenance mechanics, quality control personnel, and health physics personnel. The results of purchasing full-scale RCP mock-ups to pilot the concept of team training have far exceeded expectations and cost-limiting factors. The initial training program analysis identified RCP seal maintenance as a task that required training for maintenance department personnel. Due to radiation exposure considerations and the unavailability of actual plant equipment for training purposes, the decision was made to procure a mock-up of an RCP seal assembly and housing. This mock-up was designed to facilitate seal cartridge removal, disassembly, assembly, and installation, duplicating all internal components of the seal cartridge and housing area in exact detail

  17. RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

    Bencik, V.; Debrecin, N.; Foretic, D.

    2003-01-01

    In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

  18. Numerical Simulation of Three-Dimensional Flow Through Full Passage and Performance Prediction of Nuclear Reactor Coolant Pump

    Li Ying; Zhou Wenxia; Zhang Jige; Wang Dezhong

    2009-01-01

    In order to achieve the level of self-design and domestic manufacture of the reactor coolant pump (nuclear main pump), the software FLUENT was used to simulate the three-dimensional flow through full passage of one nuclear main pump basing on RNG κ-ε turbulence model and SIMPLE algorithm. The distribution of pressure and velocity of the flow in the impeller's surface was analyzed in different working conditions. Moreover, the performance of the pump was predicted based on the simulation results. The results show that the distributions of pressure and velocity are reasonable in both the working and back face of the blade in the steady working condition. The pressure of the flow is increased from the inlet to the outlet of the pump, and shows the maximal value in the impeller region. Comparatively satisfactory efficiency and head value were obtained in the condition of the pump design. The shaft power of the nuclear main pump is gradually increased with the increase of the flow flux. These results are helpful in understanding the change of the internal flow field in the nuclear main pump, which is of some importance for the pre-exploration and theoretical research on the domestic manufacture of the nuclear main pump. (authors)

  19. Refurbishment of primary coolant pump stuffing boxes for RAPS-1,2

    Rshikesan, P.B.; Shirolkar, K.M.; Ahmad, S.N.

    2006-01-01

    Primary coolant pumps (PCPs) are the most critical equipment in PHWR and stuffing box is one of the critical parts of the PCP. The stuffing box houses the mechanical seals, radial bearings, throttle bushings and stationary part of wearing ring. During overhauling of PCPs it was observed that the cracks are developing on the inside face of the stuffing box and at the bolt holes where the lower bearing housing is fixed. Since consequence of failure of stuffing box will be a break in primary system boundary a detailed investigation was carried out to find out cause of failure. An immediate procurement of these from OEM (Original Equipment Manufacturer) was not feasible and indigenous procurement of such a large and precision-machined PCP component would have called for extensive development work. Under the circumstances, the only immediate option left was to repair and re-use these failed stuffing boxes. However, repair of these stuffing boxes was considered to be very difficult job as weld repair could cause distortion and any other option was not found suitable. Since the industry was not geared up to produce such components, a decision to carry out a heavy weld build up after removing the cracks up to root, was taken after considering various other options. Major weld repair and subsequent machining was carried out successfully on four stuffing boxes and subsequently these have been put in to service. The paper covers the investigations done, various options considered, how the weld repairs were carried out and the salient features of the indigenous development taken up. (author)

  20. Flow Rate In Microfluidic Pumps As A Function Of Tension and Pump Motor Head Speed

    Irwin, Anthony; McBride, Krista

    2015-03-01

    As the use of microfluidic devices has become more common in recent years the need for standardization within the pump systems has grown. The pumps are ball bearing rotor microfluidic pumps and work off the idea of peristalsis. The rapid contraction and relaxation propagating down a tube or a microfluidic channel. The ball bearings compress the tube (occlusion) and move along part of the tube length forcing fluid to move inside of the tube in the same direction of the ball bearings. When the ball bearing rolls off the area occupied by the microfluidic channel, its walls and ceiling undergo restitution and a pocket of low pressure is briefly formed pulling more of the liquid into the pump system. Before looking to standardize the pump systems it must be known how the tension placed by the pumps bearing heads onto the PDMS inserts channels affect the pumps performance (mainly the flow rate produced). The relationship of the speed at which the bearings on the motor head spin and the flow rate must also be established. This research produced calibration curves for flow rate vs. tension and rpm. These calibration curves allow the devices to be set to optimal user settings by simply varying either the motor head tension or the motor head speed. I would like to acknowledge the help and support of Vanderbilt University SyBBURE program, Christina Marasco, Stacy Sherod, Franck Block and Krista McBride.

  1. Impact of mechanical- and maintenance-induced failures of main reactor coolant pump seals on plant safety

    Azarm, M.A.; Boccio, J.L.; Mitra, S.

    1985-12-01

    This document presents an investigation of the safety impact resulting from mechanical- and maintenance-induced reactor coolant pump (RCP) seal failures in nuclear power plants. A data survey of the pump seal failures for existing nuclear power plants in the US from several available sources was performed. The annual frequency of pump seal failures in a nuclear power plant was estimated based on the concept of hazard rate and dependency evaluation. The conditional probability of various sizes of leak rates given seal failures was then evaluated. The safety impact of RCP seal failures, in terms of contribution to plant core-melt frequency, was also evaluated for three nuclear power plants. For leak rates below the normal makeup capacity and the impact of plant safety were discussed qualitatively, whereas for leak rates beyond the normal make up capacity, formal PRA methodologies were applied. 22 refs., 17 figs., 19 tabs

  2. Design and instrumentation of an automotive heat pump system using ambient air, engine coolant and exhaust gas as a heat source

    Hosoz, M.; Direk, M.; Yigit, K.S.; Canakci, M.; Alptekin, E.; Turkcan, A.

    2009-01-01

    Because the amount of waste heat used for comfort heating of the passenger compartment in motor vehicles decreases continuously as a result of the increasing engine efficiencies originating from recent developments in internal combustion engine technology, it is estimated that heat requirement of the passenger compartment in vehicles using future generation diesel engines will not be met by the waste heat taken from the engine coolant. The automotive heat pump (AHP) system can heat the passenger compartment individually, or it can support the present heating system of the vehicle. The AHP system can also be employed in electric vehicles, which do not have waste heat, as well as vehicles driven by a fuel cell. The authors of this paper observed that such an AHP system using ambient air as a heat source could not meet the heat requirement of the compartment when ambient temperature was extremely low. The reason is the decrease in the amount of heat taken from the ambient air as a result of low evaporating temperatures. Furthermore, the moisture condensed from air freezed on the evaporator surface, thus blocking the air flow through it. This problem can be solved by using the heat of engine coolant or exhaust gases. In this case, the AHP system can have a higher heating capacity and reuse waste heat. (author)

  3. Motor Pump Revolution in Ethiopia: Promises at a Crossroads

    Mengistu Dessalegn

    2015-06-01

    Full Text Available In sub-Saharan Africa, motor pump irrigation is at an earlier stage than in Asia but is growing rapidly in many countries. The focus of both policy and research in Africa to date has been on facilitating supply chains to make pumps available at a reasonable price. In Africa, pump irrigation is mainly based on two sources: shallow groundwater aquifers and small streams and rivers. Both usually have limited and variable yields. We present a case study from Ethiopia where pump irrigation based on small rivers and streams is expanding rapidly, and draw parallels to experiences in Asia and other African countries. We show that while farmers understand the social nature of community-managed irrigation, they share with policymakers a narrow understanding of pump irrigation as being primarily 'technical'. They perceive pumps as liberating them from the 'social' limitations of traditional communal irrigation. However, the rapid expansion of pump irrigation is leading to increasing competition and conflict over the limited water resource. We analyse the wider implications for Africa of this blindness to the social dimension of pump irrigation and offer suggestions on future policy and applied research to address the problem before it becomes a widespread crisis.

  4. Verification of control system using inverter and canned motor pump

    Sawada, Yoshiaki; Misato, Hisashi

    2002-01-01

    Control on flow volume and so on of auxiliary systems at power stations is generally carried out by using control valves (CVs), of which numbers and kinds ranges to wide areas. CVs are required for periodical change of packing and so on, of which labor for maintenance is never few. Therefore, to reduce the maintenance of CVs, a system to operate pumps by using an inverter control was investigated. When carrying out flow control by an inverter, valves at output side of pumps was made perfectly open, but because of control on rotation numbers so as to keep required amount excess energy is never consumed. And, by reducing flow volume of a pump, consumed energy is reduced at a rate of its three powers as feature of pumps, so large energy saving effect can be established. Selected canned motor pumps have such characteristics as upgrading of reliability for leakage because of their seal-less ones and extension of periodical inspection period by setting a monitor for abrasion of bearings. As results of some investigations, it could be considered that a control system combining an inverter with a canned motor pump had equal feature as that of a control system using CVs. And, from a test result adding useless time and first order delay element to its control feature forecasting on its application to practical machine could be obtained. (G.K.)

  5. Apparatus and method for servicing an elongated suspended pump motor in an electric power plant with limited access

    Chavez, R.V.; Ekeroth, D.E.; Johnson, F.T.; Matusz, J.M.

    1994-01-01

    Elongated coolant pumps suspended under steam generators within containment in a power plant with limited access space, are removed and replaced by an elongated maintenance cart with an elongated opening along one side in which the motor is received. Rollers support the cart for conveying the elongated motor in an upright position out from under the steam generator and onto an elevator. The elevator is lowered to transfer support of the cart and motor through trunnions to saddles straddling the elevator for rotation of the cart to a generally horizontal position. The elevator then raises the horizontally disposed cart carrying the motor to a higher floor where it is rolled off the elevator and out through the auxiliary equipment hatch. 14 figures

  6. Particle image velocimetry measurement of complex flow structures in the diffuser and spherical casing of a reactor coolant pump

    Yongchao Zhang

    2018-04-01

    Full Text Available Understanding of turbulent flow in the reactor coolant pump (RCP is a premise of the optimal design of the RCP. Flow structures in the RCP, in view of the specially devised spherical casing, are more complicated than those associated with conventional pumps. Hitherto, knowledge of the flow characteristics of the RCP has been far from sufficient. Research into the nonintrusive measurement of the internal flow of the RCP has rarely been reported. In the present study, flow measurement using particle image velocimetry is implemented to reveal flow features of the RCP model. Velocity and vorticity distributions in the diffuser and spherical casing are obtained. The results illuminate the complexity of the flows in the RCP. Near the lower end of the discharge nozzle, three-dimensional swirling flows and flow separation are evident. In the diffuser, the imparity of the velocity profile with respect to different axial cross sections is verified, and the velocity increases gradually from the shroud to the hub. In the casing, velocity distribution is nonuniform over the circumferential direction. Vortices shed consistently from the diffuser blade trailing edge. The experimental results lend sound support for the optimal design of the RCP and provide validation of relevant numerical algorithms. Keywords: Diffuser, Flow Structures, Particle Image Velocimetry, Reactor Coolant Pump, Spherical Casing, Velocity Distribution

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

    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.

  8. Transient simulation of coolant peak temperature due to prolonged fan and/or water pump operation after the vehicle is keyed-off

    Pang, Suh Chyn; Masjuki, Haji Hassan; Kalam, Md. Abul; Hazrat, Md. Ali

    2014-01-01

    Automotive designers should design a robust engine cooling system which works well in both normal and severe driving conditions. When vehicles are keyed-off suddenly after some distance of hill-climbing driving, the coolant temperature tends to increase drastically. This is because heat soak in the engine could not be transferred away in a timely manner, as both the water pump and cooling fan stop working after the vehicle is keyed-off. In this research, we aimed to visualize the coolant temperature trend over time before and after the vehicles were keyed-off. In order to prevent coolant temperature from exceeding its boiling point and jeopardizing engine life, a numerical model was further tested with prolonged fan and/or water pump operation after keying-off. One dimensional thermal-fluid simulation was exploited to model the vehicle's cooling system. The behaviour of engine heat, air flow, and coolant flow over time were varied to observe the corresponding transient coolant temperatures. The robustness of this model was proven by validation with industry field test data. The numerical results provided sensible insights into the proposed solution. In short, prolonging fan operation for 500 s and prolonging both fan and water pump operation for 300 s could reduce coolant peak temperature efficiently. The physical implementation plan and benefits yielded from implementation of the electrical fan and electrical water pump are discussed.

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

    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

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

    Lefèvre Grégory

    2012-01-01

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

  11. Effects of the reactor coolant pumps following a small break in a Westinghouse PWR

    Koenig, J.E.

    1983-10-01

    Numerical simulations of the thermal-hydraulic events following a small cold-leg break in a Westinghouse pressurized water reactor were performed to address the pumps-on/off issue. The mode of pump operation was varied in each calculation to ascertain the optimum mode. It was found that pump operation was not critical for this break size and location because the fuel rods remained cool in all accidents analyzed. In terms of system mass, however, it was preferable to leave the pumps in operation

  12. Operating experience with an on-line vibration control system for PWR main coolant pumps

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1996-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors, because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW el NPP's with PWR and has been successfully tested in a 350 MW el NPP with a PWR. (orig.)

  13. Operating experience with an on-line vibration control system for PWR main coolant pumps

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1998-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors (PWRs), because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW e1 NPP's with PWR and has been successfully tested in a 350 MW e1 NPP with a PWR. (orig.)

  14. The main pump motor remote visual check in the application of the domestic nuclear power plants

    Ge Lianwei; Yu Tao; Fang Jiang; Zhang Ting; Zhang Xingtian; Ding Youyuan

    2014-01-01

    In this paper, the Qinshan nuclear power station the first main pump motor to the successful implementation of remote visual inspection the main pump motor remote visual inspection applications. Qinshan Nuclear Power Plant Units 1 and 2 of the main pump motor inspection results show that the key components of the Qinshan Nuclear Power Plant Units 1 and 2 of the main pump rotor, stator end coils good condition, its problems for 10 years in the motor does not affect the normal use of the motor state disintegration overhaul problems tracking disintegration overhaul in 10 years. (authors)

  15. Evaluation on Safety of Stainless Steels in Chemical Decontamination Process with Immersion Type of Reactor Coolant Pump for Nuclear Reactor

    Kim, Seong Jong; Han, Min Su; Jang, Seok Ki; Kim, Ki Joon

    2011-01-01

    Due to commercialization of nuclear power, most countries have taken interest in decontamination process of nuclear power plant and tried to develop a optimum process. Because open literature of the decontamination process are rare, it is hard to obtain skills on decontamination of foreign country and it is necessarily to develop proper chemical decontamination process system in Korea. In this study, applicable possibility in chemical decontamination for reactor coolant pump (RCP) was investigated for the various stainless steels. The stainless steel (STS) 304 showed the best electrochemical properties for corrosion resistance and the lowest weight loss ratio in chemical decontamination process with immersion type than other materials. However, the pitting corrosion was generated in both STS 415 and STS 431 with the increasing numbers of cycle. The intergranular corrosion in STS 431 was sporadically observed. The sizes of their pitting corrosion also increased with increasing cycle numbers

  16. Thermal hydraulic conditions inducing incipient cracking in the 900 MWe unit 93 D reactor coolant pump shafts

    Bore, C.

    1995-01-01

    From 1987, 900 MWe plant operating feedback revealed cracking in the lower part of the reactor coolant pump shafts, beneath the thermal ring. Metallurgical examinations established that this was due to a thermal fatigue phenomenon known as thermal crazing, occurring after a large number of cycles. Analysis of thermal hydraulic conditions initiating the cracks does not allow exact quantification of the thermal load inducing cracking. Only qualitative analyses are thus possible, the first of which, undertaken by the pump manufacturer, Jeumont Industrie, showed that the cracks could not be due to the major transients (stop-start, injection cut-off), which were too few in number. Another explanation was then put forward: the thermal ring, shrunk onto the shaft it is required to protect against thermal shocks, loosens to allow an alternating downflow of cold water from the shaft seals and an upflow of hot water from the primary system. However, approximate calculations showed that the flow involved would be too slight to initiate the cracking observed. A more stringent analysis undertaken with the 2D flow analysis code MELODIE subsequently refuted the possibility of alternating flows beneath the ring establishing that only a hot water upflow occurred due to a 'viscosity pump' phenomenon. Crack initiation was finally considered to be due to flowrate variations beneath the ring, with the associated temperature fluctuations. This flowrate fluctuation could be due to an unidentified transient phenomenon or to a variation in pump operating conditions. This analysis of the hydraulic conditions initiating the cracks disregards shaft surface residual stresses. These are tensile stresses and show that loads less penalizing than those initially retained could cause incipient cracking. Thermal ring modifications to reduce these risks were proposed and implemented. In addition, final metallurgical treatment of the shafts was altered and implemented. In addition, final metallurgical

  17. CNT based thermal Brownian motor to pump water in nanodevices

    Oyarzua, Elton; Zambrano, Harvey; Walther, Jens Honore

    2016-01-01

    asymmetry drive the water ow in a preferential direction. We systematically modified the magnitude of the applied thermal gradient and the axial position of the fixed points. The analysis involves measurement of the vibrational modes in the CNTs using a Fast Fourier Transform (FFT) algorithm. We observed......Brownian molecular motors are nanoscale machines that exploit thermal fluctuations for directional motion by employing mechanisms such as the Feynman-Smoluchowski ratchet. In this study, using Non Equilibrium Molecular Dynamics, we propose a novel thermal Brownian motor for pumping water through...... Carbon Nanotubes (CNTs). To achieve this we impose a thermal gradient along the axis of a CNT filled with water and impose, in addition, a spatial asymmetry by flxing specific zones on the CNT in order to modify the vibrational modes of the CNT. We find that the temperature gradient and imposed spatial...

  18. Fracture assessment of a main reactor coolant pump in a BWR with encountered defects

    Brickstad, B [Swedish Plant Inspectorate, Stockholm (Sweden)

    1988-12-31

    This document presents a case-study fracture assessment in BWR type reactor components. A cast stainless steel presenting defects due to thermal is studied. The stress analysis performed by aid of a finite element technique shows that a Leak Before Break situation could be expected. Nevertheless, it may be concluded that the cross section of the pump where the defect area was located can withstand very deep cracks before the risk of failure becomes unacceptable. (TEC).

  19. The influence of slightly different main circulation pumps on PWR coolant pressure pulsations

    Dach, K.; Pecinka, L.

    1989-01-01

    Pressure distribution along the core barrel circumference caused by the simultaneous operation of six main circulating pumps with slightly different revolutions obtained as a result of measurement in operated NPP is determined on the basis of the well-known Penzes method based on the solving of the wave equation with source term using the expansion into the infinite series of eigenfunctions. Results of calculations can be summarized as follows: the pressure distribution and the resulting force acting on the core barrel has a random character. The same is valid for core barrel vibrations and mainly for the joint between core barrel and pressure vessel. (orig.)

  20. Increasing Mud Pump Motor Reliability against Malfunctions of DC Motor Excitation System

    Nikulin, O. V.; Shabanov, V. A.

    2017-10-01

    The most widely used drilling machinery, such as mud pumps, draw-works, and rotors, use direct-current (DC) motors with independent excitation as the electric drive. Drilling machinery drives operate in harsh ambient conditions, including those with the presence of moisture, dust and vibration, which increases the malfunction rate of both drilling equipment and their electric drives. One of the frequently encountered malfunctions are DC motor excitation coil faults, which disrupt the normal functioning of electric drives, often leading to shutdown of the drilling process. In a four-pole DC motor, the malfunction of one coil leads to lack of excitation current in just one coil pair, while the other pair remains functional. In this case, DC motors and drilling equipment can remain operational, which would allow for continuing the drilling process. This paper considers the possibility of operation of a DC motor on a drilling rig in those cases when one pair of excitation coils is non-functional, and describes the device for switching between the excitation coils and the auxiliary winding in a DC motor with independent excitation.

  1. High performance magnetic bearings suitable for noise cancellation in permanent magnet motor driven pumps

    Zmood, R.; Cholewka, J.; Patak, C.; Feng, G.; Zhang, C.; Maleri, T.; Pinder, B.; McDonald, R.; Homrigh, J.

    1991-01-01

    Conventional pumps having external drive motors experience problems due to bearing noise. In addition failure of bearings and seals can lead to limited operational reliability and impaired integrity of these pumps. Pumps using DC brushless motors and magnetic bearings offer means of overcoming these problems. A design of a pump having a DC brushless motor and magnetic bearings with a potential for Naval applications in ships and submarines is discussed. In this paper attention is given to the selection of the magnetic bearings suitable for achieving active noise cancellation

  2. Research on RCP400-TB50 type reactor coolant pump shaft seal failure analysis and monitoring method

    Yuan Chaolian; Shen Yuxian; Wang Chuan; Du Pengcheng

    2014-01-01

    Mechanical seal is widely applied in mechanical devices of nuclear power plant. 3-stages mechanical seal applied in reactor coolant pump (abbreviate to RCP) is a kind of product with top technology and manufacture difficulty. As the only running machine in primary loop of nuclear power plant, RCP is designed with high security, reliability and perform ability. So performance of its key component, 3-stages mechanical seal, could directly decide whether units can operate safely and reliably. In this paper mechanical seal used in RCP400-TB50 type RCP which in designed and manufactured by Andritz AG is selected as a typical example of dynamic pressure type mechanical seal applied in second generation NPP. Its structure and working principle is expounded. Engineering fluid mechanics theory is used to establish the mathematical model using for analyzing status of mechanical seal and deducing the theoretical formula. Its correctness is verified by compare with the test data. So that research result can be used as the theoretical basis for analysis of RCP400-TB50 RCP shaft seal's working condition. According to the shaft seal operation characteristic we can establish a suitable RCP shaft seal monitoring method and interlock protection setting for NPP operation. (authors)

  3. Numerical Simulation of the Pressure Distribution in the Reactor Vessel Downcomer Region Fluctuated by the Reactor Coolant Pump

    Jeong, Dong Hwa; Jung, Byung Ryul; Jang, Ho Cheol; Yune, Seok Jeong; Kim, Eun Kee [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    In this study the numerical simulation of the pressure distribution in the downcomer region resulting from the pressure pulsation by the Reactor Coolant Pump (RCP) is performed using the Finite Difference Method (FDM). Simulation is carried out for the cylindrical shaped 2-dimensional model equivalent to the outer surface of the Core Support Barrel (CSB) of APR1400 and a 1/2 model is adopted based on the bilateral symmetry by the inlet nozzle. The fluid temperature is 555 .deg. F and the forcing frequencies are 120Hz, 240Hz, 360Hz and 480Hz. Simulation results of the axial pressure distributions are provided as the Root Mean Square (RMS) values at the five locations of 0°, 45°, 90°, 135° and 180° in the circumferential direction from the inlet nozzle location. In the study, the numerical simulation of pressure distributions in the downcomer region induced by the RCP was performed using FDM and the results were reviewed. The interference of the waves returned from both boundaries in the axial direction and the source of the sinusoidal wave is shown on the inlet nozzle interface pressure point. It seems that the maximum pressures result from the superposition of the waves reflected from the seating surface and the waves newly arrived from the inlet nozzle interface pressure location.

  4. Induction generator-induction motor wind-powered pumping system

    Miranda, M.S.; Lyra, R.O.C.; Silva, S.R. [CPDEE - UFMG, Belo Horizonte (Brazil)

    1997-12-31

    The energy storage matter plays an important role in wind-electric conversion systems for isolated applications. Having that in mind, two different approaches can be basically considered: either the immediate conversion of the generated electric energy, as in a water pumping system or electric energy storage for later use, as in a battery charging system. Due to some features such as no need of an external reactive power source and, sometimes, a gearbox, permanent-magnet synchronous generators have been broadly used in low rated power isolated systems. Despite that, system performance can be affected when the generator is feeding an inductive load (e.g., an induction motor) under variable-speed-variable-frequency operational conditions. Since there is no effective flux control, motor overload may occur at high wind speeds. Thus, good system performance can be obtained through additional control devices which may increase system cost. Although being rugged and cheap, induction machines always work as a reactive power drain; therefore, they demand an external reactive power source. Considering that, reactive static compensators appear as an attractive alternative to the cost x performance problem. In addition to that, different control strategies can be used so that system performance can be improved.

  5. The Performance Test for Reactor Coolant Pump (RCP) adopting Variable Restriction Orifice Type Control Valve

    Kim, S.; Bae, B. U.; Cho, Y. J. and others

    2014-05-15

    The design values of the RCPTF are 17.2 MPa, 343 .deg. C, 11.7 m{sup 3}/s, and 13 MW in the maximum pressure, temperature, flow rate, and electrical power, respectively. In the RCPTF, various types of tests can be performed including a hydraulic performance test to acquire a H-Q curve as well seal transient tests, thrust bearing transient test, cost down test, NPSHR verification test, and so on. After a commissioning startup test was successfully perfomed, mechanical structures are improved including a flow stabilizer and variable restriction orifice. Two- branch pipe (Y-branch) was installed to regulate the flow rate in the range of performance tests. In the main pipe, a flow restrictor (RO: Restriction Orifice) for limiting the maximum flow rate was installed. In the branch pipe line, a globe valve and a butterfly valves for regulating the flow rate was located on the each branch line. When the pressure loss of the valve side is smaller than that of the RO side, the flow rate of valve side was increasing and the flow disturbance was occurred in the lower pipe line. Due to flow disturbnace, it is to cause an error when measuring RCP head and flow measurement of the venturi flow meter installed in the lower main pipe line, and thus leading to a decrease in measurement accuracy as a result. To increase the efficiency of the flow control availability of the test facility, the variable restriction orifice (VRO) type flow control valve was designed and manufactured. In the RCPTF in KAERI, the performance tests and various kinds of transient tests of the RCP were successfully performed. In this study, H-Q curve of the pump using the VRO revealed a similar trend to the result from two ROs. The VRO was confirmed to effectively cover the full test range of the flow rate.

  6. On-line PWR RHR pump performance testing following motor and impeller replacement

    DiMarzo, J.T.

    1996-12-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  7. Present situation and the future task of pumps and motors for mobile application; Kensetsu kikai sharyoyo pump/motor no genjo to kadai

    Ishii, S.

    1994-09-15

    The current technologies of hydraulic pumps and motors for construction machinery and vehicles are discussed. A bent axis type axial double piston pump for negative feedback control systems and a cam plate type double piston pump superior in responsibility for load sensing control systems are usually used for hydraulic shovels. A split flow type double piston pump is on the increase for mini-shovels, having an unsolved problem as pressure fluctuation. The use of piston pumps is increasing rapidly for cranes, and a cam plate type axial piston pump is mainly used for large cranes. A traveling motor integrated with a transmission gear is usually used for hydraulic shovels, and a swing motor with valves and a brake in its casing is also used. A bent axis type variable displacement motor combined with a transmission is on the increase for winches, and a reliable preventing system from drop of suspended loading is expected to be developed rapidly. HST for traveling is also diffusing into small construction machinery. 12 figs.

  8. Expanding the applicable duration for shrink fitting of the ultrathin-walled reactor coolant pump rotor-can

    Li, Ruiqin; Zhang, Chi; Zhang, Liwen; Cui, Yan; Shen, Wenfei

    2017-01-01

    Highlights: •A thermal-mechanical coupled finite element model was developed to simulate the whole process. •Heat capacity added layer was used to extend the limited time for the process. •Shrink-fitted experiments were performed to verify the simulation results. -- Abstract: The rotor-can of reactor coolant pump (RCP) is generally assembled on the rotor using shrink fitting technique. The rotor-can is characterized by large height and ultrathin-walled cylinder, thus, its rigidity is weak and heat capacity is quite limited. The shrink fitting process has to be completed within a short limited-time, which makes it difficult for rotor to insert in the rotor-can completely. In order to solve this problem, a new method was proposed to extend the limited time by using a heat capacity added layer (HCAL) during the shrink fitting process. A thermal-mechanical coupled finite element (FE) model was developed to simulate the whole process. The transient heat exchange with a narrow gap between rotor and rotor-can during the shrink fitting process was taken into consideration. The limited time was predicted by calculating and analyzing the evolutions of temperature field and radial displacement field of the rotor-can. The simulation results indicate that the limited time of the shrink fitting process can be significantly extended with the increase of HCAL in thickness. Then, shrink fitting experiments were performed to confirm the extending effect of the HCAL. The experimental results of limited time show good agreement with the predicted values. The current results will certainly help the designer to improve the shrink fitting technique.

  9. Improvement in motor performance during high pressure pump starting at NDDP, Kalpakkam

    Nagaraj, R.; Murugan, V.; Thalor, K.L.; Saxena, A.K.; Dangore, A.Y.; Prabhakar, S.; Tiwari, P.K.

    2007-01-01

    The major energy requirement required for a Sea Water Reverse Osmosis is in the form of Electrical Energy. The primary energy requirement in the process is the electrical energy fed to High Pressure Pumps to pressurize the feed sea water to membranes. This High pressure pump being a high inertia load requires very high torque at the time of starting. This high starting torque requirement results in increased acceleration time of the motor which subsequently increases the strain on the upstream electrical system from motor feeder to transformer. Such starting characteristic necessitates provision of special starting scheme for the high pressure pump motors. Sea water reverse osmosis (SWRO) plant of Nuclear Desalination Demonstration Project (NDDP) was commissioned in October 2002 at Kalpakkam, India. This paper presents the experiences of problems faced due to the typical starting characteristics of High Pressure pumps and provision of series reactor type motor starter for the same. (author)

  10. Study of brushless fuel pump (improvement of pump and motor parts). 2nd Report. Blushless dendo fuel pump no kento. 2

    Mine, K; Takada, S; Tatematsu, M; Takeuchi, H [Aisan Industry Co. Ltd., Aichi (Japan)

    1992-10-01

    A methanol use electrically driven fuel pump was developed as reported in the present report. Mixed fuel of gasoline with alcohol can be handled by a brushless fuel pump which was proposed and improved as reported. The flow rate performance was heightened to 25g/sec by heightening in output power of motor, while the high temperature performance was 17% heightened against the conventional ratio of lowering in flow rate by heightening in vapor jet capacity. Against the corrosiveness of methanol, an in-tank type was applied to the pump, and all its electrically conductive and other mechanical parts were made to be both anti-corrosive and anti-abrasive. It is structurally of a two-stage series turbine type of non-volume form. A sensor method was applied to the motor by confining the miniaturized control circuit of brushless motor in the motor so that the transistor is controlled against the heightening in temperature. The motor is a three-phase half-wave driving motor. Also developed was a fuel supply system which is useful for the mixed fuel covering a range of 100% methanol through 100% gasoline. The present pump is dimensionally interchangeable with the conventional gasoline use one. Its operational life is more than 10000 hours. 3 refs., 17 figs., 1 tab.

  11. Pump

    Mole, C.J.

    1983-01-01

    An electromagnetic pump for circulating liquid -metal coolant through a nuclear reactor wherein opposite walls of a pump duct serve as electrodes to transmit current radially through the liquid-metal in the ducts. A circumferential electric field is supplied to the liquid-metal by a toroidal electromagnet which has core sections interposed between the ducts. The windings of the electromagnet are composed of metal which is superconductive at low temperatures and the electromagnet is maintained at a temperature at which it is superconductive by liquid helium which is fed through the conductors which supply the excitation for the electromagnet. The walls of the ducts joining the electrodes include metal plates insulated from the electrodes backed up by insulators so that they are capable of withstanding the pressure of the liquid-metal. These composite wall structures may also be of thin metal strips of low electrical conductivity backed up by sturdy insulators. (author)

  12. Analysis of canned-motor pump jamming failure in start hot standby mode

    Mo Zhengyu; Liu Hua; Wu Chao; Li Haifeng; Yu Zhongbin

    2009-01-01

    The reason of a canned-motor pump start failure is deduced in this paper. It is shown that the undersized gap of the lower guide bearing causes the start failure. By the stripping inspection data, the design value of the guide bearing gap, the cumulative departure of manufacturing assembly and the preloading of the bearing liner lead to the undersized gap. The test result shows that the improved canned-motor pump can start up successfully. (authors)

  13. Seismic stress analysis of feeder lines to LOFT primary coolant pump motors

    Kuehster, C.J.

    1978-01-01

    The conduit system in the LOFT Support Building was analyzed for seismic loading. The conduit itself plus its various supports were subjected to both horizontal and vertical forces. The results show the system loads or stresses to be within allowables

  14. On-line PWR RHR pump performance testing following motor and impeller replacement

    DiMarzo, J.T.

    1996-01-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump's 'B' impeller. The spare was installed into the 'B' train. The motor had never been run in the system before. A pump performance test was developed to verify it's operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the 'B' Train showed performance well in excess of the minimum required. The motor that was originally in the 'B' train was similarly overhauled and equipped with 'A' pump's original impeller, re-installed in the 'A' train, and tested. Analysis of the 'A' train results indicate that the RHR pump's performance was also well in excess of the vendors requirements

  15. Coolant clean-up and recycle systems

    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.)

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

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

    2011-01-01

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

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

    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

  18. Procedure to determine the optimal parameters of the main primary coolant pump after compacting the FRG-1 reactor. Pt. 2. Partial structures of the procedure

    Pihowicz, W.

    1999-01-01

    On the basis of an extensive physical and technical analysis the partial structures of the procedure had been developed. They represent a logical linkage of determination elements in the form of decision and result units. The developed partial structures enable to determine the physical parameters, which characterize the primary circuit together with the compact core as well as the main primary coolant pump coming into question after compacting the core. The report also contains a discussions and a comparison of the partial structures. (orig.) [de

  19. Modification of the motor control system of secondary cooling pumps of the multipurpose reactor - GA Siwabessy (RSG-GAS) to overcome the electrical flicking

    Yayan Andriyanto; Tufiq, M.; Adin Sudirman; Koes Indrakoesoema

    2012-01-01

    Modification of the motor control system of secondary cooling pumps that have been carried out to overcome the electrical flicking. Consequences of unanticipated flicking is the cooling pumps will shut down. Modification is done by replacing the control system PLN to Uninterruptible Power Supply (UPS), plus a component of Miniature Circuit Breaker (MCB) 220 Volt 2 Amp 2 Pole, 2 Auxiliary Contactor-type 3RH1131-1APOO with 1 NO auxiliary contact (normally Open) and two time-1BP30-3RP1505 type relay (off delay). Delay time to overcome the electrical flicking is set to 1 seconds and 2 seconds, taking into calculate the electrical flicking that occurred while in the setting of time limits, the predicted secondary cooling water pressure inside the pipeline is still eligible and the motor cooling system operation can still continue to operate. In the flicking of an electrical incident on 13 April 2010 at 0:45 Pm showed that the modifications of the secondary coolant pump motors operation when the reactor operates with a power of 15 MW. (author)

  20. Research on the induction motor current signature for centrifugal pump at cavitation condition

    Yin Luo

    2015-11-01

    Full Text Available Cavitation is a major undesirable phenomenon for centrifugal pump because it can cause hydraulic performance deterioration, pump damage by pitting and material erosion, and structural vibration and noise. Cavitation can appear within the entire range of the operating conditions; therefore, it must be prevented by all means. Sensorless monitoring technology based on motor current signature analysis is non-intrusive and economic for monitoring motor-driven equipment. Thus, this technology is suitable for centrifugal pump systems. The motor current signature for centrifugal pump load at the cavitation condition is the basis of this technology. However, systematic research is lacking on sensorless monitoring technology based on motor current signature. As a result, the tentative exploration for motor current signature at cavitation load was conducted in this study. The results show that the stator current is still a sinusoidal alternating current strictly to the law of sine. Moreover, the root mean square of the current fluctuates because of different flow regimes in the cavitation progress and decreases because vapor density is smaller than water density when cavitation is fully formed. For the stator current spectrum, the noise level, noise distribution, rotation speed, and vane pass frequency components show features in the cavitation process. These indicator indexes change according to the stage of cavitation development. Thus, the motor current signature analysis is found to be a feasible and cost-effective method for the stages of cavitation condition.

  1. Sensorless Sinusoidal Drives for Fan and Pump Motors by V/f Control

    Kiuchi, Mitsuyuki; Ohnishi, Tokuo

    This paper proposes sensorless sinusoidal driving methods of permanent magnet synchronous motors for fans and pumps by V/f control. The proposed methods are simple methods that control the motor peak current constant by voltage or frequency control, and are characterized by DC link current detection using a single shunt resistor at carrier wave signal bottom timing. As a result of the dumping factor from square torque load characteristics of fan and pump motors, it is possible to control stable starting and stable steady state by V/f control. In general, pressure losses as a result of the fluid pass of fan and pump systems are nearly constant; therefore, the flow rate and motor torque are determined by revolutions. Accordingly, high efficiency driving is possible by setting corresponding currents to q-axis currents (torque currents) at target revolutions. Because of the simple current detection and motor control methods, the proposed methods are optimum for fan and pump motor driving systems of home appliances.

  2. Assessment of load of beam-balanced pumping units by electric motor power indicators

    Д. И. Шишлянников

    2017-10-01

    Full Text Available The results of experimental studies on the loading of beam-balanced pumping units (BP of sucker rod- pumping equipment (SRPE are presented. It is noted that the key factor that has the most significant effect causing the SRPE failure is the balance of the beam pumping unit, which determines the amount of specific energy consumption for the rise of reservoir fluid and the level of dynamic loads on the machine units. The urgency of using software-recording systems for estimating the loading of units of oil field pumping installations is substantiated. The principle of operation and design of the «AKD-SK» software recording system is described. The prospects of using this method for controlling the performance parameters and evaluating the technical state of the sicker rod-pumping units is proved on the basis of an analysis of the magnitude and nature of the changes in the loads of drive motors determined by the registration of the instantaneous values of the consumed power. The main provisions of the methodology for analyzing the watt-meters of drive motors of the sucker rod-pumping units are outlined. The nature of the manifestation of the main defects of submersible pumps and beam-balanced pumping units is described. The results of pilot-industrial tests of the beam-balanced pumping units equipped with advanced permanent magnet motors and intelligent control stations are presented. It is proved that the use of permanent magnet motors allows to reduce the specific energy consumption for the rise of reservoir fluid, which increases the efficiency of the SRPE.However, the presence of transient processes and generator operating modes of the permanent magnet motors results in the occurrence of significant dynamic loads, which, due to the rigid fixing of the rotor of magnet motor on the reducer shaft, negatively affect the life of the gearbox bearings. It has been shown that the lack of its own bearings in the tested motors causes a high probability

  3. Primary system hydraulic characteristics after modification of reactor coolant pumps' impeller wheels at Bohunice NPP executed in 2012 and 2013

    Hermansky, Jozef; Zavodsky, Martin

    2014-01-01

    A coolant flow through the reactor is usually determined after annual outages at Slovak NPP (VVER 440) in two distinct ways. First method is determination on the basis of the secondary system parameters - measurement of thermal balances. The value achieved by this method is used as the input parameter in the Table of allowed reactor operation modes. The second method draws from the primary system parameters - measurement of primary system hydraulic characteristics. Flow nozzles used for the measurement of feed water flow behind high pressure heaters were replaced at both Bohunice Units during outages in 2008. The feed water flow behind high pressure heaters is one of the main parameters used for the determination of coolant flow through the reactor by the first method. Compared to the measurement executed during previous fuel cycles, the calculated coolant flow through the reactor decreased considerably after the change of flow nozzles. The imaginary change of coolant flow through the reactor at Unit 3 was -1,6 %; and at Unit 4 -2,6 %. This change was not proved by the parallel measurement of primary system hydraulic characteristics. Later it was found out that the original flow nozzles used for 25 years were substantially deposited (original inner diameter of the nozzles was reduced by about 0,6-0,9 mm). Therefore feed water flow measurement was untrustworthy within the recent years. On the findings stated above, Bohunice NPP has decided to increase coolant flow through the reactor by changing the reactor coolant pumps impeller wheels. The modification of impellers wheels is planned within years 2012 to 2014. During the outages in 2013 two impeller wheels were replaced at both units. Nowadays Unit 4 is operated with all 6 new impeller wheels and Unit 3 with four new impeller wheels. Modification of last two impeller wheels at Unit 3 will be performed during the outage in 2014. On account of impeller wheels modification, non-standard measurement of PS hydraulic

  4. A Peristaltic Micro Pump Driven by a Rotating Motor with Magnetically Attracted Steel Balls

    Zhaoying Zhou

    2009-04-01

    Full Text Available In this paper, we present a membrane peristaltic micro pump driven by a rotating motor with magnetically attracted steel balls for lab-on-a-chip applications. The fabrication process is based on standard soft lithography technology and bonding of a PDMS layer with a PMMA substrate. A linear flow rate range ~490 μL/min was obtained by simply varying the rotation speed of a DC motor, and a maximum back pressure of 592 Pa was achieved at a rotation speed of 43 rpm. The flow rate of the pump can also be adjusted by using steel balls with different diameters or changing the number of balls. Nevertheless, the micro pump can also work in high speed mode. A high back pressure up to 10 kPa was achieved at 500 rpm using a high speed DC motor, and an utmost flow rate up to 5 mL/min was reached.

  5. Canned motor pumps at Heavy Water Project, Baroda [Paper No.: II-2

    Batra, R.K.; Waishampayan, S.G.

    1981-01-01

    Heavy Water Project, Baroda is having special canned motor pumps for pumping ammonia and potassium amide. These pumps work under a pressure of 640 kg/cm 2 and are lubricated and cooled by the same fluid. These pumps are having special bearings consisting of mating surfaces of tungsten carbine Vs ceramic or stellite Vs carbon depending upon application. The total number of such pumps installed in the plant is around 52. These pumps being installed in high pressure vessels working at a pressure of 640 kg/cm 2 have special maintenance problems and need special care during maintenance. Pumps once boxed up are completely out of reach for immediate maintenance if needed. The failure of these pumps may mean a down time of one to two months for the plant. Besides above, there are multistage barrel type ammonia pumps with 24 stages developing a pressure of 140 kg/cm 2 . All these equipments need special maintenance methods as the problems faced are varied and difficult to solve. This paper deals with general and unique type of problems faced on these pumps. (author)

  6. Computer model verification for seismic analysis of vertical pumps and motors

    McDonald, C.K.

    1993-01-01

    The general principles of modeling vertical pumps and motors are discussed and then two examples of verifying the models are presented in detail. The first examples is a vertical pump and motor assembly. The model and computer analysis are presented and the first four modes (frequencies) calculated are compared to the values of the same modes obtained from a shaker test. The model used for this example is a lumped mass connected by massless beams model. The shaker test was performed by National Technical Services, Los Angeles, CA. The second example is a larger vertical motor. The model used for this example is a finite element three dimensional shell model. The first frequency obtained from this model is compared to the first frequency obtained from shop tests for several different motors. The shop tests were performed by Reliance Electric, Stratford, Ontario and Siemens-Allis, Inc., Norwood, Ohio

  7. Comparative Study on Photovoltaic Pumping Systems Driven by Different Motors Optimized with Sliding Mode Control

    Abdelhak Bouchakour

    2017-06-01

    Full Text Available This study investigates the performance of three different photovoltaic (PV water pumping systems driven by three types of motors, namely a separately excited DC motor (DCM, an asynchronous motor (ASM, and a permanent magnet synchronous motor (PMSM, via a DC/DC buck-boost converter coupled to a centrifugal pump. The purpose of this study is to implement a fast and robust control for this type of a nonlinear system, controlled by sliding mode (SM. This paper presents an SM control technique for controlling a DC/DC buck-boost converter to transfer the maximum power delivered by the PV generator. Each component is studied and analyzed to simulate the global system in MATLAB/SIMULINK. The three systems are then compared to determine the overall effectiveness of the proposed command. The study concludes that the ASM-driven PV system yields highly favorable results and requires less maintenance compared with other systems.

  8. What we learn from surveillance testing of standby turbine driven and motor driven pumps

    Christie, B.

    1996-12-01

    This paper describes a comparison of the performance information collected by the author and the respective system engineers from five standby turbine driven pumps at four commercial nuclear electric generating units in the United States and from two standby motor driven pumps at two of these generating units. Information was collected from surveillance testing and from Non-Test actuations. Most of the performance information (97%) came from surveillance testing. {open_quotes}Conditional Probabilities{close_quotes} of the pumps ability to respond to a random demand were calculated for each of the seven standby pumps and compared to the historical record of the Non-Test actuations. It appears that the Conditional Probabilities are comparable to the rate of success for Non-Test actuations. The Conditional Probabilities of the standby motor driven pumps (approximately 99%) are better than the Conditional Probabilities of the standby turbine driven pumps (82%-96% range). Recommendations were made to improve the Conditional Probabilities of the standby turbine driven pumps.

  9. What we learn from surveillance testing of standby turbine driven and motor driven pumps

    Christie, B.

    1996-01-01

    This paper describes a comparison of the performance information collected by the author and the respective system engineers from five standby turbine driven pumps at four commercial nuclear electric generating units in the United States and from two standby motor driven pumps at two of these generating units. Information was collected from surveillance testing and from Non-Test actuations. Most of the performance information (97%) came from surveillance testing. open-quotes Conditional Probabilitiesclose quotes of the pumps ability to respond to a random demand were calculated for each of the seven standby pumps and compared to the historical record of the Non-Test actuations. It appears that the Conditional Probabilities are comparable to the rate of success for Non-Test actuations. The Conditional Probabilities of the standby motor driven pumps (approximately 99%) are better than the Conditional Probabilities of the standby turbine driven pumps (82%-96% range). Recommendations were made to improve the Conditional Probabilities of the standby turbine driven pumps

  10. Compartmentalized safety coolant injection system

    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)

  11. A superconducting linear motor drive for a positive displacement bellows pump for use in the g-2 cryogenics system

    Green, M.A.

    1994-10-01

    Forced two-phase cooling of indirectly cooled magnets requires circulation of liquid helium through the magnet cooling channel. A bellows helium pump is one possible way of providing helium flow to a magnet cooling system. Since the bellows type of helium pump is immersed in liquid helium, a superconducting linear motor drive appears to be an attractive option. This report describes a linear motor drive that employs oriented permanent magnet materials such as samarium-cobalt as the stator magnet system and a superconducting loud speaker voice coil type of drive as the armature of the linear motor. This report examines drive motor requirements for a helium pump

  12. 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

    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.

  13. 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

    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.

  14. Adiabatic and non-adiabatic charge pumping in a single-level molecular motor

    Napitu, B.D.; Thijssen, J.M.

    2015-01-01

    We propose a design for realizing quantum charge pump based on a recent proposal for a molecular motor (Seldenthuis J S et al 2010 ACS Nano 4 6681). Our design is based on the presence of a moiety with a permanent dipole moment which can rotate, thereby modulating the couplings to metallic contacts

  15. INTENSITY SETTER FOR A DEVICE OF SMOOTH START OF SUBMERSIBLE PUMP ELECTRIC MOTOR

    V.I. Lobov

    2016-06-01

    Full Text Available Purpose. Development of an intensity setter, which in a rational law changes the opening thyristor the voltage regulator and effectively to changing power supply voltage stator windings of the electric pump deepening, ensuring a smooth start in a wide range. Methodology. Electric submersible pump belongs to the small inertia electric, since it is not significant total moment of inertia, not exceeding two moments of inertia of the motor and static moment on the shaft does not exceed forty percent of the nominal torque. For technical requirements that electric acceleration time should have no less than twenty seconds or more. Office starting modes of electric submersible pumps economically justified using thyristor voltage regulator by forming the dial changes the intensity of the necessary legislation in time voltage feeding the stator windings. This ensures a smooth start right rotor of the electric submersible pump. Results. A block diagram of the intensity setter that is: with control unit, two units that form the exponential voltage supply emitter follower and regulatory elements. The mathematical expressions for voltage at the stator windings of the motor, changing exponentially, opening the angle of thyristor power unit thyristor the voltage regulator, which is determined through the initial angle of opening. Provided formula for pick-up voltage and minimum voltage, time constants, which are determined from the basic equations of motion and mechanical characteristics of the electric motor. Analytical investigated by the voltage dependence violation by changing the time constant flowing and growing exhibitor supply voltage stator electric circuit deepening pump. Originality. Proposed in the initial time on the stator windings of the electric pick-up voltage is applied. Under the influence of this voltage, motor rotor begins to accelerate. At the same time, pick-up voltage decreases the minimum startup voltage varies exponentially in which

  16. The Performance Evaluation of Overall Heat Transfer and Pumping Power of γ-Al2O3/water Nanofluid as Coolant in Automotive Diesel Engine Radiator

    Navid Bozorgan

    2013-05-01

    Full Text Available The efficiency of γ-Al2O3/water nanofluid as coolant is investigated in the present study. γ-Al2O3 nanoparticles with diameters of 20 nm dispersed in water with volume concentrations up 2% are selected and their performance in a radiator of Chevrolet Suburban diesel engine under turbulent flow conditions are numerically studied. The performance of an automobile radiator is a function of overall heat transfer coefficient and total heat transfer area. The heat transfer relations between nanofluid and airflow have been investigated to evaluate the overall heat transfer and the pumping power of γ-Al2O3/water nanofluid in the radiator with a given heat exchange capacity. In the present paper, the effects of the automotive speed and Reynolds number of the nanofluid in the different volume concentrations on the radiator performance are also investigated. As an example, the results show that for 2% γ-Al2O3 nanoparticles in water with Renf=6000 in the radiator while the automotive speed is 50 mph, the overall heat transfer coefficient and pumping power are approximately 11.11% and 29.17% more than that of water for given conditions, respectively. These results confirm that γ-Al2O3/water nanofluid offers higher overall heat transfer performance than water and can be reduced the total heat transfer area of the radiator.

  17. The Best Efficiency Point of the Performance of Solar Cell Panel System for Pumping Water at Various Lifting Heads Using 100 W Motor Pump Unit

    Himran, Sukri

    2013-01-01

    This study was carried out experimentally and analytically about the performance of solar cell panel system for operating the pump coupled by dc motor. The solar cell panel with total area 1.9848 m2 consists of three modules of 80 Wp each. The small centrifugal pump powered by dc motor is operated to lift water from 1m to 7m heads in sequence and gives the amount of water pumped over the whole day from 08.00 to 16.00 h are 11988, 10851, 8874, 7695, 5760, 3600...

  18. The Best Efficiency Point of the Performance of Solar Cell Panel System for Pumping Water at Various Lifting Heads Using 100 W Motor- Pump Unit

    S. Himran; B. Mire; N. Salam; L. Sule

    2013-01-01

    This study was carried out experimentally and analytically about the performance of solar cell panel system for operating the pump coupled by dc-motor. The solar cell panel with total area 1.9848 m2 consists of three modules of 80 Wp each. The small centrifugal pump powered by dc-motor is operated to lift water from 1m to 7m heads in sequence and gives the amount of water pumped over the whole day from 08.00 to 16.00 h are 11988, 10851, 8874, 7695, 5760, 3600, 2340 L/d respectively. The hourl...

  19. Vectorial Command of Induction Motor Pumping System Supplied by a Photovoltaic Generator

    Makhlouf, Messaoud; Messai, Feyrouz; Benalla, Hocine

    2011-01-01

    With the continuous decrease of the cost of solar cells, there is an increasing interest and needs in photovoltaic (PV) system applications following standard of living improvements. Water pumping system powered by solar-cell generators are one of the most important applications. The fluctuation of solar energy on one hand, and the necessity to optimise available solar energy on the other, it is useful to develop new efficient and flexible modes to control motors that entrain the pump. A vectorial control of an asynchronous motor fed by a photovoltaic system is proposed. This paper investigates a photovoltaic-electro mechanic chain, composed of a PV generator, DC-AC converter, a vector controlled induction motor and centrifugal pump. The PV generator is forced to operate at its maximum power point by using an appropriate search algorithm integrated in the vector control. The optimization is realized without need to adding a DC-DC converter to the chain. The motor supply is also ensured in all insolation conditions. Simulation results show the effectiveness and feasibility of such an approach.

  20. Experience of decontaminating of the ventilation system electric motors and main circulation and waterside pump drive electric motors in conditions of the Chernobyl' NPP

    Denisov, V.I.; Karataev, B.A.; Platonov, A.I.; Potekhin, V.B.; Ryumin, G.V.; Sorokin, N.M.; Fenogenov, V.A.; Cheremshin, P.I.

    1989-01-01

    The ventilation system electric motor decontamination is described. In decontaminating waterside pump rotors and stators with high level of contamination was developed a procedure, eliminating corrosion action and damage of electric isolation coatings

  1. Residual heat removal pump retrofit program

    Dudiak, J.G.; McKenna, J.M.

    1990-01-01

    Residual Heat Removal (RHR) pumps installed in pressurized water reactor power plants are used to provide the removal of decay heat from the reactor and to provide low head safety injection in the event of loss of coolant in the reactor coolant system. These pumps are subjected to rather severe temperature and pressure transients, therefore, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. RHR pumps have traditionally been a significant maintenance item for many utilities. The close-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. The casing separation requires the loosening of numerous highly torqued studs. Once the casing is separated, the impeller is dropped from the motor shaft to allow removal of the mechanical seal and casing cover from the motor shaft. Galling of the impeller to the motor shaft is not uncommon. The RHR pump internals are radioactive and the separation of the pump casing to perform routine maintenance exposes the maintenance personnel to high radiation levels. The handling of the impeller also exposes the maintenance personnel to high radiation levels. This paper introduces a design modification developed to convert the close-coupled RHR pumps to a coupled configuration

  2. Experimental investigation of material chemical effects on emergency core cooling pump suction filter performance after loss of coolant accident

    Park, Jong Woon; Park, Byung Gi; Kim, Chang Hyun

    2009-01-01

    Integral tests of head loss through an emergency core cooling filter screen are conducted, simulating reactor building environmental conditions for 30 days after a loss of coolant accident. A test rig with five individual loops each of whose chamber is established to test chemical product formation and measure the head loss through a sample filter. The screen area at each chamber and the amounts of reactor building materials are scaled down according to specific plant condition. A series of tests have been performed to investigate the effects of calcium-silicate, reactor building spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the filter screen is strongly affected by spray duration and the head loss increase is rapid at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKON TM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

  3. Sensitivity Analysis of Core Damage from Reactor Coolant Pump Seal Leakage during Extended Loss of All AC Power

    Park, Da Hee; Kim, Min Gi; Lee, Kyung Jin; Hwang, Su hyun; Lee, Byung Chul [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Yoon, Duk Joo; Lee, Seung Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-10-15

    In this study, in order to comprehend the Fukushima accident, the sensitivity analysis was performed to analyze the behavior of Reactor Coolant System (RCS) during ELAP using the RELAP5/MOD3.3 code. The Fukushima accident was caused by tsunami resulted in Station Black Out (SBO) followed by the reactor core melt-down and release of radioactive materials. After the accident, the equipment and strategies for the Extended Loss of All AC Power (ELAP) were recommended strongly. In this analysis, sensitivity studies for the RCP seal failure of the OPR1000 type NPP were performed by using RELAP5/MOD3.3 code. Six cases with different leakage rate of RCP seal were studied for ELAP with operator action or not. The main findings are summarized as follows: (1) Without the operator action, the core uncovery time is determined by the leakage rate of RCP seal. When the leakage rate per RCP seal are 5 gpm, 50 gpm, and 300 gpm respectively, the core uncovery time are 1.62 hr, 1.58 hr, and 1.29 hr respectively. Namely, If the leakage rate of RCP seal was much bigger, the uncover time of core would be shorter. (2) In case that the cooling by SG secondary side was performed using the TDAFP and SG ADV, the core uncovery time was significantly extended.

  4. Pump Coupling & Motor bearing damage detection using Condition Monitoring at DTPS

    Bari, H. M.; Deshpande, A. A.; Jalkote, P. S.; Patil, S. S.

    2012-05-01

    This paper shares a success story out of the implementation of Co-ordinated Condition Monitoring techniques at DTPS, wherein imminent Mis-alignment of HT auxiliary BFP - 1B and Motor bearing failure of ID FAN - 1B was diagnosed. On 30/12/2010, Booster Pump DE horizontal reading increased from 4.8 to 5.1 and then upto 5.9 mm/sec. It was suspected that Booster pump was mis-aligned with Motor. To confirm misalignment, Phase Analysis was also done which showed that Coupling phase difference was 180 Degrees. Vibration & Phase Analysis helped in diagnosing the exact root cause of abnormity in advance, saving plant from huge losses which could have caused total cost of £ 104,071. On 06/01/2011, ID fan 1B Motor NDE & DE horizontal vibration readings deviated from 0.5 to 0.8 and 0.6 to 0.8 mm/sec (RMS) respectively. Noise level increased from 99.1 to 101.9 db. It was suspected that Motor bearings had loosened over the shaft. Meanwhile, after opening of Motor, Inner race of NDE side was found cracked and loosened over the shaft. Vibration Analysis & Noise Monitoring helped in diagnosing the exact root cause of abnormity in advance, saving plant from huge losses which could have caused total cost of £ 308,857.

  5. Failure analysis of motor bearing of sea water pump in nuclear power plant

    Bian Chunhua; Zhang Wei

    2015-01-01

    The motor bearing of sea water pump in Qinshan Phase II Nuclear Power plant broke after only one year's using. This paper introduces failure analysis process of the motor bearing. Chemical composition analysis, metallic phase analysis, micrographic examination, and hardness analysis, dimension analysis of each part of the bearing, as well as the high temperature and low temperature performance analysis of lubricating grease are performed. According to the analysis above mentioned, the failure mode of the bearing is wearing, and the reason of wearing is inappropriate installation of the bearing. (authors)

  6. Method for Lumped Parameter simulation of Digital Displacement pumps/motors based on CFD

    Rømer, Daniel; Johansen, Per; Pedersen, Henrik C.

    2013-01-01

    Digital displacement fluid power pumps/motors offers improved efficiency and performance compared to traditional variable displacement pump/motors. These improvements are made possible by using efficient electronically controlled seat valves and careful design of the flow geometry. To optimize...... the design and control of digital displacement machines, there is a need for simulation models, preferably models with low computational cost. Therefore, a low computational cost generic lumped parameter model of digital displacement machine is presented, including a method for determining the needed model...... parameters based on steady CFD results, in order to take detailed geometry information into account. The response of the lumped parameter model is compared to a computational expensive transient CFD model for an example geometry....

  7. SLIPPER PERFORMANCE INVESTIGATION IN AXIAL PISTON PUMPS AND MOTORS-FLOW AND VISCOUS POWER LOSSES

    A. Osman KURBAN

    1997-01-01

    Full Text Available In this study, the slippers being the most effective on the performance of swash plate type axial piston pumps and motors, which is a good example of hydrodynamic-hydrostatic bearing applications, have been investigated. With respect to this, having derived the viscous moment loss, viscous flow leakage loss and power loss equations, the variations of these parameters under different operating conditions have been examined experimentally.

  8. Assessment of the integrity and functional requirement of moderator pump-motor units

    Soni, R.S.; Chawla, D.S.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1995-01-01

    The design of various active components in a nuclear power plant calls for a satisfactory analysis of these components for various loadings from the point of view of safety because a designated number of these components must always remain functional. Presented herein is the structural and seismic qualification of one the active components namely the moderator system pump-motor units for a typical PHWR. (author). 5 refs., 8 figs

  9. TRAC analysis of an 80% pump-side, cold-leg, large-break loss-of-coolant accident for the Westinghouse AP600 advanced reactor design

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

    1996-01-01

    An updated TRAC 80% pump-side, cold-leg, large-break (LB) loss-of-coolant accident (LOCA) has been calculated for the Westinghouse AP600 advanced reactor design. The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The break size and location were calculated by Westinghouse to be the most severe LBLOCA for the AP600 design. The LBLOCA transient was calculated to 280 s, which is the time of in-containment refueling water-storage-tank injection. All fuel rods were quenched completely by 240 s. Peak cladding temperatures (PCTs) were well below the licensing limit of 1,478 K (2,200 F) but were very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCTs for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown and refill periods. The reasons for these differences are still being investigated

  10. Enhanced Component Performance Study: Motor-Driven Pumps 1998-2014

    Schroeder, John Alton

    2016-01-01

    This report presents an enhanced performance evaluation of motor-driven pumps at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from fiscal year 1998 through 2014 for the component reliability as reported in the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The motor-driven pump failure modes considered for standby systems are failure to start, failure to run less than or equal to one hour, and failure to run more than one hour; for normally running systems, the failure modes considered are failure to start and failure to run. An eight hour unreliability estimate is also calculated and trended. The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire active period. Statistically significant increasing trends were identified in pump run hours per reactor year. Statistically significant decreasing trends were identified for standby systems industry-wide frequency of start demands, and run hours per reactor year for runs of less than or equal to one hour.

  11. Enhanced Component Performance Study: Motor-Driven Pumps 1998–2014

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    This report presents an enhanced performance evaluation of motor-driven pumps at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from fiscal year 1998 through 2014 for the component reliability as reported in the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The motor-driven pump failure modes considered for standby systems are failure to start, failure to run less than or equal to one hour, and failure to run more than one hour; for normally running systems, the failure modes considered are failure to start and failure to run. An eight hour unreliability estimate is also calculated and trended. The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire active period. Statistically significant increasing trends were identified in pump run hours per reactor year. Statistically significant decreasing trends were identified for standby systems industry-wide frequency of start demands, and run hours per reactor year for runs of less than or equal to one hour.

  12. High-inertia drive motors and their starting characteristics

    Anon.

    1980-01-01

    The motor for a large reactor coolant pump failed while starting. The motor-application and the motor-failure are discussed in detail. A review of applications of motors for high-inertia drives shows that a motor designed and built to today's industry-standards might be overstressed while experiencing abnormal starting conditions, even though its protection is in accord with accepted practice. The inter-relationship between motor characteristics and characteristics of various types of protection are discussed, briefly. The review concludes that motor specifications and motor standards should be augmented. 1 ref

  13. Optimization of geometry of annular seat valves suitable for Digital Displacement fluid power pumps/motors

    Rømer, Daniel; Johansen, Per; Pedersen, Henrik C.

    2013-01-01

    Digital Displacement Fluid Power is an upcoming technology setting new standards for the achievable efficiency of fluid power pumps and motors. The core element of the Digital Displacement technology is high performance electronically controlled seat valves, which must exhibit very low flow...... work an annular seat valve suitable for use in Digital Displacement units is considered, and the ring geometry is optimized using finite element analysis including non-linear material behaviour, contact elements and fluid pressure penetrating load, closely reflecting the actual load of the seat valve...

  14. Dynamic response characteristics evaluation of hydrostatic bearing in hydraulic piston pump/motor

    Ham, Young Bog; Yun, So Nam; Kim, Dong Soo; Choi, Byoung Oh; Kim, Sung Dong

    2001-01-01

    In swash plate type axial piston hydraulic pump and motor, the piston shoe is periodically pressurized with square function shape by supply pressure load as rotation of cylinder barrel. Therefore the recess pressure ono bottom part of piston shoe is suddenly increase through orifice in the piston shoe. In this study, we simulated that the frequency response of the recess pressure against with change of supply pressure with analysis tool. Also, we evaluate the dynamic response characteristics of overbalanced hydrostatic bearing with change of the orifice diameter

  15. Residual heat removal pump and low pressure safety injection pump retrofit program

    Dudiak, J.G.; McKenna, J.M.

    1992-01-01

    Residual Heat Removal (RHR) and low pressure safety injection (LPSI) pumps installed in pressurized water-to-reactor power plants are used to provide low-head safety injection in the event of loss of coolant in the reactor coolant system. Because these pumps are subjected to rather severe temperature and pressure transients, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. Typically the pump impeller is mounted on an extended motor shaft (close-coupled configuration) and a mechanical seal is employed at the pump end of the shaft. Traditionally RHR and LPSI pumps have been a significant maintenance item for many utilities. Periodic mechanical seal of motor bearing replacement often is considered routine maintenance. The closed-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. This paper introduces a design modification developed to convert the close-coupled RHR and LPSI pumps to a coupled configuration

  16. LOFT pump speed controller stability and accuracy analysis

    Good, R.R.

    1978-01-01

    Two system modifications to the primary coolant pumps motor generators control systems have recently been completed. The range of pump speed operation has been extended and the scoop tube positioner motor replaced. This has necessitated a re-analysis of PSMG stability throughout its range of operation. System accuracy requirements of less than 4 Hz differential pump speed when operating at less than 35 Hz and 8.5 Hz differential pump speed when operating at greater than 35 Hz can be guaranteed by specifying the gain of the system. The installation of the new scoop tube positioner motor will increase the PSMG system's bandwidth and stability. Low speed pump trips should be carefully evaluated if the pump's operational range is to extend to 10 Hz

  17. A Methodology for Evaluation of Inservice Test Intervals for Pumps and Motor-Operated Valves

    Cox, D.F.; Haynes, H.D.; McElhaney, K.L.; Otaduy, P.J.; Staunton, R.H.; Vesely, W.E.

    1999-01-01

    Recent nuclear industry reevaluation of component inservice testing (IST) requirements is resulting in requests for IST interval extensions and changes to traditional IST programs. To evaluate these requests, long-term component performance and the methods for mitigating degradation need to be understood. Determining the appropriate IST intervals, along with component testing, monitoring, trending, and maintenance effects, has become necessary. This study provides guidelines to support the evaluation of IST intervals for pumps and motor-operated valves (MOVs). It presents specific engineering information pertinent to the performance and monitoring/testing of pumps and MOVs, provides an analytical methodology for assessing the bounding effects of aging on component margin behavior, and identifies basic elements of an overall program to help ensure component operability. Guidance for assessing probabilistic methods and the risk importance and safety consequences of the performance of pumps and MOVs has not been specifically included within the scope of this report, but these elements may be included in licensee change requests

  18. Advanced development and operating experience with a canned motor pump under pressurized water reactor conditions

    Dittmer, H.; Reymann, A.; Seibig, B.; Reinecker, H.

    1988-01-01

    At the research reactor FRG-2, Geesthacht, an irradiation device is in operation for testing defective light-water-reactor (LWR) test fuel rods under pressurized water reactor conditions (320 0 C, 160 bar). The requirements to the canned motor pump for cooling water circulation: medium: Demineralized water, operating temperature 320 0 C, operating pressure 155 bar, radiation field of the reactor, integration in the irradiation capsule, helium leak rate -6 mbar.dm 3 .s -1 , minimum working life 3000 hours, were high and caused difficulties in the acquisition of this component. First test runs with supplied pumps showed that the desired working life could not be achieved. The results of the development steps, the test runs, and the performance in service show that for our range of applications, the best combination of materials for the radial bearings is silicon-infiltrated SiC (8% free Si) against the same material. These bearings allowed a good working life for the pump to be achieved. (orig./GL) [de

  19. LMFBR with booster pump in pumping loop

    Rubinstein, H.J.

    1975-01-01

    A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation

  20. Failure analysis of the stud bolt of a canned motor pump of heavy water plant, Talcher

    Kumar, Sunil; Sethumadhavan, V.; Sah, D.N.; Sivaramakrishnan, K.S.; Kain, Vivekanand; Gadiyar, H.S.

    1990-01-01

    Detailed investigations have been carried out on the failed stud bolts (made of DIN 1.4021 X 20Cr 13) of a canned motor pump of Heavy Water Plant, Talcher, using metallographic, microhardness testing, scanning electron microscopy and electron probe micro-analysis(EPMA) techniques. The studs had failed in a brittle manner in the mid-length portion. The origin of the fracture has been identified to be a corroded region on the stud surface. Branching cracks propagating through silicon rich inclusions have been noted. Two types of inclusions, one containing Mn and S and other containing S, Mn and O 2 have been found in the material. Clusters of large and small inclusions of the above types have been found near the origin of the fracture. It has been concluded that the fracture was caused by corrosion fatigue, initiating at the cluster of inclusions present on the surface of the stud bolts. (author). 7 refs., 5 figs

  1. Apparatus for facilitating the servicing of inverted canned pump motors having limited access space and restricted access time especially in nuclear power plants

    Matusz, J.M.

    1991-01-01

    This patent describes a service apparatus for one or more inverted canned motor pumps installed above a floor and beneath a steam generator in a nuclear or fossil power plant with limited access space and limited access time at least in the case of nuclear power plants, each of the canned motor pumps having a pump casing and a depending motor having a flange secured to a pump casing flange by tensioned studs with tightened nuts. It comprises a maintenance cart having a height greater than the height of the motor beneath the motor flange and further having a generally U-shaped frame means with an open vertical side that permits the cart to be moved horizontally such that the cart frame means can be moved under the pump casing to surround the depending motor; actuator means supported by the cart frame means and having translating arm means engageable with support means on the motor; means for operating the translating arm means to support, raise and lower the motor; means supported by the frame means to support the motor flange prior to raising the motor to its installed position and after the motor has been released from its installed position and lowered to the cart; work platform means provided on the cart frame means at an elevation beneath the motor flange elevation; and roller means provided on the bottom of the cart frame means to facilitate horizontal cart movement along the floor

  2. Installation design of pump motor control systems for supplied of the RSG-GAS secondary raw water cooling system

    Kiswanto; Teguh Sulistyo; M-Taufik

    2013-01-01

    It has designed already of an installation of the pump motor control system for supplied of raw water to fulfil the RSG-GAS secondary cooling system. The installation design of this plant is used to supply electrical energy from PLN and 3 phase generator to operate the pump motor embedded multilevel type, capacity, Q = 30 m 3 /h; electric power, PN = 4 kW; voltage, 380V/3-/50Hz, and Y connections that can be operated manually or automatically by using the automatic transfer switch. The results obtained recapitulation total load of 4 kW, the magnitude of the nominal current of 9.5 A; kind of safety and capacity are used NFB 16 A, use of this type of cable to the control panel is PLN NYY 6 mm 2 diameter maximum current capacity of 25 A cable and use the control panel to the pump motor cable type NYY 4 mm 2 diameter maximum current capacity of 20 A. The design of the pump motor control system installation is ready to be implemented. (author)

  3. Electromagnetic pump technology

    Prabhakar, R.

    1994-01-01

    Fast Breeder Reactors have an important role to play in our nuclear power programme. Liquid metal sodium is used as the coolant for removing fission heat generated in fast reactors and a distinctive physical property of sodium is its high electrical conductivity. This enables application of electromagnetic (EM) pumps, working on same principle as electric motors, for pumping liquid sodium. Due to its lower efficiency as compared to centrifugal pumps, use of EM pumps has been restricted to reactor auxiliary circuits and experimental facilities. As part of our efforts to manufacture fast reactor components indigenously, work on the development of two types of EM pumps namely flat linear induction pump (FLIP) and annular linear induction pump (ALIP) has been undertaken. Design procedures based on an equivalent circuit approach have been established and results from testing a 5.6 x 10E-3 Cum/s (20 Cum/h) FLIP in a sodium loop were used to validate the procedure. (author). 7 refs., 6 figs

  4. An experimental and theoretical investigation on the effects of adding hybrid nanoparticles on heat transfer efficiency and pumping power of an oil-based nanofluid as a coolant fluid

    Asadi, Meisam; Asadi, Amin; Aberoumand, Sadegh

    2018-01-01

    The present work aims to study heat transfer performance and pumping power of MgO-MWCNT/ thermal oil hybrid nanofluid. Using a KD2 Pro thermal analyzer, the thermal conductivity of the samples have been measured. The results showed an increasing trend for the thermal conductivity of the nanofluid...... by increasing the mass concentration and temperature, in which the maximum enhancement of thermal conductivity was approximately 65%. Predicting the thermal conductivity of the nanofluid, a highly accurate correlation in terms of solid concentration and temperature has been proposed. Moreover, the heat transfer...... nanofluid is highly efficient in heat transfer applications as a coolant fluid in both the laminar and turbulent flow regimes, although it causes a certain penalty in the pumping power....

  5. PUMPS

    Thornton, J.D.

    1959-03-24

    A pump is described for conveving liquids, particure it is not advisable he apparatus. The to be submerged in the liquid to be pumped, a conduit extending from the high-velocity nozzle of the injector,and means for applying a pulsating prcesure to the surface of the liquid in the conduit, whereby the surface oscillates between positions in the conduit. During the positive half- cycle of an applied pulse liquid is forced through the high velocity nozzle or jet of the injector and operates in the manner of the well known water injector and pumps liquid from the main intake to the outlet of the injector. During the negative half-cycle of the pulse liquid flows in reverse through the jet but no reverse pumping action takes place.

  6. An improved efficiency of fuzzy sliding mode control of permanent magnet synchronous motor for wind turbine generator pumping system

    Benchabane, F.; Titaouine, A.; Guettaf, A.; Yahia, K.; Taibi, D.; Bennis, O.

    2012-01-01

    This paper presents an analysis by which the dynamic performances of a permanent magnet synchronous motor (PMSM) motor is controlled through a hysteresis current loop and an outer speed loop with different controllers. The dynamics of the wind turbine pumping drive system with (PI) and a fuzzy sliding mode (FSM) speed controllers are presented. In order to optimize the overall system efficiency, a maximum power point tracker is also used. Simulation is carried out by formatting the mathematical model for wind turbine generator, motor and pump load. The results for such complicated and nonlinear system, with fuzzy sliding mode speed controller show improvement in transient response of the PMSM drive over conventional PI. The effectiveness of the FSM controller is also demonstrated. (author)

  7. Stator winding service life analysis of wet winding motor RCP

    Yeoh Eingyee; Li Linsen; Wang Ziguan; Shen Feng

    2017-01-01

    The reactor coolant pump type RUV of KSB is specially designed for the Westinghouse AP1000 reactor. It is hermetically sealed, wet winding motor pump. This wet winding motor concept has higher efficiency as compared to a canned motor since there are no eddy current losses, but this makes the stator cable insulator face the challenge of degradation by γ-irradiation in coolant. Radiation dose at the RUV is evaluated in interest from the material degradation point of view and the service life of stator winding in RUV is preliminarily analyzed. The values obtained are less than the dose rate and the dose which change the mechanical properties of the insulator material. According to the preliminary analysis, it can be concluded that the degradation of cable insulator of stator winding by γ-irradiation in coolant is negligible for the 60 years' service life of reactor coolant pump. This result also ensures crosslinked polyethylene used as cable insulator of stator winding is reliable in reactor coolant pump. (author)

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

    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.)

  9. An experimental and theoretical investigation on the effects of adding hybrid nanoparticles on heat transfer efficiency and pumping power of an oil-based nanofluid as a coolant fluid

    Asadi, Meisam; Asadi, Amin; Aberoumand, Sadegh

    2018-01-01

    The present work aims to study heat transfer performance and pumping power of MgO-MWCNT/ thermal oil hybrid nanofluid. Using a KD2 Pro thermal analyzer, the thermal conductivity of the samples have been measured. The results showed an increasing trend for the thermal conductivity of the nanofluid...... nanofluid is highly efficient in heat transfer applications as a coolant fluid in both the laminar and turbulent flow regimes, although it causes a certain penalty in the pumping power....... efficiency and pumping power in all the studied range of solid concentrations and temperatures have been theoretically investigated, based on the experimental data of dynamic viscosity and thermal conductivity, for both the internal laminar and turbulent flow regimes. It was observed that the studied......The present work aims to study heat transfer performance and pumping power of MgO-MWCNT/ thermal oil hybrid nanofluid. Using a KD2 Pro thermal analyzer, the thermal conductivity of the samples have been measured. The results showed an increasing trend for the thermal conductivity of the nanofluid...

  10. Development of the pump protection system against cavitation on the basis of the stator current signature analysis of drive electric motor

    Kipervasser, M. V.; Gerasimuk, A. V.; Simakov, V. P.

    2018-05-01

    In the present paper a new registration method of such inadmissible phenomenon as cavitation in the operating mode of centrifugal pump is offered. Influence of cavitation and extent of its development on the value of mechanical power consumed by the pump from the electric motor is studied. On the basis of design formulas the joint mathematical model of centrifugal pumping unit with the synchronous motor is created. In the model the phenomena accompanying the work of a pumping installation in the cavitation mode are considered. Mathematical modeling of the pump operation in the considered emergency operation is carried out. The chart of stator current of the electric motor, depending on the degree of cavitation development of is received. On the basis of the analysis of the obtained data the conclusion on the possibility of registration of cavitation by the current of drive electric motor is made and the functional diagram of the developed protection system is offered, its operation principle is described.

  11. Vibration Analysis and Experimental Research of the Linear-Motor-Driven Water Piston Pump Used in the Naval Ship

    Ye-qing Huang

    2016-01-01

    Full Text Available Aiming at the existing problems of traditional water piston pump used in the naval ship, such as low efficiency, high noise, large vibration, and nonintelligent control, a new type of linear-motor-driven water piston pump is developed and its vibration characteristics are analyzed in this research. Based on the 3D model of the structure, the simulation analyses including static stress analysis, modal analysis, and harmonic response analysis are conducted. The simulation results reveal that the mode shape under low frequency stage is mainly associated with the eccentricity swing of the piston rod. The vibration experiment results show that the resonance frequency of linear-motor-driven water piston pump is concentrated upon 500 Hz and 800 Hz in the low frequency range. The dampers can change the resonance frequency of the system to a certain extent. The vibration under triangular motion curve is much better than that of S curve, which is consistent with the simulation conclusion. This research provides an effective method to detect the vibration characteristics and a reference for design and optimization of the linear-motor-driven water piston pump.

  12. Optimum Design of a Moving Coil Actuator for Fast-Switching Valves in Digital Hydraulic Pumps and Motors

    Roemer, Daniel Beck; Bech, Michael Møller; Johansen, Per

    2015-01-01

    Fast-switching seat valves suitable for digital hydraulic pumps and motors utilize direct electromagnetic actuators, which must exhibit superior transient performance to allow efficient operation of the fluid power pump/motor. A moving coil actuator resulting in a minimum valve switching time...... is designed for such valves using transient finite-element analysis of the electromagnetic circuit. The valve dynamics are coupled to the fluid restrictive forces, which significantly influence the effective actuator force. Fluid forces are modeled based on transient computational fluid dynamics models....... The electromagnetic finite-element model is verified against experimental measurement, and used to design an optimum moving coil actuator for the application considering different voltage-current ratios of the power supply. Results show that the optimum design depends on the supply voltage-current ratio, however...

  13. A New Low-Cost Hybrid Switched Reluctance Motor for Adjustable-Speed Pump Applications

    Lu, Kaiyuan; Rasmussen, Peter Omand; Watkins, Steve

    2011-01-01

    with conventional single-phase switched reluctance motors, it has an increased torque density. The cogging torque is beneficially used in this motor for reducing the torque ripple. It is demonstrated that such a motor drive system can be a suitable candidate to advantageously compete with the existing motor drive...

  14. Thermal hydraulic conditions inducing incipient cracking in the 900 MWe unit 93 D reactor coolant pump shafts; Pompes primaires 93 D des tranches de 900 MW. Conditions thermo-hydrauliques d`amorcage des fissures d`arbres

    Bore, C.

    1995-12-31

    From 1987, 900 MWe plant operating feedback revealed cracking in the lower part of the reactor coolant pump shafts, beneath the thermal ring. Metallurgical examinations established that this was due to a thermal fatigue phenomenon known as thermal crazing, occurring after a large number of cycles. Analysis of thermal hydraulic conditions initiating the cracks does not allow exact quantification of the thermal load inducing cracking. Only qualitative analyses are thus possible, the first of which, undertaken by the pump manufacturer, Jeumont Industrie, showed that the cracks could not be due to the major transients (stop-start, injection cut-off), which were too few in number. Another explanation was then put forward: the thermal ring, shrunk onto the shaft it is required to protect against thermal shocks, loosens to allow an alternating downflow of cold water from the shaft seals and an upflow of hot water from the primary system. However, approximate calculations showed that the flow involved would be too slight to initiate the cracking observed. A more stringent analysis undertaken with the 2D flow analysis code MELODIE subsequently refuted the possibility of alternating flows beneath the ring establishing that only a hot water upflow occurred due to a `viscosity pump` phenomenon. Crack initiation was finally considered to be due to flowrate variations beneath the ring, with the associated temperature fluctuations. This flowrate fluctuation could be due to an unidentified transient phenomenon or to a variation in pump operating conditions. This analysis of the hydraulic conditions initiating the cracks disregards shaft surface residual stresses. These are tensile stresses and show that loads less penalizing than those initially retained could cause incipient cracking. Thermal ring modifications to reduce these risks were proposed and implemented. In addition, final metallurgical treatment of the shafts was altered and implemented. (Abstract Truncated)

  15. Secondary coolant purification system

    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

  16. Primary coolant circuits in FBR type reactors

    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.)

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

    Rathbun, R. J.

    1966-01-01

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

  18. Short-term memory of motor network performance via activity-dependent potentiation of Na+/K+ pump function.

    Zhang, Hong-Yan; Sillar, Keith T

    2012-03-20

    Brain networks memorize previous performance to adjust their output in light of past experience. These activity-dependent modifications generally result from changes in synaptic strengths or ionic conductances, and ion pumps have only rarely been demonstrated to play a dynamic role. Locomotor behavior is produced by central pattern generator (CPG) networks and modified by sensory and descending signals to allow for changes in movement frequency, intensity, and duration, but whether or how the CPG networks recall recent activity is largely unknown. In Xenopus frog tadpoles, swim bout duration correlates linearly with interswim interval, suggesting that the locomotor network retains a short-term memory of previous output. We discovered an ultraslow, minute-long afterhyperpolarization (usAHP) in network neurons following locomotor episodes. The usAHP is mediated by an activity- and sodium spike-dependent enhancement of electrogenic Na(+)/K(+) pump function. By integrating spike frequency over time and linking the membrane potential of spinal neurons to network performance, the usAHP plays a dynamic role in short-term motor memory. Because Na(+)/K(+) pumps are ubiquitously expressed in neurons of all animals and because sodium spikes inevitably accompany network activity, the usAHP may represent a phylogenetically conserved but largely overlooked mechanism for short-term memory of neural network function. Copyright © 2012 Elsevier Ltd. All rights reserved.

  19. Ion pump as Brownian motor: theory of electroconformational coupling and proof of ratchet mechanism for Na,K-ATPase action

    Tsong, Tian Yow; Chang, Cheng-Hung

    2003-04-01

    This article reviews some concepts of the Brownian Ratchet which are relevant to our discussion of mechanisms of action of Na,K-ATPase, a universal ion pump and an elemental motor protein of the biological cell. Under wide ranges of ionic compositions it can hydrolyze an ATP and use the γ-phosphorous bond energy of ATP to pump 3 Na + out of, and 2 K + into the cell, both being uphill transport. During the ATP-dependent pump cycle, the enzyme oscillates between E1 and E2 states. Our experiment replaces ATP with externally applied electric field of various waveforms, amplitudes, and frequencies. The field enforced-oscillation, or fluctuation of E1 and E2 states enables the enzyme to harvest energy from the applied field and convert it to the chemical gradient energy of cations. A theory of electroconformational coupling (TEC), which embodies all the essential features of the Brownian Ratchet, successfully simulates these experimental results. Our analysis based on a four-state TEC model indicates that the equilibrium and the rate constants of the transport system define the frequency and the amplitude of the field for the optimal activation. Waveform, frequency, and amplitude are three elements of signal. Thus, electric signal of the ion pump is found by TEC analysis of the experimental data. Electric noise (white) superimposed on an electric signal changes the pump efficiency and produces effects similar to the stochastic resonance reported in other biological systems. The TEC concept is compared with the most commonly used Michaelis-Menten enzyme mechanism (MME) for similarities and differences. Both MME and TEC are catalytic wheels, which recycle the catalyst in each turnover. However, a MME can only catalyze reaction of descending free energy while a TEC enzyme can catalyze reaction of ascending free energy by harvesting needed energy from an off-equilibrium electric noise. The TEC mechanism is shown to be applicable to other biological motors and engines, as

  20. Design and fabrication of magnetic coolant filter

    Prashanth, B. N.

    2017-07-01

    lose its strength even number of years of use. Dirty coolant is fed from the machines in to the reservoir of the coolant filter either by a pump or taken by the gravity and flows under the tray. This attracts the ferrous particles and builds up a cake of ferrous material and finally taken away by the scraper. The moving permanent magnets mounted on the shaft attracts ferrous chips and slide them on to plate and then to the discharge end or sludge bin. The coolant separated from chips flow back to the coolant tank. Well in this fast changing growth of metal working operation the recycling of cutting fluids become very important for the management of coolant. With the help of this developed model of magnetic coolant separator we can get highly efficient way of filtration guarantying fine finish, dimensional accuracy and increased tool life. The most significant role of this filter is that, it will reduce the waste disposal of coolant and a net profit for the production industries.

  1. BWR series pump recirculation system

    Dillmann, C.W.

    1992-01-01

    This patent describes a recirculation system for driving reactor coolant water contained in an annular downcomer defined between a boiling water reactor vessel and a reactor core spaced radially inwardly therefrom. It comprises a plurality of circumferentially spaced second pumps disposed in the downcomer, each including an inlet for receiving from the downcomer a portion of the coolant water as pump inlet flow, and an outlet for discharging the pump inlet flow pressurized in the second pump as pump outlet flow; and means for increasing pressure of the pump inlet flow at the pump inlet including a first pump disposed in series flow with the second pump for first receiving the pump inlet flow from the downcomer and discharging to the second pump inlet flow pressurized in the first pump

  2. Systematic Design of a Magnetically Levitated Brushless DC Motor for a Reversible Rotary Intra-Aortic Blood Pump.

    Wang, Yaxin; Logan, Thomas G; Smith, P Alex; Hsu, Po-Lin; Cohn, William E; Xu, Liping; McMahon, Richard A

    2017-10-01

    The IntraVAD is a miniature intra-aortic ventricular assist device (VAD) designed to work in series with the compromised left ventricle. A reverse-rotation control (RRc) mode has been developed to increase myocardial perfusion and reduce ventricular volume. The RRc mode includes forward rotation in systole and reverse rotation in diastole, which requires the IntraVAD to periodically reverse its rotational direction in synchrony with the cardiac cycle. This periodic reversal leads to changes in pressure force over the impeller, which makes the entire system less stable. To eliminate the mechanical wear of a contact bearing and provide active control over the axial position of the rotor, a miniature magnetically levitated bearing (i.e., the PM-Coil module) composed of two concentric permanent magnetic (PM) rings and a pair of coils-one on each side-was proposed to provide passive radial and active axial rotor stabilization. In the early design stage, the numerical finite element method (FEM) was used to optimize the geometry of the brushless DC (BLDC) motor and the maglev module, but constructing a new model each time certain design parameters were adjusted required substantial computation time. Because the design criteria for the module had to be modified to account for the magnetic force produced by the motor and for the hemodynamic changes associated with pump operation, a simplified analytic expression was derived for the expected magnetic forces. Suitable bearings could then be designed capable of overcoming these forces without repeating the complicated FEM simulation for the motor. Using this method at the initial design stage can inform the design of the miniature maglev BLDC motor for the proposed pulsatile axial-flow VAD. © 2017 International Center for Artificial Organs and Transplantation and Wiley Periodicals, Inc.

  3. Coolant cleanup method in a nuclear reactor

    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.)

  4. Coolant leakage detection device

    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.)

  5. Maximum Power Point Tracking for Brushless DC Motor-Driven Photovoltaic Pumping Systems Using a Hybrid ANFIS-FLOWER Pollination Optimization Algorithm

    Neeraj Priyadarshi

    2018-04-01

    Full Text Available In this research paper, a hybrid Artificial Neural Network (ANN-Fuzzy Logic Control (FLC tuned Flower Pollination Algorithm (FPA as a Maximum Power Point Tracker (MPPT is employed to amend root mean square error (RMSE of photovoltaic (PV modeling. Moreover, Gaussian membership functions have been considered for fuzzy controller design. This paper interprets the Luo converter occupied brushless DC motor (BLDC-directed PV water pump application. Experimental responses certify the effectiveness of the suggested motor-pump system supporting diverse operating states. The Luo converter, a newly developed DC-DC converter, has high power density, better voltage gain transfer and superior output waveform and can track optimal power from PV modules. For BLDC speed control there is no extra circuitry, and phase current sensors are enforced for this scheme. The most recent attempt using adaptive neuro-fuzzy inference system (ANFIS-FPA-operated BLDC directed PV pump with advanced Luo converter, has not been formerly conferred.

  6. Transient Analysis for Evaluating the Potential Boiling in the High Elevation Emergency Cooling Units of PWR Following a Hypothetical Loss of Coolant Accident (LOCA) and Subsequent Water Hammer Due to Pump Restart

    Husaini, S. Mahmood; Qashu, Riyad K.

    2004-01-01

    The Generic Letter GL-96-06 issued by the U.S. Nuclear Regulatory Commission (NRC) required the utilities to evaluate the potential for voiding in their Containment Emergency Cooling Units (ECUs) due to a hypothetical Loss Of Coolant Accident (LOCA) or a Main Steam Line Break (MSLB) accompanied by the Loss Of Offsite Power (LOOP). When the offsite power is restored, the Component Cooling Water (CCW) pumps restart causing water hammer to occur due to cavity closure. Recently EPRI (Electric Power Research Institute) performed a research study that recommended a methodology to mitigate the water hammer due to cavity closure. The EPRI methodology allows for the cushioning effects of hot steam and released air, which is not considered in the conventional water column separation analysis. The EPRI study was limited in scope to the evaluation of water hammer only and did not provide any guidance for evaluating the occurrence of boiling and the extent of voiding in the ECU piping. This paper presents a complete methodology based on first principles to evaluate the onset of boiling. Also, presented is a methodology for evaluating the extent of voiding and the water hammer resulting from cavity closure by using an existing generalized computer program that is based on the Method of Characteristics. The EPRI methodology is then used to mitigate the predicted water hammer. Thus it overcomes the inherent complications and difficulties involved in performing hand calculations for water hammer. The heat transfer analysis provides an alternative to the use of very cumbersome modeling in using CFD (computational fluid dynamics) based computer programs. (authors)

  7. HANARO secondary coolant management

    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

  8. Electromagnetic pump

    Ito, Koji; Suetake, Norio; Aizawa, Toshie; Nakasaki, Masayoshi

    1998-01-01

    The present invention provides an electromagnetic pump suitable to a recycling pump for liquid sodium as coolants of an FBR type reactor. Namely, a stator module of the electromagnetic pump of the present invention comprises a plurality of outer laminate iron core units and outer stator modules stacked alternately in the axial direction. With such a constitution, even a long electromagnetic pump having a large number of outer stator coils can be manufactured without damaging electric insulation of the outer stator coils. In addition, the inner circumferential surface of the outer laminate iron cores is urged and brought into contact with the outer circumferential surface of the outer duct by an elastic material. With such a constitution, Joule loss heat generated in the outer stator coils and internal heat generated in the outer laminate iron cores can be released to an electroconductive fluid flowing the inner circumference of the outer duct by way of the outer duct. (I.S.)

  9. Nuclear reactor coolant channels

    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)

  10. Coolant cleanup system for BWR type reactor

    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.)

  11. Coolant clean up system in nuclear reactor

    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.)

  12. A Methodology for Evaluation of Inservice Test Intervals for Pumps and Motor Operated Valves

    McElhaney, K.L.

    1999-01-01

    The nuclear industry has begun efforts to reevaluate inservice tests (ISTs) for key components such as pumps and valves. At issue are two important questions--What kinds of tests provide the most meaningful information about component health, and what periodic test intervals are appropriate? In the past, requirements for component testing were prescribed by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The tests and test intervals specified in the Code were generic in nature and test intervals were relatively short. Operating experience has shown, however, that performance and safety improvements and cost savings could be realized by tailoring IST programs to similar components with comparable safety importance and service conditions. In many cases, test intervals may be lengthened, resulting in cost savings for utilities and their customers

  13. Design, in-sodium testing and performance evaluation of annular linear induction pump for a sodium cooled fast reactor

    Nashine, B.K.; Rao, B.P.C.

    2014-01-01

    Highlights: • Derivation of applicable design equations. • Design of an annular induction pump based on these equations. • Testing of the designed pump in a sodium test facility. • Performance evaluation of the designed pump. - Abstract: Annular linear induction pumps (ALIPs) are used for pumping electrically conducting liquid metals. These pumps find wide application in fast reactors since the coolant in fast reactors is liquid sodium which a good conductor of electricity. The design of these pumps is usually done using equivalent circuit approach in combination with numerical simulation models. The equivalent circuit of ALIP is similar to that of an induction motor. This paper presents the derivation of equivalent circuit parameters using first principle approach. Sodium testing of designed ALIP using the equivalent circuit approach is also described and experimental results of the testing are presented. Comparison between experimental and analytical calculations has also been carried out. Some of the reasons for variation have also been listed in this paper

  14. Methodology to monitor and diagnostic vibrations of the motor-pumps used in the primary cooling system of IEAR-1 nuclear research reactor

    Benevenuti, Erion de Lima

    2004-01-01

    The objectives of this study are to establish a strategy to monitor and diagnose vibrations of the motor pumps used in the primary reactor cooling system of the IEA-R1 nuclear research reactor, to verify the possibility of using the existing installed monitoring vibration system and to implement such strategy in a continuous way. Four types of mechanical problems were considered: unbalancing, misalignment, gaps and faults in bearings. An adequate set of analysis tools, well established by the industry, was selected. These are: global measurements of vibration, velocity spectrum and acceleration envelope spectrum. Three sources of data and information were used; the data measured from the primary pumps, experimental results obtained with a Spectra Quest machine used to simulate mechanical defects and data from the literature. The results show that, for the specific case of the motor-pumps of IEA-R1 nuclear research reactor, although the technique using the envelope of acceleration, which is not available in the current system used to monitor the vibration of the motor pumps, is the one with best performance, the other techniques available in the system are sufficient to monitor the four types of mechanical problems mentioned. The proposed strategy is shown and detailed in this work. (author)

  15. Reactor auxiliary cooling facility and coolant supplying method therefor

    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.)

  16. Reactor auxiliary cooling facility and coolant supplying method therefor

    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.)

  17. Sizing modelling and simulation of PV generation system feeding a 6 MW induction motor dedicated to pumping applications

    Samir HADJERI

    2017-06-01

    Full Text Available This paper proposes an autonomous photovoltaic (PV pumping system for rural/remote applications especially in the large desert in northern Africa where the sun radiation is abundant. Since PV generators exhibit nonlinear I-V characteristics and their maximum power point varies with solar radiation. For this reason, the MPPT controller is used to optimize the solar energy conversion by guaranteeing fast maximum power point tracking (MPPT. This feature has an essential role in dynamic response and efficiency of the photovoltaic system, thus it maximizes the amount of extracted natural gas to be conveyed to another site for an ulterior utilization. Here a robust maximum power point tracker (MPPT using incremental conductance algorithm is applied to the duty cycle value of the DC-DC converter which acts directly on the drive speed. A DC/DC boost is used to enhance voltage up to the favourite level and the SVPWM inverter connects it to a powerful induction motor for a pumping of natural gas application, these systems are at the heart of many industrial sectors such as the oil industry, the production of thermal and nuclear energy etc. An understanding of how these systems operate is essential to increase their performance and reduce their operating costs. Furthermore, three phase voltage-fed PWM inverters are recently showing growing popularity for multi-megawatt industrial drive applications, the main purpose of these topologies is to provide a three-phase voltage source, where the amplitude, phase, and frequency of the voltages should always be controllable. The model of a three-phase voltage source inverter is modelled and discussed based on space vector modulation theory. Simulation results are obtained using MATLAB/Simulink environment for effectiveness of the proposed system.

  18. Efficiency increase of a high-efficient IEC 3 kW permanent-magnet-motor - Implementation in a pump system; Effizienzsteigerung eines hocheffizienten IEC 3 kW Permanent-Magnet-Motors inklusiv energetische Messung als Pumpenanwendung - Schlussbericht

    Lindegger, M. [Circle Motor AG, Guemligen (Switzerland); Biner, H.-P.; Evequoz, B. [University of Applied Sciences, Western Switzerland, Delemont (Switzerland); Salathe, D. [Fachhochschule Luzern, Technik und Architektur, Horw (Switzerland)

    2009-11-15

    The IEC norm 60034-30 defines new international standards for efficiency of electric motors. IE1 is standard efficiency, IE2 is high efficiency, and IE3/IE4 are the premium class. Circle Motor AG assumes that the premium class efficiency factor for motors with power less than 22 kW is easier to achieve with Permanent-Magnet-Motors enhanced by electric inverters than with asynchronous motors. The goal of this work is to achieve an overall 'efficiency factor' of 90% through a high-efficiency 3 kW-rated input Permanent-Magnet-Motor enhanced with an electric inverter, and then to perform measurements on the pump drive. The 'efficiency factor' measurement results showed worse performance than expected for both the inverter and the motor. One reason is due to high iron losses in the motor that are produced by high frequency currents of inverter. Another reason is the relatively high ohmic resistance of the copper motor winding caused by the chosen coil production method. By filtering high frequency currents at the rated power, the motor efficiency factor reaches up to 89.5%, which complies with premium class IE3. In generator mode the efficiency reaches 92.8%. We note that the positive result of this work is that the inverter can be produced with very low standby power losses - in our current work with 160 mW. It is necessary to admit that every detail can be responsible for success or failure during research and development activities of premium class motors. The optimization activities mainly belong to the area of additional losses of electric machines. In this report we have encountered all research results on the theme of Additional Losses that resulted from the current research work. The concept of drive control from the previous research project has shown itself as efficiency aware. The total efficiency factor of 90% can be reached by a system coupling a drive controller with a synchronous motor and the 'know how' from this work. The

  19. Pump characteristics and applications

    Volk, Michael

    2013-01-01

    Providing a wealth of information on pumps and pump systems, Pump Characteristics and Applications, Third Edition details how pump equipment is selected, sized, operated, maintained, and repaired. The book identifies the key components of pumps and pump accessories, introduces the basics of pump and system hydraulics as well as more advanced hydraulic topics, and details various pump types, as well as special materials on seals, motors, variable frequency drives, and other pump-related subjects. It uses example problems throughout the text, reinforcing the practical application of the formulae

  20. Maintenance experience on reactor recirculation pumps at Tarapur Atomic Power Station

    Singh, A.K.

    1995-01-01

    Reactor recirculation pumps at Tarapur Atomic Power Station (TAPS) are vertical, single stage centrifugal pumps having mechanical shaft seals and are driven by vertical mounted 3.3 kV, 3 phase, 1500 h.p. electric motors. During these years of operation TAPS has gained enough experience and expertise on the maintenance of reactor recirculation pumps which are dealt in this article. Failure of mechanical shaft seals, damage on pump carbon bearings, motor winding insulation failures and motor shaft damage have been the main areas of concern on recirculation pump. A detailed procedure step by step with component sketches has helped in eliminating errors during shaft seal assembly and installation. Pressure breakdown devices in seal assembly were rebuilt. Additional coolant water injection for shaft seal cooling was provided. These measures have helped in extending the reactor recirculation pump seal life. Pump bearing problems were mainly due to failure of anti-rotation pins and dowel pins of bearing assembly. These pins were redesigned and strengthened. Motor stator winding insulation failures were detected. Stator winding replacement program has been taken up on regular basis to avoid winding insulation failure due to aging. 3 refs., 2 tabs., 7 figs

  1. The selection of flying roller as an effort to increase the power of scooter-matic as the main power of centrifugal pump for fire fighter motor cycle

    Hadi Sutrisno, Himawan

    2018-03-01

    In densely populated settlements, fires often occur and cause losses. In some instances, the process of the occurrence of fires takes place so quickly that to reduce and avoid the occurrence of a fire disaster effort is required in accordance with the existing environmental condition. Fire fighter motorcycle by using motorcycle scooter-matic is considered suitable as one alternative to combating fire hazard in densely populated residential settlements. The use of motorcycle engines as the driving force of the pump often leads to unstable and not optimum power. Thus, the water spray on the centrifugal pump also becomes not maximum. To increase the engine power at scooter-matic engine idle rotation (700-2000 rpm), then the flying roller replacement with certain mass weight becomes an option. By selecting a 10 to 14 gram flying roller mass, the power analysis using a dynotest engine produces several variations. Of the calculation, the mass of a 14 gram flying roller provides a significant increase in motor power on the upper rotation. Meanwhile, on the lower power rotation using a flying roller with a mass of 10 grams provides an increase in power compared to a standard flying roller on a scooter matic motor engine. As a reference to the use of scooter-matic motor power as the pump power, the result of use of the flying roller with a mass of 10 grams becomes the best option.

  2. LWR primary coolant pipe rupture test rig

    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.)

  3. The sodium coolant

    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.)

  4. Nuclear reactor coolant channels

    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.)

  5. Extended Life Coolant Testing

    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

  6. Research on coolant radiochemistry

    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

  7. Seismic qualification of moderator system pump-motor units for RAPP-3,4 and KAIGA-1,2 235 MWe PHWRs

    Neelwarne, A.; Soni, R.S.; Kushawaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1992-01-01

    Smooth operation of active components like primary heat transport pumps, moderator pumps, emergency core cooling pumps etc. is always required to ensure safety of any nuclear power plants in case of normal as well as abnormal conditions such as earthquake loading. In order to ensure the functional requirement of such rotating equipment, is necessary to demonstrate, either through theoretical means or through experimental means, that in an event like earthquake loading, the static parts and the rotating parts of the equipment do not rub against each other giving rise to trouble during their operation. The moderator system pump units for RAPP-3,4 and Kaiga-1,2 have been analysed theoretically to demonstrate the structural integrity of various components of the unit as well as the functional requirement during an earthquake loading. A detailed Finite Element Model (FEM) was prepared for this which includes the modelling of static parts, rotating parts, anti-friction bearings and fluid-film journal bearings. Response spectrum analysis of the unit was carried out using the applicable floor response spectra for RAPP-3,4 and Kaiga-1,2 sites. It was concluded from this analysis that the pump-motor unit analysed meets the required design intent in terms of structural integrity and operability of the unit. The present report gives a detailed description of the problem, the development of FEM model, results and the conclusions arrived at. (author). 23 refs., 9 tabs., 17 figs

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

    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

  9. Pumps in nuclear power plants

    Kim, J.H.

    1991-01-01

    This paper reports that pumps play an important role in nuclear plant operation. For instance, reactor coolant pumps (RCPs) should provide adequate cooling for reactor core in both normal operation and transient or accident conditions. Pumps such as Low Pressure Safety Injection (LPSI) pump in the Emergency Core Cooling System (ECCS) play a crucial role during an accident, and their reliability is of paramount importance. Some key issues involved with pumps in nuclear plant system include the performance of RCP under two-phase flow conditions, piping vibration due to pump operating in two-phase flows, and reliability of LPSI pumps

  10. Triboengineering problems of lead coolant in innovative fast reactors

    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

  11. Coolant leakage detecting device

    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.)

  12. Reactor coolant cleanup device

    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.)

  13. Internal pump monitoring device

    Kurosaki, Toshikazu.

    1996-01-01

    In the present invention, a thermometer is disposed at the upper end of an internal pump casing of a coolant recycling system in a BWR type reactor to detect leakage of reactor water thereby ensuring the improvement of reliability of the internal pump. Namely, a thermometer is disposed, which can detect temperature elevation occurred when water in the internal pump leaked from a reactor pressure vessel passes through the gap between a stretch tube and an upper end of the pump casing. Signals from the thermometer are transmitted to a signal processing device by an instrumentation cable. The signal processing device generates an alarm when the temperature signal exceeds a predetermined value and announces that leakage of reactor water occurs in the internal pump. Since the present invention can detect the leakage of the reactor water in the pump casing in an early stage, it can contribute to the improvement of the safety and reliability of the internal pump. (I.S.)

  14. Coolant system decontamination

    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)

  15. Research on Coolant Radiochemistry

    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

  16. Coolant clean-up system in nuclear reactor

    Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

    1987-01-01

    Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

  17. Primary coolant recycling device for FBR type reactor

    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.)

  18. Sodium pumping: pump problems

    Guer, M.; Guiton, P.

    Information on sodium pumps for LMFBR type reactors is presented concerning ring pump design, pool reactor pump design, secondary pumps, sodium bearings, swivel joints of the oscillating annulus, and thermal shock loads

  19. Water pumping system using solar photovoltaic induction motor; Sistema de bombeamento de agua com energia solar fotovoltaica utilizando motor de inducao trifasico

    Andrade, Eduardo Henrique Pereira de; Bezerra, Luiz Daniel Santos; Antunes, Fernando Luiz Marcelo [Universidade Federal do Ceara (DEE/PPGEE/UFC), Fortaleza, CE (Brazil). Dept. de Engenharia Eletrica. Programa de Pos -Graduacao em Engenharia Eletrica; Borges Neto, Manuel Rangel [Centro Federal de Educacao Tecnologica de Petrolina (CEFET), PE (Brazil)

    2008-07-01

    One of the main difficulties to people who live in remote areas or isolated community and not grid connected, certainly is to access potable drink water. In the world, more than 6000 children dies everyday by some kind of illnesses associated to non-potable drink water. At state of Ceara, during the dry weather periods, remain water reservoir becomes practically a mud puddle, as a result, people and animals are forced to drink this inappropriate water. To minimize this consequences in this periods some water is distributed by tankers but, sometimes, even this water is not enough potable. Underground water is an alternative to mitigate this problem. The most common technique is the use of direct current (DC) pumps set supplied by solar photovoltaic systems. However, this kind of pump-set is relatively expensive and too hard to maintain. This paper brings an alternative lower expensive and sustainable to water pumping system, it uses a three phase induction machine coupled to an underwater centrifugal pump supplied by solar photovoltaic energy system. (author)

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

    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.)

  1. Detection of pump degradation

    Greene, R.H.; Casada, D.A.; Ayers, C.W.

    1995-08-01

    This Phase II Nuclear Plant Aging Research study examines the methods of detecting pump degradation that are currently employed in domestic and overseas nuclear facilities. This report evaluates the criteria mandated by required pump testing at U.S. nuclear power plants and compares them to those features characteristic of state-of-the-art diagnostic programs and practices currently implemented by other major industries. Since the working condition of the pump driver is crucial to pump operability, a brief review of new applications of motor diagnostics is provided that highlights recent developments in this technology. The routine collection and analysis of spectral data is superior to all other technologies in its ability to accurately detect numerous types and causes of pump degradation. Existing ASME Code testing criteria do not require the evaluation of pump vibration spectra but instead overall vibration amplitude. The mechanical information discernible from vibration amplitude analysis is limited, and several cases of pump failure were not detected in their early stages by vibration monitoring. Since spectral analysis can provide a wealth of pertinent information concerning the mechanical condition of rotating machinery, its incorporation into ASME testing criteria could merit a relaxation in the monthly-to-quarterly testing schedules that seek to verify and assure pump operability. Pump drivers are not included in the current battery of testing. Operational problems thought to be caused by pump degradation were found to be the result of motor degradation. Recent advances in nonintrusive monitoring techniques have made motor diagnostics a viable technology for assessing motor operability. Motor current/power analysis can detect rotor bar degradation and ascertain ranges of hydraulically unstable operation for a particular pump and motor set. The concept of using motor current or power fluctuations as an indicator of pump hydraulic load stability is presented

  2. Detection of pump degradation

    Greene, R.H.; Casada, D.A.; Ayers, C.W. [and others

    1995-08-01

    This Phase II Nuclear Plant Aging Research study examines the methods of detecting pump degradation that are currently employed in domestic and overseas nuclear facilities. This report evaluates the criteria mandated by required pump testing at U.S. nuclear power plants and compares them to those features characteristic of state-of-the-art diagnostic programs and practices currently implemented by other major industries. Since the working condition of the pump driver is crucial to pump operability, a brief review of new applications of motor diagnostics is provided that highlights recent developments in this technology. The routine collection and analysis of spectral data is superior to all other technologies in its ability to accurately detect numerous types and causes of pump degradation. Existing ASME Code testing criteria do not require the evaluation of pump vibration spectra but instead overall vibration amplitude. The mechanical information discernible from vibration amplitude analysis is limited, and several cases of pump failure were not detected in their early stages by vibration monitoring. Since spectral analysis can provide a wealth of pertinent information concerning the mechanical condition of rotating machinery, its incorporation into ASME testing criteria could merit a relaxation in the monthly-to-quarterly testing schedules that seek to verify and assure pump operability. Pump drivers are not included in the current battery of testing. Operational problems thought to be caused by pump degradation were found to be the result of motor degradation. Recent advances in nonintrusive monitoring techniques have made motor diagnostics a viable technology for assessing motor operability. Motor current/power analysis can detect rotor bar degradation and ascertain ranges of hydraulically unstable operation for a particular pump and motor set. The concept of using motor current or power fluctuations as an indicator of pump hydraulic load stability is presented.

  3. Device for preventing coolant outflow in a reactor

    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.)

  4. Photovoltaic pump systems

    Klockgether, J.; Kiessling, K. P.

    1983-09-01

    Solar pump systems for the irrigation of fields and for water supply in regions with much sunshine are discussed. For surface water and sources with a hoisting depth of 12 m, a system with immersion pumps is used. For deep sources with larger hoisting depths, an underwater motor pump was developed. Both types of pump system meet the requirements of simple installation and manipulation, safe operation, maintenance free, and high efficiency reducing the number of solar cells needed.

  5. Gas engine driven reversible heat pumps: Innovative design. Realizzazione di una pompa di calore reversibile azionata da motore a gas

    Canci, F.; Zecchin, M.

    1992-01-01

    This paper describes the development of a series of gas engine driven air-water compression heat pumps designed for reversible summer-winter operation. The development work was carried out within the framework of a joint venture combing the efforts of the Italian Gas Society, Natural Gas of Barcellona and Climaveneta of Vicenza (Italy), who acted as the heat pump constructor. The main objective of this venture was to develop a series of machines that would be suitable for the contemporaneous summer air conditioning and winter space heating of medium-sized buildings. The designs were optimized to allow cost and energy savings with respect to conventional equipment. The useful cooling power range of the innovative heat pump systems goes from 100 to 250 kW thus giving them the flexibility not yet afforded by conventional equipment currently sold on international markets. In addition to pointing out the new heat pumps' main design and performance features, this paper suggests some feasible applications.

  6. Design of automotive engine coolant hoses

    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. Cooling device for leaking fluid from a centrifugal pump

    Raymond, J.R.; Thomson, C.I.

    1978-01-01

    The patented device consists of an integrated heat exchanger in a centrifugal primary cooling circuit pump whose purpose is to cool the coolant medium which leaks along the pump shaft so that the shaft seals are not damaged. The cooling water passes through spirally arranged banks of tubes round the shaft, with baffle plates to direct the leaking coolant. (JIW)

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

    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

  9. Emergency recirculation pump driving mechanism

    Morooka, Shin-ichi.

    1980-01-01

    Purpose: To sufficiently secure the coolant flow rate in a reactor core and restrict the temperature on the surface of fuel elements to low degree when the coolant is lost in a BWR type reactor. Constitution: In order to secure sufficient coolant flow rate in a reactor core and to sufficiently cool the reactor core when the coolant is lost in a BWR type reactor, it is tripped upon loss of power supply simultaneously when an accident occurs, a recycling pump at the side of normal reactor where its rotating speed is decelerated in accordance with its inertia is restarted by the pressure water stored in a tank out of the reactor to increase the coolant flow rate in the reactor core so as to sufficiently cool the reactor core. (Aizawa, K.)

  10. Making effective use of rod pumping systems in coalbed methane applications

    Crivello, A. [eProduction Solutions Inc., Kingwood, TX (United States)

    2003-07-01

    The advantages of optimizing coalbed methane (CBM) operations are increased production, reduced expenses, improved efficiency, and better inventory. The author discussed the CBM production cycle and the possible artificial lift options, including electric submersible pump (ESP), plunger lift, primary coolant pump (PCP), and reciprocating rod lift. The presentation focused on the rod lift, as it represents a low to moderate capital expenditure, has good system efficiency, an excellent fluid volume range, an excellent salvage value, excellent familiarity with equipment, and has readily available parts and service. The major disadvantage of the rod lift is that the fixed operating range does not adapt to changing reservoir characteristics. A comparison between the rod pump controller and the variable speed drive was presented. The well can be operated at or near the pumped off condition with variable speed drives with rod pumping intelligence. The author provided a closer examination of the variable frequency drive and the vector flux drive. The presentation also included a discussion of prime movers, drive and inclinometer, gearbox loading, rod load limiter, and dynamometer cards. Three case studies were presented: CSW1, CSW2, and CSW3. It was concluded that wells must be kept pumping, and that a Flux Vector Drive should be used along with an NEMA B motor and properly sized pumping unit and pump. tabs., figs.

  11. Analysis of pump's shaft torsional vibrations in transient conditions

    Pasqualini, G.R.; Cauquelin, C.

    1989-01-01

    When the voltage is applied to an induction motor, the currents in the stator's phases are subject to a transient period. It is consequently also the case for the torques. A method to calculate the torque in the case of an induction motor with deep bars is presented. A model is proposed to represent the squirrel cage. It allows to take into account the fact the currents are not sinusoidal and that, in this case, the rotor's winding cannot be represented by only one resistance and once reactance. The electrical model is completed by a mechanical model for the shaftline. The calculation is realized for the start up of an reactor coolant pump. A comparison is made between the results given by the new model, by the classical model and by tests

  12. Pumps for German pressurized water reactors

    Dernedde, R.

    1984-01-01

    The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

  13. Dual coolant blanket concept

    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

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

    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

  15. Hydrodynamic problems of heavy liquid metal coolants technology in loop-type and mono-block-type reactor installations

    Orlov, Yuri I.; Efanov, Alexander D.; Martynov, Pyotr N.; Gulevsky, Valery A.; Papovyants, Albert K.; Levchenko, Yuri D.; Ulyanov, Vladimir V.

    2007-01-01

    In the report, the influence of hydrodynamics of the loop with heavy liquid metal coolants (Pb and Pb-Bi) on the realization methods and efficiency of the coolant technology for the reactor installations of loop, improved loop and mono-block type of design has been studied. The last two types of installations, as a rule, are characterized by the following features: availability of loop sections with low hydraulic head and low coolant velocities, large squares of coolant free surfaces; absence of stop and regulating valve, auxiliary pumps on the coolant pumping-over lines. Because of the different hydrodynamic conditions in the installation types, the tasks of the coolant technology have specific solutions. The description of the following procedures of coolant technology is given in the report: purification by hydrogen (purification using gas mixture containing hydrogen), regulation of dissolved oxygen concentration in coolant, coolant filtrating, control of dissolved oxygen concentration in coolant. It is shown that change of the loop design made with economic purpose and for improvement of the installation safety cause additional requirements to the procedures and apparatuses of the coolant technology realization

  16. Role of system characteristics in evolution of pump hydraulic design

    Walia, Mohinder; Misri, Vijay; Sharma, A.K.; Bapat, C.N.

    1994-01-01

    Primary heat transport (PHT) main circuit provides the means for transferring the heat produced in the fuel by circulating heavy water in the main circuit loop by primary coolant pumps (PCPs). The procurement specification of PCPs for 500 MWe pressurised heavy water reactor (PHWR) was prepared based upon the first order hydraulic analysis of the primary heat transport system and accordingly duty point was fixed. With this specification the manufacturer carried out model testing to arrive at optimum size of the impeller followed by determination of pump characteristics curves using full scale impeller during type testing. The duty point thus obtained was higher than specified necessitating the trimming of impeller. However, in order to make use of available higher duty point from system considerations, the duty point was redefined for production of subsequent pumps within specified tolerances governed by manufacturing limitations. PHT main system sizing (piping and feeders) was carried out based upon pump (delivering maximum flow) characteristics curve. Pressure profiles of PHT system at various operating modes were drawn and corresponding power drawn by motor was calculated. The interfacing of reactor coolant main system with hydraulic characteristics of PCP plays a significant role in establishing the requisite capability and capacity of PHT system in performing its intended functions. Therefore the paper traces the evolution of design parameters for PCP and subsequent generation of pressure profiles commensurate with the changes made in power profile including their impact on feeder sizing. The paper also highlights the scope of interaction between process designer and pump manufacturer in formulating a mutually acceptable and efficient hydraulic performance for PCP. (author). 3 refs., 8 figs., 3 tabs

  17. Energy efficiency in pumps

    Kaya, Durmus; Yagmur, E. Alptekin [TUBITAK-MRC, P.O. Box 21, 41470 Gebze, Kocaeli (Turkey); Yigit, K. Suleyman; Eren, A. Salih; Celik, Cenk [Engineering Faculty, Kocaeli University, Kocaeli (Turkey); Kilic, Fatma Canka [Department of Air Conditioning and Refrigeration, Kocaeli University, Kullar, Kocaeli (Turkey)

    2008-06-15

    In this paper, ''energy efficiency'' studies, done in a big industrial facility's pumps, are reported. For this purpose; the flow rate, pressure and temperature have been measured for each pump in different operating conditions and at maximum load. In addition, the electrical power drawn by the electric motor has been measured. The efficiencies of the existing pumps and electric motor have been calculated by using the measured data. Potential energy saving opportunities have been studied by taking into account the results of the calculations for each pump and electric motor. As a conclusion, improvements should be made each system. The required investment costs for these improvements have been determined, and simple payback periods have been calculated. The main energy saving opportunities result from: replacements of the existing low efficiency pumps, maintenance of the pumps whose efficiencies start to decline at certain range, replacements of high power electric motors with electric motors that have suitable power, usage of high efficiency electric motors and elimination of cavitation problems. (author)

  18. Energy efficiency in pumps

    Kaya, Durmus; Yagmur, E. Alptekin; Yigit, K. Suleyman; Kilic, Fatma Canka; Eren, A. Salih; Celik, Cenk

    2008-01-01

    In this paper, 'energy efficiency' studies, done in a big industrial facility's pumps, are reported. For this purpose; the flow rate, pressure and temperature have been measured for each pump in different operating conditions and at maximum load. In addition, the electrical power drawn by the electric motor has been measured. The efficiencies of the existing pumps and electric motor have been calculated by using the measured data. Potential energy saving opportunities have been studied by taking into account the results of the calculations for each pump and electric motor. As a conclusion, improvements should be made each system. The required investment costs for these improvements have been determined, and simple payback periods have been calculated. The main energy saving opportunities result from: replacements of the existing low efficiency pumps, maintenance of the pumps whose efficiencies start to decline at certain range, replacements of high power electric motors with electric motors that have suitable power, usage of high efficiency electric motors and elimination of cavitation problems

  19. SNR coolant system components

    De Haas Van Dorsser, A.H.; Mausbeck, H.

    1976-01-01

    The DEBENELUX prototype fast reactor power plant SNR 300 at Kalkar has a loop-type heat transfer system similar to that of the prototype LMFBR plants in the USA and Japan. There exist three 257 MW/sub th/ primary sodium loops, each with a hot leg centrifugal pump and three 85.6 MW/sub th/ intermediate heat exchangers in parallel. From there the heat is transferred to the steam generators via three secondary sodium loops with one cold leg sodium circulating pump in each. At a nominal reactor outlet temperature of 819 0 K and a turbine inlet power of 771 MW/sub th/ super heated steam of 166 bar and 733 0 K is produced, giving rise to a plant rating of 327 MW/sub e/ gross. The primary and secondary loops are described in detail

  20. Magnetohydrodynamic generator and pump system

    Birzvalk, Yu.A.; Karasev, B.G.; Lavrentyev, I.V.; Semikov, G.T.

    1983-01-01

    The MHD generator-pump system, or MHD coupling, is designed to pump liquid-metal coolant in the primary circuit of a fast reactor. It contains a number of generator and pump channels placed one after another and forming a single electrical circuit, but hydraulically connected parallel to the second and first circuits of the reactor. All the generator and pump channels are located in a magnetic field created by the magnetic system with an excitation winding that is fed by a regulated direct current. In 500 to 2000 MW reactors, the flow rate of the coolant in each loop of the primary circuit is 3 to 6 m 3 /s and the hydraulic power is 2 to 4 MW. This paper examines the primary characteristics of an MHD generator-pump system with various dimensions and number of channels, wall thicknesses, coolant flow rates, and magnetic fields. It is shown that its efficiency may reach 60 to 70%. The operating principle of the MHD generator-pump system is explained in the referenced patent and involves the transfer of hydraulic power from generator channels to pump channels using a magnetic field and electrical circuit common to both channels. Variations of this system may be analyzed using an equivalent circuit. 7 refs., 5 figs

  1. Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

    Luo Bangqi

    2012-01-01

    The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

  2. Coolant channel module CCM

    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

  3. Tendency of nuclear pumps for PWR primary system

    Shibata, Takeshi

    1976-01-01

    At present, large PWR power stations of more than 1,000 MW are successively constructed, and the pumps used there have become large. The progress and tendency of the technical development of main pumps in primary system are described. The increase of the capacity of power stations is accomplished by increasing the circulating coolant quantity per loop or the number of loops. Same standard primary coolant pumps are employed in the plants from 500 to 1,100 MW. The type of primary coolant pumps changed from canned type to shaft seal type, and the advantages of the shaft seal type are cheap production cost, high efficiency, and the easy utilization of inertia force. The bearings and shaft seals are thermally insulated from primary coolant. As for auxiliary pumps, reciprocating filling-up pumps and centrifugal high pressure injection pumps are used for 500 MW plants, but only centrifugal pumps are used for both purposes in 800 MW plants, and in 1,100 MW plants, the pumps of both types for separate purposes and centrifugal pumps for combined purposes are installed. Horizontal or vertical pumps of same type are used as containment vessel-spraying pumps and excess heat-eliminating pumps. The type of boric acid pumps changed from canned type to mechanical seal type. (Kako, I.)

  4. Coolant controls of a PEM fuel cell system

    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

  5. Improving motor reliability in nuclear power plants: Volume 3, Failure analysis and diagnostic tests on a naturally aged large electric motor

    Subudhi, M.; Taylor, J.H.; Sheets, M.W.

    1987-11-01

    Stator coils of a naturally failed 400 hp motor from the Brookhaven National Laboratory test reactor facility were tested for their dielectric integrities. The motor was used to drive the primary reactor coolant pump for the last 20 years. Maintenance activities on this motor during its entire service life were minimal, with the exception of meggering it periodically. The stator consisted of ninety individual coils which were separated for testing. Seven different dielectric tests were performed on the coils. Each set of data from the tested coils indicated a spectrum of variation depending on their aging conditions and characteristics. By comparing the test data to baseline data, the test methods were assessed for application to motor maintenance programs in nuclear power plants. Also included in this study are results of an investigation to determine the cause of this motor failure. Recommendations are provided on the aged condition of a second identical primary pump motor which is the same age and presently in operation. Recommendations are also presented relating to each of the dielectric test methods applicability to motor maintenance programs. 6 refs., 11 figs., 5 tabs

  6. Full reactor coolant system chemical decontamination qualification programs

    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.

  7. Pumps for nuclear facilities

    1999-01-01

    The guide describes how the Finnish Radiation and Nuclear Safety Authority (STUK) controls pumps and their motors at nuclear power plants and other nuclear facilities. The scope of the control is determined by the Safety Class of the pump in question. The various phases of the control are: (1) review of construction plan, (2) control of manufacturing, and construction inspection, (3) commissioning inspection, and (4) control during operation. STUK controls Safety Class 1, 2 and 3 pumps at nuclear facilities as described in this guide. STUK inspects Class EYT (non-nuclear) pumps separately or in connection with the commissioning inspections of the systems. This guide gives the control procedure and related requirements primarily for centrifugal pumps. However, it is also applied to the control of piston pumps and other pump types not mentioned in this guide

  8. Indigenous development of 20 Cu. M/hr flat linear induction pump (Paper No. 047)

    Prabhakar, R; Prakash, V; Sundarasekaran, S

    1987-01-01

    A distinctive physical property of sodium metal which is used as a coolant in fast reactors, is its high electrical conductivity. This together with its ability to wet stainless steel permits fluid pumping techniques using electromagnetic devices. Electromagnetic pumps are analogous to the electric motor, in which a force is produced by the interaction of magnetic field and current flowing in a conductor. Flat linear induction pump (FLIP) whose operating principle is similar to that of an induction motor is one of the types of electromagnetic pumps in wide use in auxilary circuits of fast reactors. As part of efforts to develop fast reactor components indigenously, work on the design and construction of a prototype FLIP rated for 20Cu.M/hr and 5Kg/sq.cm at 550degC was initiated. Under Board of Research in Nuclear Sciences scheme, the design was carried out by the Electrical Engineering Department of IIT, Madras. Pump was constructed at Engineering Development Division, Indira Gandhi Centre for Atomic Research, Kalpakkam. This paper presents in detail the work carried out for the fabrication of flow channel and for the stator assembly. Results obtained from dry electrical tests are also reported. Appendix summarises the design data. (author).

  9. Reactor coolant flow measurements at Point Lepreau

    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)

  10. Internal pump

    Kushima, Jun; Hayashi, Youjiro; Ueda, Masayuki.

    1997-01-01

    The present invention relates to an internal pump. A water hole allowing communication between internal and external circumferences of a stretch tube is provided at the portion of the stretch tube corresponding to a position where an end face of a nozzle portion of a motor case and an end face of a diffuser are joined with each other so that hot filtered water inside a pressure container which has entered from where the end face of the nozzle portion of the motor case and the end face of the diffuser are joined with each other is combined with the purged water so that it can be sent back to the pressure container again. (author) figs

  11. Implementation of multiple measures to improve reactor recirculation pump sealing performance in nuclear boiling water reactor service

    Loenhout, Gerard van [Flowserve B.V., Etten-Leur (Netherlands). Nuclear Services and Solutions Engineering; Hurni, Juerg

    2015-05-15

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core by feeding into multiple stationary jet pumps inside the vessel. Together with the jet pumps, they allow station operators to vary coolant flow and variable pump speed provides the best and most stable reactor power control. A crucial technical problem with a recirculation pump, such as a mechanical seal indicating loss of sealing pressure, may result in a power station having to shut down for repair. This article describes the sudden increase in stray current phenomenon leading to rapid and severe deterioration of the mechanical end face shaft seal in a reactor recirculation pump. This occurred after the installation of a variable frequency converter replacing the original motor-generator set. This article will also discuss the 2,500 hour laboratory test results conducted under reactor recirculation pump sealing conditions using a newly developed seal face technology recently implemented to overcome challenges when sealing neutral, ultra-pure water. In addition, the article will describe the elaborate shaft grounding arrangement and the preliminary measurement results achieved in order to eliminate potential damages to both pump and mechanical seal.

  12. Heat Radiators for Electromagnetic Pumps

    Campana, R. J.

    1986-01-01

    Report proposes use of carbon/carbon composite radiators in electromagnetic coolant pumps of nuclear reactors on spacecraft. Carbon/carbon composite materials function well at temperatures in excess of 2,200 K. Aluminum has melting temperature of only 880 K.

  13. Improvement to liquid metal pumps

    Pennell, W.E.

    1981-01-01

    This invention concerns the coolant pumps of nuclear reactors. It resolves the problems of structures which have to withstand high temperatures, the difficulties in keeping the multiple bearings of the shaft aligned, the excessive fluid flows, the risks of scoring and seizing-up by self welding, the need for narrow machining tolerances and the difficulties of access for inspection and repairs [fr

  14. Selected technological problems of repair of hydraulic drive systems for shearer loaders with the example of A2V-107 pumps and SHT-630W motors

    Kusak, E.; Paluch, J.

    1983-07-01

    Major elements of hydraulic drive systems as well as their wear and repair characteristics are described. Types of wear and standardized repair methods are comparatively evaluated. The evaluations are aimed at development of standardized procedures for use in large repair shops. The following stages of repair operations of A2V-107 pumps and SHT-630W motors are analyzed: disassembling hydraulic systems, washing and cleaning, classification of equipment elements (elements for scrapping and for regeneration), regeneration, assembling and final tests. The following regeneration methods are discussed: cutting, burnishing, bushing (e.g. the Heli Coil method), regeneration using copper, tin or zinc dusts and a temperature from 950 to 1,000 K under inert atmosphere, heat treatments. Methods are reviewed for comparative evaluations of repair efficiency and repair quality as well as documents used for recording repair in the shops. Economic aspects of using standardized procedures for repair of hydraulic equipment for shearer loaders are discussed and recommendations are made. (6 refs.)

  15. A Heuristic T-S Fuzzy Model for the Pumped-Storage Generator-Motor Using Variable-Length Tree-Seed Algorithm-Based Competitive Agglomeration

    Jianzhong Zhou

    2018-04-01

    Full Text Available With the fast development of artificial intelligence techniques, data-driven modeling approaches are becoming hotspots in both academic research and engineering practice. This paper proposes a novel data-driven T-S fuzzy model to precisely describe the complicated dynamic behaviors of pumped storage generator motor (PSGM. In premise fuzzy partition of the proposed T-S fuzzy model, a novel variable-length tree-seed algorithm based competitive agglomeration (VTSA-CA algorithm is presented to determine the optimal number of clusters automatically and improve the fuzzy clustering performances. Besides, in order to promote modeling accuracy of PSGM, the input and output formats in the T-S fuzzy model are selected by an economical parameter controlled auto-regressive (CAR model derived from a high-order transfer function of PSGM considering the distributed components in the water diversion system of the power plant. The effectiveness and superiority of the T-S fuzzy model for PSGM under different working conditions are validated by performing comparative studies with both practical data and the conventional mechanistic model.

  16. Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

    Navid Bozorgan

    2015-07-01

    Full Text Available Nanofluids are expected to be a promising coolant condidate in chemical processes for heat transfer system size reduction. This paper focuses on reducing the number of turns in a helical coil heat recovery exchanger with a given heat exchange capacity in a biomass heating plant using γ-Al2O3/n-decane nanofluid as coolant. The nanofluid flows through the tubes and the hot n-hexane flows through the shell. The numerical results show that using nanofluid as coolant in a helical coil heat exchanger can reduce the manufacturing cost of the heat exchanger and pumping power by reducing the number of turns of the coil.

  17. Optimum pump armament in supply networks; Optimal pumpebestykning i forsynignssystemer

    Kristjansson, Halldor

    2005-06-15

    The report presents recommendations as regards pump armament in district heating pump centres and the linked savings. A model has been built which shows variations of efficiency for motor and pump. (BA)

  18. Modal method for crack identification applied to reactor recirculation pump

    Miller, W.H.; Brook, R.

    1991-01-01

    Nuclear reactors have been operating and producing useful electricity for many years. Within the last few years, several plants have found cracks in the reactor coolant pump shaft near the thermal barrier. The modal method and results described herein show the analytical results of using a Modal Analysis test method to determine the presence, size, and location of a shaft crack. The authors have previously demonstrated that the test method can analytically and experimentally identify shaft cracks as small as five percent (5%) of the shaft diameter. Due to small differences in material property distribution, the attempt to identify cracks smaller than 3% of the shaft diameter has been shown to be impractical. The rotor dynamics model includes a detailed motor rotor, external weights and inertias, and realistic total support stiffness. Results of the rotor dynamics model have been verified through a comparison with on-site vibration test data

  19. Pump and Flow Control Subassembly of Thermal Control Subsystem for Photovoltaic Power Module

    Motil, Brian; Santen, Mark A.

    1993-01-01

    The pump and flow control subassembly (PFCS) is an orbital replacement unit (ORU) on the Space Station Freedom photovoltaic power module (PVM). The PFCS pumps liquid ammonia at a constant rate of approximately 1170 kg/hr while providing temperature control by flow regulation between the radiator and the bypass loop. Also, housed within the ORU is an accumulator to compensate for fluid volumetric changes as well as the electronics and firmware for monitoring and control of the photovoltaic thermal control system (PVTCS). Major electronic functions include signal conditioning, data interfacing and motor control. This paper will provide a description of each major component within the PFCS along with performance test data. In addition, this paper will discuss the flow control algorithm and describe how the nickel hydrogen batteries and associated power electronics will be thermally controlled through regulation of coolant flow to the radiator.

  20. Survey of pumps for tritium gas

    Dowell, T.M.

    1983-05-01

    This report considers many different types of pumps for their possible use in pumping tritium gas in the low, intermediate and high vacuum ranges. No one type of pump is suitable for use over the wide range of pumping pressure required in a typical pumping system. The favoured components for such a system are: bellows pump (low vacuum); orbiting scroll pump (intermediate vacuum); magnetically suspended turbomolecular pump (high vacuum); cryopump (high vacuum). Other pumps which should be considered for possible future development are: mound modified vane pump; SRTI wobble pump; roots pump with canned motor. It is proposed that a study be made of a future tritium pumping system in a Canadian tritium facility, e.g. a tritium laboratory

  1. Multi-stage internal gear/turbine fuel pump

    Maier, Eugen; Raney, Michael Raymond

    2004-07-06

    A multi-stage internal gear/turbine fuel pump for a vehicle includes a housing having an inlet and an outlet and a motor disposed in the housing. The multi-stage internal gear/turbine fuel pump also includes a shaft extending axially and disposed in the housing. The multi-stage internal gear/turbine fuel pump further includes a plurality of pumping modules disposed axially along the shaft. One of the pumping modules is a turbine pumping module and another of the pumping modules is a gerotor pumping module for rotation by the motor to pump fuel from the inlet to the outlet.

  2. Sodium as a reactor coolant

    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

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

    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.

  4. Analysis of water flow, power and electrical current of a motor pump powered by photovoltaic panels; Analise da vazao, potencia e corrente em uma motobomba acionada por paines fotovoltaicos

    Oyama, Paulo Takashi; Santos, Jose Airton Azevedo dos; Michels, Roger Nabeyama [Universidade Tecnologica Federal do Parana (UTFPR-MD), Medianeira, PR (Brazil)], E-mail: airton@utfpr.edu.br; Ricieri, Reinaldo Prandini [Universidade do Oeste Paulista (UNIOESTE), Cascavel, PR (Brazil). Dept. de Engenharia Agricola], E-mail: ricieri@unioeste.br; Halmeman, Maria Cristina Rodrigues [Universidade Tecnologica Federal do Parana (UTFPR-CM), Campo Mourao, PR (Brazil)

    2009-07-01

    Because of the high cost of electricity connection to distant places, far away from conventional nets, photovoltaic solar energy is widely used in public illumination, water pumping, medical care centers, etc. With the purpose to contribute to a better utilization in this type of energy, the object of the project is to analyze the water flow, power and electrical current of a motor pump to be used for water lift, driven by photovoltaic panels, the data will be collected at every 2 meters in height from 2.2 to 18.2 meters. The analyses show the outflow linear dependence in relation to the height of the water lift, a second order dependence of the electrical power provided by the photovoltaic panels in relation to the solar irradiance on the panels and a linear dependence of the electrical current in relation to the solar irradiance. This study is a part of a project developed at the 'Universidade Tecnologica Federal do Parana' (UTFPR) - Campus Medianeira - PR. Two photovoltaic panels were used from Solarex Brand, model MSX 56, a motor pump made by Shurflo Ltd, model 2088-732, two pyranometers from KIPP and ZONEN CM3, a type K thermocouple, a pressure transducer model HUBA-510 and a water flow transducer. (author)

  5. Cooling devices and methods for use with electric submersible pumps

    Jankowski, Todd A; Hill, Dallas D

    2014-12-02

    Cooling devices for use with electric submersible pump motors include a refrigerator attached to the end of the electric submersible pump motor with the evaporator heat exchanger accepting all or a portion of the heat load from the motor. The cooling device can be a self-contained bolt-on unit, so that minimal design changes to existing motors are required.

  6. Nuclear reactor coolant and cover gas system

    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

  7. Application of damage function analysis to reactor coolant circuits

    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)

  8. Application of damage function analysis to reactor coolant circuits

    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)

  9. Detection of pump degradation

    Casada, D.

    1995-01-01

    There are a variety of stressors that can affect the operation of centrifugal pumps. Although these general stressors are active in essentially all centrifugal pumps, the stressor level and the extent of wear and degradation can vary greatly. Parameters that affect the extent of stressor activity are manifold. In order to assure the long-term operational readiness of a pump, it is important to both understand the nature and magnitude of the specific degradation mechanisms and to monitor the performance of the pump. The most commonly applied method of monitoring the condition of not only pumps, but rotating machinery in general, is vibration analysis. Periodic or continuous special vibration analysis is a cornerstone of most pump monitoring programs. In the nuclear industry, non-spectral vibration monitoring of safety-related pumps is performed in accordance with the ASME code. Pump head and flow rate are also monitored, per code requirements. Although vibration analysis has dominated the condition monitoring field for many years, there are other measures that have been historically used to help understand pump condition; advances in historically applied technologies and developing technologies offer improved monitoring capabilities. The capabilities of several technologies (including vibration analysis, dynamic pressure analysis, and motor power analysis) to detect the presence and magnitude of both stressors and resultant degradation are discussed

  10. Development of Reactor Coolant Pump for APR1400

    Bang, Sang-Youn; Chu, Sung-Min; Chang, Jin-Young [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of)

    2015-10-15

    The development was focused on the performance requirements for APR1400 and to achieve the goals of the safety, reliability and adaptability for APR1400 system design. In addition, APR1400 RCP design was customized considering convenience of installation, operation and maintainability. This paper describes the details of the development process, improved design feature and type test results. Based on development of core technology of RCP, DOOSAN supplies the localized and improved APR1400 RCP to Shin-Hanul 1 and 2 Project. This would be good experience that the RCP core technology can break foreign monopoly in supplying the domestic nuclear industry. Also, there expect APR1400 RCP can be sustainable revenue models in nuclear industry. Moreover, development of RCP will be a catalyst to enhance design capacity for equipment and system of nuclear power plant as well as evaluation and verification skills of Korean nuclear industry.

  11. Examination of reactor coolant pump shaft at Crystal River-3

    Hayner, G.O.; Frye, C.R.; Clary, M.D.

    1987-01-01

    A detailed examination was performed on a broken RCP shaft assembly. The primary fracture was located in a groove under the upper end of the hydrostatic bearing journal. Additionally, all four impeller-to-shaft bolts and one drive pin failed. Mechanical properties, bulk chemistry, hardness, and microstructure were normal for the A-286 material used for the shaft, bolts and pins. A zone of axial surface cracking was seen just above the top of the hydrostatic bearing journal. According to Yoon et.al. these cracks are caused by thermal fatigue resulting from turbulent hot and cold water mixing in this area and have a self limiting depth. The primary RCP shaft fracture was caused by high cycle fatigue. Crack initiation probably occurred during initial use of the RCP and according to Yoon was caused by a combination postulated effect of comined surface residual stresses and stress concentration in the groove area. Several combinations of effects including broken impeller bolts probably were responsible for the initial crack propagation. Fracture mechanics testing results in 550 0 F air and simulated PWR water were used to estimate the stress intensity range of the primary crack and the crack propagation time by comparison of the fracture surface features. These estimates indicated that the propagation time was probably in the range from ≅ 191 to ≅ 323 days with a maximum stress intensity level of ≅ 30 ksi √(in). (orig.)

  12. Coolant monitoring systems for PWR reactors

    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. Alignment analysis of a vertical sodium pump

    Gupta, V.K.; Fair, C.E.

    1981-01-01

    With the objective of identifying important alignment features of pumps such as FFTF, HALLAM, EBR II, PNC, PHENIX, and CRBR, alignment of the vertical sodium pump for the Clinch River Breeder Reactor Plant (CRBRP) is investigated. The CRBRP pump includes a flexibly coupled pump shaft and motor shaft, two oil-film tilting-pad hydrodynamic radial bearings in the motor plus a vertical thrust bearing, and two sodium hydrostatic bearings straddling the double-suction centrifugal impeller in the pump. The assembled CRBRP prototype pump shows smooth predictable vibration behavior experienced during water test. An ealier swing check of the pump shaft about the motor shaft hub demonstrated that the pump is relatively insensitive to manufacturing and assembly tolerances, a consequence of close dimensional control and unique alignment features. (orig./GL)

  14. An economic evaluation comparison of solar water pumping system with engine pumping system for rice cultivation

    Treephak, Kasem; Thongpron, Jutturit; Somsak, Dhirasak; Saelao, Jeerawan; Patcharaprakiti, Nopporn

    2015-08-01

    In this paper we propose the design and economic evaluation of the water pumping systems for rice cultivation using solar energy, gasoline fuel and compare both systems. The design of the water and gasoline engine pumping system were evaluated. The gasoline fuel cost used in rice cultivation in an area of 1.6 acres. Under same conditions of water pumping system is replaced by the photovoltaic system which is composed of a solar panel, a converter and an electric motor pump which is compose of a direct current (DC) motor or an alternating current (AC) motor with an inverter. In addition, the battery is installed to increase the efficiency and productivity of rice cultivation. In order to verify, the simulation and economic evaluation of the storage energy battery system with batteries and without batteries are carried out. Finally the cost of four solar pumping systems was evaluated and compared with that of the gasoline pump. The results showed that the solar pumping system can be used to replace the gasoline water pumping system and DC solar pump has a payback less than 10 years. The systems that can payback the fastest is the DC solar pumping system without batteries storage system. The system the can payback the slowest is AC solar pumping system with batteries storage system. However, VAC motor pump of 220 V can be more easily maintained than the motor pump of 24 VDC and batteries back up system can supply a more stable power to the pump system.

  15. Impulse pumping modelling and simulation

    Pierre, B; Gudmundsson, J S

    2010-01-01

    Impulse pumping is a new pumping method based on propagation of pressure waves. Of particular interest is the application of impulse pumping to artificial lift situations, where fluid is transported from wellbore to wellhead using pressure waves generated at wellhead. The motor driven element of an impulse pumping apparatus is therefore located at wellhead and can be separated from the flowline. Thus operation and maintenance of an impulse pump are facilitated. The paper describes the different elements of an impulse pumping apparatus, reviews the physical principles and details the modelling of the novel pumping method. Results from numerical simulations of propagation of pressure waves in water-filled pipelines are then presented for illustrating impulse pumping physical principles, and validating the described modelling with experimental data.

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

    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. Trace organics in AGR coolants

    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)

  18. Quantitative determination of a hydrogen impurity in a sodium coolant by hydride thermal dissociation

    Ivanovskiy, M. N.; Pavlova, G. D.; Shmatko, B. A.; Milovidova, A. V.; Konovalov, E. YE.; Arnoldov, M. N.; Pleshivtsev, A. D.

    1988-01-01

    A molten sodium coolant containing hydrogen was heated in a vacuum at 450 C, and the gases generated pumped through a liquid nitrogen trap, and the H2 was then oxidized on a copper oxide substrate heated to 400 C. The accuracy of the method is 1.5 percent and the sensitivity is 1x10 to the -5 wt percent hydrogen.

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

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

    1997-04-01

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

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

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

    1997-01-01

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

  1. A novel static frequency converter based on multilevel cascaded H-bridge used for the startup of synchronous motor in pumped-storage power station

    Wang Feng, E-mail: sjtuwfeng@hotmail.co [Key Lab of Control of Power Transmission and Transformation, Ministry of Education, Department of Electrical Engineering, Shanghai Jiaotong University, Minhang District, Shanghai 200240 (China); Jiang Jianguo, E-mail: jiang@sjtu.edu.c [Key Lab of Control of Power Transmission and Transformation, Ministry of Education, Department of Electrical Engineering, Shanghai Jiaotong University, Minhang District, Shanghai 200240 (China)

    2011-05-15

    Research highlights: {yields} A novel Static Frequency Converter (SFC) based on multilevel cascaded H-bridge (CHB) topology is proposed and used for the reversible pump-generating units in pumped-storage power station. {yields} The novel SFC based on CHB has compact configuration, low current harmonic distortion and fast speed response. {yields} Rotor position and Grid connection are realized successfully by the novel SFC. -- Abstract: A novel static frequency converter (SFC) is proposed and is used firstly to start the reversible pump-generating units in pumped-storage power station. Multilevel cascaded H-bridge (CHB) topology and Insulated Gate Bipolar Transistor (IGBT) are applied in the novel SFC. In comparison with the conventional SFC adopting load-commutated inverter (LCI) which is composed of silicon-controlled rectifier (SCR), the novel one has plenty of advantages such as compact configuration, low current harmonic distortion and fast speed response, and these advantages have been verified during 2-year operation at Xiang Hong Dian Pumped-storage power station in China. This application shows that the novel SFC greatly enhances the reliability and success rate of connecting to grid for starting up the pump-generating units. The principle, characteristic and performance of the novel SFC are described in this paper, and some key issues related to the startup of the units of the pumped-storage power station are also presented.

  2. Vision system for precision alignment of coolant channels

    Kar, S.; Rao, Y.V.; Valli Kumar; Joshi, D.G.; Chadda, V.K.; Nigam, R.K.; Kayal, J.N.; Panwar, S.; Sinha, R.K.

    1997-01-01

    This paper describes a vision system which has been developed for precision alignment of Coolant Channel Replacement Machine (CCRM) with respect to the front face of the coolant channel under repair/replacement. It has provisions for automatic as well as semi-automatic alignment. A special lighting scheme has been developed for providing illumination to the front face of the channel opening. This facilitates automatic segmentation of the digitized image. The segmented image is analysed to obtain the centre of the front face of the channel opening and thus the extent of misalignment i.e. offset of the camera with respect to the front face of the channel opening. The offset information is then communicated to the PLC to generate an output signal to drive the DC servo motors for precise positioning of the co-ordinate table. 2 refs., 5 figs

  3. Penis Pump

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  4. Method of decontaminating primary coolant circuits

    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.)

  5. Equivalent mechanical parameters of oscillating rotary motors used in transport equipment

    A. Andziulis

    2006-12-01

    Full Text Available In various appliances and equipment of sundry transport means there is a lot of diverse mechanisms of periodical movement. So, the various piston or membrane pumps of fuel feeding and lubrication systems, circulation pumps, air and refrigerating coolant compressors, etc. are the typical examples of innovative and well promising application of the oscillating motors. In these cases the moving part of a motor can be directly connected to the working body of driven mechanism without the additional gears. Consequently, the drive can be simplified in design and improved in efficiency and reliability. Application of the oscillating rotary motors, if used in the aforesaid devices, strictly depends on specific properties of mechanical system of a motor aggregated with the driven mechanism and considered as the one-piece unit on the whole. So, this study analyses how the properties of mechanical system, comprised of two moving parts interconnected eccentrically or centrically, can be evaluated by the equivalent rotational inertia, equivalent mass and by equivalent mechanical power factor which, in turn, determine the operating characteristics and basic possibilities of the motor.

  6. High reliability flow system - an assessment of pump reliability and optimisation of the number of pumps

    Butterfield, J.M.

    1981-01-01

    A system is considered where a number of pumps operate in parallel. Normally, all pumps operate, driven by main motors fed from the grid. Each pump has a pony motor fed from an individual battery supply. Each pony motor is normally running, but not engaged to the pump shaft. On demand, e.g. failure of grid supplies, each pony motor is designed to clutch-in automatically when the pump speed falls to a specified value. The probability of all the pony motors failing to clutch-in on demand must be demonstrated with 95% confidence to be less than 10 -8 per demand. This assessment considers how the required reliability of pony motor drives might be demonstrated in practice and the implications on choice of the number of pumps at the design stage. The assessment recognises that not only must the system prove to be extremely reliable, but that demonstration that reliability is adequate must be done during plant commissioning, with practical limits on the amount of testing performed. It is concluded that a minimum of eight pony motors should be provided, eight pumps each with one pony motor (preferred) or five pumps each with two independent pony motors. A minimum of two diverse pony motor systems should be provided. (author)

  7. Alignment and operability analysis of a vertical sodium pump

    Gupta, V.K.; Fair, C.E.

    1981-01-01

    With the objective of identifying important alignment features of pumps such as FFTF, HALLAM, EBR II, PNC, PHENIX, and CRBR, alignment of the vertical sodium pump for the Clinch River Breeder Reactor Plant (CRBRP) is investigated. The CRBRP pump includes a flexibly coupled pump shaft and motor shaft, two oil-film tilting-pad hydrodynamic radial bearings in the motor plus a vertical thrust bearing, and two sodium hydrostatic bearings straddling the double-suction centrifugal impeller in the pump

  8. Detection of pump degradation

    Casada, D.

    1994-01-01

    There are a variety of stressors that can affect the operation of centrifugal pumps. Although these general stressors are active in essentially all centrifugal pumps, the stressor level and the extent of wear and degradation can vary greatly. Parameters that affect the extent of stressor activity are manifold. In order to assure the long-term operational readiness of a pump, it is important to both understand the nature and magnitude of the specific degradation mechanisms and to monitor the performance of the pump. The most commonly applied method of monitoring the condition of not only pumps, but rotating machinery in general, is vibration analysis. Periodic or continuous spectral vibration analysis is a cornerstone of most pump monitoring programs. In the nuclear industry, non-spectral vibration monitoring of safety-related pumps is performed in accordance with the ASME code. Although vibration analysis has dominated the condition monitoring field for many years, there are other measures that have been historically used to help understand pump condition: advances in historically applied technologies and developing technologies offer improved monitoring capabilities. The capabilities of several technologies (including vibration analysis, dynamic pressure analysis, and motor power analysis) to detect the presence and magnitude of both stressors and resultant degradation are discussed

  9. Review of magnetohydrodynamic pump applications

    O.M. Al-Habahbeh

    2016-06-01

    Full Text Available Magneto-hydrodynamic (MHD principle is an important interdisciplinary field. One of the most important applications of this effect is pumping of materials that are hard to pump using conventional pumps. In this work, the progress achieved in this field is surveyed and organized according to the type of application. The literature of the past 27 years is searched for the major developments of MHD applications. MHD seawater thrusters are promising for a variety of applications requiring high flow rates and velocity. MHD molten metal pump is important replacement to conventional pumps because their moving parts cannot stand the molten metal temperature. MHD molten salt pump is used for nuclear reactor coolants due to its no-moving-parts feature. Nanofluid MHD pumping is a promising technology especially for bioapplications. Advantages of MHD include silence due to no-moving-parts propulsion. Much progress has been made, but with MHD pump still not suitable for wider applications, this remains a fertile area for future research.

  10. Experimental study on utilization of air-borne jet sound in coolant leak detector

    Hayamizu, Y.; Kitahara, T.; Hayashi, T.; Nishimura, M.

    1975-10-01

    Studies have been undertaken to develop a new coolant leak detection method by the use of a microphone to pick up jet sound generated when pressurized high temperature water is discharged from a pressure boundary into the atmosphere. Leakage was simulated in three shapes, such as two machine-made circular holes and longitudinal and transverse slits in an inlet tube of a blowdown test facility. The measured power level of the jet sound was in agreement with theoretical values calculated from Lighthill's equation. In the study of utilization, this new method has been confirmed as applicable, and to be calculated theoretically for design on 'signal to noise ratio' evaluation. Detection of a small coolant leakage of 1 kg/sec is possible in a recirculation pump room which has large background noise from the pump if a suitable isolation wall, such as hot boxes, is installed between the monitored pipes and the pump. (auth.)

  11. Coolant mixing in pressurized water reactors. Proceedings

    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

  12. Experimental and numerical investigation of the coolant mixing during fast deboration transients

    Hoehne, T.; Rohde, U.; Weiss, F.P.

    1999-01-01

    For the analysis of boron dilution transients and main steam line break scenarios the modeling of the coolant mixing inside the reactor vessel is important, because the reactivity insertion strongly depends on boron acid concentration or the coolant temperature distribution. Calculations for steady state flow conditions for the VVER-440 were performed with a CFD code (CFX-4). The comparison with experimental data and an analytical mixing model which is implemented in the neutron-kinetic code DYN3D showed a good agreement for near-nominal conditions. First experiments at the Rossendorf Mixing Test Facility ROCOM were performed simulating the start-up of the first main coolant pump. The reference reactor for the geometrically 1:5 scaled Plexiglas model is the German Konvoi type PWR. After demonstrating the capability of the CFD code to simulate these complicated flow transients, calculations were performed for the start-up of the first pump in a VVER-440 type reactor. These calculations are a first step of understanding the coolant mixing in the RPV of a VVER-440 type reactor under transient conditions. The results of the calculation show a very complex flow in the downcomer. A high downcomer of VVER-440 and the existence of the lower control rod chamber support coolant mixing is concluded. (author)

  13. Centrifugal pumps

    Anderson, HH

    1981-01-01

    Centrifugal Pumps describes the whole range of the centrifugal pump (mixed flow and axial flow pumps are dealt with more briefly), with emphasis on the development of the boiler feed pump. Organized into 46 chapters, this book discusses the general hydrodynamic principles, performance, dimensions, type number, flow, and efficiency of centrifugal pumps. This text also explains the pumps performance; entry conditions and cavitation; speed and dimensions for a given duty; and losses. Some chapters further describe centrifugal pump mechanical design, installation, monitoring, and maintenance. The

  14. Coolant inlet device for nuclear reactors

    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.)

  15. Organic coolant for ARIES-III

    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

  16. Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

    Elina Syrjaelahti; Anitta Haemaelaeinen [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

    2005-07-01

    Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

  17. Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

    Elina Syrjaelahti; Anitta Haemaelaeinen

    2005-01-01

    Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

  18. Physical properties of organic coolants

    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)

  19. Cleaning of aluminum after machining with coolants

    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

  20. Method of suppressing the deposition of Co-60 to primary coolant pipeways in a nuclear reactor

    Hoshi, Michio; Tachikawa, Enzo; Goto, Satoshi; Sagawa, Chiaki; Yonezawa, Chushiro.

    1987-01-01

    Purpose: To suppress the deposition of Co-60 to primary coolant pipeways in a nuclear reactor. Method: To reduce the accumulation of Co-60 by causing chemical species of extremely similar chemical property with soluble Co-60 to be present together in coolants and replacing the deposition of Co-60 to the primary coolant pipeways in a nuclear reactor with that of the coexistent chemical spacies. Ni or Zn is used as the coexistet chemical spacies of similar chemical property with Co-60. The coexistent amount is from 5 to 10 times of the soluble Co-60 in the primary coolants. Ni or Zn solution adjusted with concentration is poured into and mixed with the coolants from a water feed source by using a high pressure constant volume pump. The amount of Co-60 taken into the pipeways caused by corrosion due to high temperature coolant is reduced to about 1/5 as compared with the case of Co-60 alone if 1 ppb of soluble Co-60 is present in water and 5 ppb of soluble Ni or Zn is added and, reduced to 1/12 if the amount of Ni or Zn is 10 ppb. (Kamimura, M.)

  1. Assessing the energy efficiency of pumps and pump units background and methodology

    Bernd Stoffel, em Dr-Ing

    2015-01-01

    Assessing the Energy Efficiency of Pumps and Pump Units, developed in cooperation with Europump, is the first book available providing the background, methodology, and assessment tools for understanding and calculating energy efficiency for pumps and extended products (pumps+motors+drives). Responding to new EU requirements for pump efficiency, and US DOE exploratory work in setting pump energy efficiency guidelines, this book provides explanation, derivation, and illustration of PA and EPA methods for assessing energy efficiency. It surveys legislation related to pump energy eff

  2. Preventive maintenance basis: Volume 10 -- High voltage electric motors (5 kV and greater). Final report

    Worledge, D.; Hinchcliffe, G.

    1997-07-01

    US nuclear plants are implementing preventive maintenance (PM) tasks with little documented basis beyond fundamental vendor information to support the tasks or their intervals. The Preventive Maintenance Basis project provides utilities with the technical basis for PM tasks and task intervals associated with 40 specific components such as valves, electric motors, pumps, and HVAC equipment. This report provides an overview of the PM Basis project and describes use of the PM Basis database. Volume 10 of the report provides a program of PM tasks suitable for application to high voltage (5kV and greater) electric motors in nuclear power plants. The PM tasks that are recommended provide a cost-effective way to intercept the causes and mechanisms that lead to degradation and failure. They can be used, in conjunction with material from other sources, to develop a complete PM program or to improve an existing program. Users of this information will be utility managers, supervisors, craft technicians, and training instructors responsible for developing, optimizing, or fine-tuning PM programs. Reactor Coolant Pumps motors (RCP's) are not excluded from this report in so far as good PM practices for motors of the appropriate class are concerned. However, the special auxiliary equipment normally associated with RCP's has not been included. Consequently, this report does not provide a complete PM program for RCP's. Industry and vendor programs for RCP's should be consulted for complete definition of RCP motor PM programs

  3. Report on measurements at the pump Avala - Annex 7

    Nikolic, M.

    1963-01-01

    Visual inspection and measuring results have shown that the surface of the upper pump bearing is much more worn-out than the lower radial bearing. This has proved that most of the cobalt (contained in the stellite alloy) came from the upper pump bearings. It could be stated that about 60 grams of cobalt from the upper pump bearings could come into the coolant system [sr

  4. Pump audit identifies the energy consumers; Pump Audit identifiziert gezielt die Energieverschwender

    Siekemann, Ralf [Grundfos GmbH, Erkrath (Germany)

    2011-03-14

    Engineering and services, which enhance and secure the energy efficiency of pump systems, are demanded. Speed-regulated pumps, included high efficient motors and a service philosophy, be designed on value conservation and cost control, are as well as professional ''pump-audits''. (orig./GL)

  5. Interacting adiabatic quantum motor

    Bruch, Anton; Kusminskiy, Silvia Viola; Refael, Gil; von Oppen, Felix

    2018-05-01

    We present a field-theoretic treatment of an adiabatic quantum motor. We explicitly discuss a motor called the Thouless motor which is based on a Thouless pump operating in reverse. When a sliding periodic potential is considered to be the motor degree of freedom, a bias voltage applied to the electron channel sets the motor in motion. We investigate a Thouless motor whose electron channel is modeled as a Luttinger liquid. Interactions increase the gap opened by the periodic potential. For an infinite Luttinger liquid the coupling-induced friction is enhanced by electron-electron interactions. When the Luttinger liquid is ultimately coupled to Fermi liquid reservoirs, the dissipation reduces to its value for a noninteracting electron system for a constant motor velocity. Our results can also be applied to a motor based on a nanomagnet coupled to a quantum spin Hall edge.

  6. Organic coolant in Winnipeg riverbed sediments

    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)

  7. Development of model pump for establishing hydraulic design of primary sodium pumps in PFBR

    Chougule, R.J.; Sahasrabudhe, H.G.; Rao, A.S.L.K.; Balchander, K.; Kale, R.D.

    1994-01-01

    Indira Gandhi Centre for Atomic Research, Kalpakkam indicated requirement of indigenous development of primary sodium pump, handling liquid sodium as coolant in Fast Breeder Reactor. The primary sodium pump concept selected in its preliminary design is a vertical, single stage, with single suction impeller, suction facing downwards. The pump is having diffuser, discharge casing and discharge collector. The 1/3 rd size model pump is developed to establish the hydraulic performance of the prototype primary sodium pump. The main objectives were to verify the hydraulic design to operate on low net positive suction head available (NPSHA), no evidence of visible cavitation at available NPSHA, the pump should be designed with a diffuser etc. The model pump PSP 250/40 was designed and successfully developed by Research and Development Division of M/s Kirloskar Brothers Ltd., Kirloskarvadi. The performance testing using model pump was successfully carried out on a closed circuit test rig. The performance of a model pump at three different speeds 1900 rpm, 1456 rpm and 975 rpm was established. The values of hydraulic axial thrust with and without balancing holes on impeller at 1900 rpm was measured. Visual cavitation study at 1900 rpm was carried out to establish the NPSH at bubble free operation of the pump. The tested performance of the model pump is converted to the full scale prototype pump. The predicted performance of prototype pump at 700 rpm was found to be meeting fully with the expected duties. (author). 6 figs., 3 tabs

  8. Heat pumps

    Macmichael, DBA

    1988-01-01

    A fully revised and extended account of the design, manufacture and use of heat pumps in both industrial and domestic applications. Topics covered include a detailed description of the various heat pump cycles, the components of a heat pump system - drive, compressor, heat exchangers etc., and the more practical considerations to be taken into account in their selection.

  9. Application of the voltage biased digital relay for the optimal protection of high inertia drive induction motors

    Choi, D. B.

    1999-01-01

    This paper describes typical protection schemes for large-size high inertia drive motor that are generally rotor thermal limited. Difficult and variable starting conditions of the large-size high inertia drive motor and compromises in the selection and setting of the protective devices are frequently encountered. The motors that typically encounter severe starting duty and present difficulties in achieving full motor protection are reactor coolant pumps (RCPs), blowers and compressors. For difficult starting conditions that are encountered by the large-size high inertia drive motors, state-of-the-art computer based calculations are capable of providing realistic predictions of the band of margin available for applying the protective relay. Based on the analysis of starting characteristics of large-size high inertia drive motors, this paper recommends that the optimal protection scheme for high inertia drive motors for nuclear power plants can be achieved by using the voltage biased digital relay instead of a speed switch and conventional overcurrent relays. (author)

  10. A system for cooling electronic elements with an EHD coolant flow

    Tanski, M; Kocik, M; Barbucha, R; Garasz, K; Mizeraczyk, J; Kraśniewski, J; Oleksy, M; Hapka, A; Janke, W

    2014-01-01

    A system for cooling electronic components where the liquid coolant flow is forced with ion-drag type EHD micropumps was tested. For tests we used isopropyl alcohol as the coolant and CSD02060 diodes in TO-220 packages as cooled electronic elements. We have studied thermal characteristics of diodes cooled with EHD flow in the function of a coolant flow rate. The transient thermal impedance of the CSD02060 diode cooled with 1.5 ml/min EHD flow was 7.8°C/W. Similar transient thermal impedance can be achieved by applying to the diode a large RAD-A6405A/150 heat sink. We found out that EHD pumps can be successfully applied for cooling electronic elements.

  11. Nonazeotropic Heat Pump

    Ealker, David H.; Deming, Glenn

    1991-01-01

    Heat pump collects heat from water circulating in heat-rejection loop, raises temperature of collected heat, and transfers collected heat to water in separate pipe. Includes sealed motor/compressor with cooling coils, evaporator, and condenser, all mounted in outer housing. Gradients of temperature in evaporator and condenser increase heat-transfer efficiency of vapor-compression cycle. Intended to recover relatively-low-temperature waste heat and use it to make hot water.

  12. Heat pumps

    Brodowicz, Kazimierz; Wyszynski, M L; Wyszynski

    2013-01-01

    Heat pumps and related technology are in widespread use in industrial processes and installations. This book presents a unified, comprehensive and systematic treatment of the design and operation of both compression and sorption heat pumps. Heat pump thermodynamics, the choice of working fluid and the characteristics of low temperature heat sources and their application to heat pumps are covered in detail.Economic aspects are discussed and the extensive use of the exergy concept in evaluating performance of heat pumps is a unique feature of the book. The thermodynamic and chemical properties o

  13. Flow boiling test of GDP replacement coolants

    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

  14. Continuous surveillance of reactor coolant circuit integrity

    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

  15. Economic study of a water pumping system for traveling gun machine; Estudo economico de uma adutora, para sistema autopropelido de irrigacao, operando com motor diesel e alcool

    Faria, Lessandro Coll; Viola, Marcelo R.; Bilibios, Carolina; Oliveira, Henrique F.E. de; Lima Junior, Joaquim A. de [Universidade Federal de Lavras (UFLA), Lavras, MG (Brazil)

    2009-07-01

    This study aimed at comparing the set-up and operation costs of a water pumping system using engines which run on two different fuels (diesel oil and alcohol) for application in traveling gun machines. The study was carried out by considering fuel prices in Minas Gerais State on 10/11/2008. The results showed that annual costs with diesel oil engines (R$ 67,460.64) were less than those for alcohol engines (R$ 82,608.12). Nevertheless, it is worthwhile to point out that the operation with alcohol engines was found to be viable only when this fuel price does not exceed R$ 1.40 per liter. (author)

  16. Numerical FEM Analyses of primary coolant system at NPP Temelin

    Junek, L.; Slovacek, M.; Ruzek, L.; Moulis, P.

    2003-01-01

    The main goal of this paper is to inform about the beginning and first steps of implementation of an aging management system at the Temelin NPP. The aging management system is important not only for achieving the current safety level but also for reaching operational reliability of a production unit equipment above the life time assumed by the original design, typically over 40 years. A method to locate the most prominent degradation regions is described. A global shell model of the primary coolant system including all loops and their components - reactor pressure vessel (RPV), steam generator (SG), main coolant pump (MCP), pressurizer, feed water and steam pipelines system is presented. The results of stress-strain analysis on the measured service parameters base are given. Validation of the results is very important and the method to compare the service measurement data with the numerical results is described. The global/local approach is mentioned and discussed. The effects of the complete global system on the individual components under monitoring are transformed into more accurate local spatial models. The local spatial models are used to analyze the gradual lifetime exhaustion of a facility during its service operation. Two spatial local models are presented, viz. feed water nozzle of SG and main coolant piping system T-brunch. The results of analysis of the local spatial models are processed by the neural network computing method, which is also described. The actual gradual damage of the material of the components under monitoring can be obtained based on the analyses performed and on the results from the neural network in combination with the knowledge of the real material characteristics. The procedures applied are included in the DIALIFE diagnostic system

  17. Evaluation of stress histories of reactor coolant loop piping for pipe rupture prediction

    Lu, S.C.; Larder, R.A.; Ma, S.M.

    1981-01-01

    This paper describes the analyses used to evaluate stress histories in the primary coolant loop piping of a selected four-loop PNR power station. In order to make the simulation as realistic as possible, best estimates rather than conservative assumptions were considered throughout. The best estimate solution, however, was aided by a sensitivity study to assess the possible variation of outcomes resulted from uncertainties associated with these assumptions. Sources of stresses considered in the evaluation were pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, pump vibrations, and finally seismic excitations. The best estimates of pressure and thermal transient histories arising from plant operations were based on actual plant operation records supplemented by specified plant design conditions. Seismic motions were generated from response spectrum curves developed specifically for the region surrounding the plant site. Stresses due to dead weight and thermal expansion were computed from a three dimensional finite element model which used a combination of pipe, truss, and beam elements to represent the coolant loop piping, the pressure vessel, coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients were obtained by closed form solutions. Seismic stress calculations considered the soil structure interaction, the coupling effect between the containment structure and the reactor coolant system. A time history method was employed for the seismic analysis. Calculations of residual stresses accounted for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation were estimated by a dynamic analysis using existing measurements of pump vibrations. (orig./HP)

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

    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)

  19. Consideration of hot channel factors in design for providing operating margins on coolant channel outlet temperature

    Sharma, V.K.; Surendar, C.; Bapat, C.N.

    1994-01-01

    The Indian Pressurized Heavy Water Reactors (IPHWR) are horizontal pressure tube reactors using natural uranium oxide fuel in the form of short (495 mm) clusters. The fuel clusters in the Zr-Nb pressure tubes are cooled by high pressure, high temperature and subcooled circulating heavy water. Coolant flow distribution to individual channels is designed to match the power distribution so as to obtain uniform coolant outlet temperature. However, during operation, the coolant outlet temperature in individual channels deviate from their nominal value due to: tolerances in process design; effects of grid frequency on the pump speed; deviation in channel powers from the nominal values due to on-power fuelling and movement of reactivity devices, and so on. Thus an operating margin, between the highest permissible and nominal coolant outlet temperatures, is required taking into account various hot channel factors that contribute to higher coolant outlet temperatures. The paper discusses the methodology adopted to assess various hot channel factors which would provide optimum operating margins while ensuring sub-cooling. (author)

  20. Heat transfer properties of organic coolants containing high boiling residues

    Debbage, A.G.; Driver, M.; Waller, P.R.

    1964-01-01

    Heat transfer measurements were made in forced convection with Santowax R, mixtures of Santowax R and pyrolytic high boiling residue, mixtures of Santowax R and CMRE Radiolytic high boiling residue, and OMRE coolant, in the range of Reynolds number 10 4 to 10 5 . The data was correlated with the equation Nu = 0.015 Re b 0.85 Pr b 0.4 with an r.m.s. error of ± 8.5%. The total maximum error arising from the experimental method and inherent errors in the physical property data has been estimated to be less than ± 8.5%. From the correlation and physical property data, the decrease in heat transfer coefficient with increasing high boiling residue concentration has been determined. It has been shown that subcooled boiling in organic coolants containing high boiling residues is a complex phenomenon and the advantages to be gained by operating a reactor in this region may be marginal. Gas bearing pumps used initially in these experiments were found to be unsuitable; a re-designed ball bearing system lubricated with a terphenyl mixture was found to operate successfully. (author)

  1. Small break LOCA [loss of coolant accident] mitigation for Bellefonte

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

    1986-01-01

    Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S 2 D sequence. Operator actions to mitigate the S 2 D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S 2 D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were also able to prevent core damage for the S 2 D sequence

  2. Coolant processing device for nuclear reactor

    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.)

  3. Fatigue management considering LWR coolant environments

    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)

  4. Selection of nuclear reactor coolant materials

    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)

  5. Standardized sampling system for reactor coolants

    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

  6. Centrifugal pumps

    Gülich, Johann Friedrich

    2014-01-01

    This book gives an unparalleled, up-to-date, in-depth treatment of all kinds of flow phenomena encountered in centrifugal pumps including the complex interactions of fluid flow with vibrations and wear of materials. The scope includes all aspects of hydraulic design, 3D-flow phenomena and partload operation, cavitation, numerical flow calculations, hydraulic forces, pressure pulsations, noise, pump vibrations (notably bearing housing vibration diagnostics and remedies), pipe vibrations, pump characteristics and pump operation, design of intake structures, the effects of highly viscous flows, pumping of gas-liquid mixtures, hydraulic transport of solids, fatigue damage to impellers or diffusers, material selection under the aspects of fatigue, corrosion, erosion-corrosion or hydro-abrasive wear, pump selection, and hydraulic quality criteria. As a novelty, the 3rd ed. brings a fully analytical design method for radial impellers, which eliminates the arbitrary choices inherent to former design procedures. The d...

  7. A Linear Electromagnetic Piston Pump

    Hogan, Paul H.

    Advancements in mobile hydraulics for human-scale applications have increased demand for a compact hydraulic power supply. Conventional designs couple a rotating electric motor to a hydraulic pump, which increases the package volume and requires several energy conversions. This thesis investigates the use of a free piston as the moving element in a linear motor to eliminate multiple energy conversions and decrease the overall package volume. A coupled model used a quasi-static magnetic equivalent circuit to calculate the motor inductance and the electromagnetic force acting on the piston. The force was an input to a time domain model to evaluate the mechanical and pressure dynamics. The magnetic circuit model was validated with finite element analysis and an experimental prototype linear motor. The coupled model was optimized using a multi-objective genetic algorithm to explore the parameter space and maximize power density and efficiency. An experimental prototype linear pump coupled pistons to an off-the-shelf linear motor to validate the mechanical and pressure dynamics models. The magnetic circuit force calculation agreed within 3% of finite element analysis, and within 8% of experimental data from the unoptimized prototype linear motor. The optimized motor geometry also had good agreement with FEA; at zero piston displacement, the magnetic circuit calculates optimized motor force within 10% of FEA in less than 1/1000 the computational time. This makes it well suited to genetic optimization algorithms. The mechanical model agrees very well with the experimental piston pump position data when tuned for additional unmodeled mechanical friction. Optimized results suggest that an improvement of 400% of the state of the art power density is attainable with as high as 85% net efficiency. This demonstrates that a linear electromagnetic piston pump has potential to serve as a more compact and efficient supply of fluid power for the human scale.

  8. Hybrid Switch Reluctance Drives For Pump Applications

    Jakobsen, Uffe

    be the single phase hybrid switched reluctance motor (HSRM). Due to the simple construction of the single phase HSRM, manufacturing may be simplified compared to a three phase permanent magnet motor and consumption of copper may be lowered when compared to both the induction motor and some three phase permanent...... magnet synchronous motor (PMSM), the HSRM needs information about rotor position to be properly controlled. For BLDC, PMSM, induction motors, and the normal SRM position sensorless methods are relatively well established and have been used for some time. For the single phase switched reluctance motor......The initial research problem is to investigate an alternative motor drive to the existing permanent magnet synchronous and brushless DC-motor drives for pump applications. A review of different motor types showed that a possible candidate for another low cost permanent magnet motor may...

  9. Pumping life

    Sitsel, Oleg; Dach, Ingrid; Hoffmann, Robert Daniel

    2012-01-01

    The name PUMPKIN may suggest a research centre focused on American Halloween traditions or the investigation of the growth of vegetables – however this would be misleading. Researchers at PUMPKIN, short for Centre for Membrane Pumps in Cells and Disease, are in fact interested in a large family o......’. Here we illustrate that the pumping of ions means nothing less than the pumping of life....

  10. Design capability of CANDU heat transport pump shafts against cracking

    Kumar, A.N.; Sheikh, Z.B.; Padgett, A.

    1993-01-01

    During 1986 three different Light Water Reactors (LWR's) in the U.S. reported either a cracked or fractured shaft on one or more of their reactor coolant (RC) pumps. The RC pumps for all these stations were supplied by Byron Jackson (BJ) Pump Company. A majority of CANDU heat transport (HT) pumps (equivalent of RC pumps) are supplied by BJ Pump Company and are similar in design to RC pumps. Hence the failure of these RC pumps in the U.S. utilities caused concern regarding the relevance of these failures to the BJ supplied CANDU HT pumps (HTP). This paper presents the results of AECL assessment to establish the capability of the HT pump shaft against cracking. Two methods were used for assessment: (a) detailed comparative design review of the HTP and RCP shafts; (b) semi-empirical analysis of the HTP shafts. The results of the AECL assessment showed significant differences in detailed design, materials, assembly and fits of various components and the control of operating parameters between the HT and RC pumps. It was concluded that because of these differences the failures similar to RC pump shafts are not likely to appear in HT pump shafts. This conclusion is further reinforced by about 140,000 hours of operating history of the longest running HT pump of comparable size to RC Pumps, without failures

  11. Methodology to monitor and diagnostic vibrations of the motor-pumps used in the primary cooling system of IEAR-1 nuclear research reactor; Metodologia para monitoracao e diagnostico de vibracao das bombas moto-operadas do circuito primario de refrigeracao do Reator IEA-R1

    Benevenuti, Erion de Lima

    2004-07-01

    The objectives of this study are to establish a strategy to monitor and diagnose vibrations of the motor pumps used in the primary reactor cooling system of the IEA-R1 nuclear research reactor, to verify the possibility of using the existing installed monitoring vibration system and to implement such strategy in a continuous way. Four types of mechanical problems were considered: unbalancing, misalignment, gaps and faults in bearings. An adequate set of analysis tools, well established by the industry, was selected. These are: global measurements of vibration, velocity spectrum and acceleration envelope spectrum. Three sources of data and information were used; the data measured from the primary pumps, experimental results obtained with a Spectra Quest machine used to simulate mechanical defects and data from the literature. The results show that, for the specific case of the motor-pumps of IEA-R1 nuclear research reactor, although the technique using the envelope of acceleration, which is not available in the current system used to monitor the vibration of the motor pumps, is the one with best performance, the other techniques available in the system are sufficient to monitor the four types of mechanical problems mentioned. The proposed strategy is shown and detailed in this work. (author)

  12. Walking beam pumping unit system efficiency measurements

    Kilgore, J.J.; Tripp, H.A.; Hunt, C.L. Jr.

    1991-01-01

    The cost of electricity used by walking beam pumping units is a major expense in producing crude oil. However, only very limited information is available on the efficiency of beam pumping systems and less is known about the efficiency of the various components of the pumping units. This paper presents and discusses measurements that have been made on wells at several Shell locations and on a specially designed walking beam pump test stand at Lufkin Industries. These measurements were made in order to determine the overall system efficiency and efficiency of individual components. The results of this work show that the overall beam pumping system efficiency is normally between 48 and 58 percent. This is primarily dependent on the motor size, motor type, gearbox size, system's age, production, pump size, tubing size, and rod sizes

  13. Modular 3-D solid finite element model for fatigue analyses of a PWR coolant system

    Garrido, Oriol Costa; Cizelj, Leon; Simonovski, Igor

    2012-01-01

    Highlights: ► A 3-D model of a reactor coolant system for fatigue usage assessment. ► The performed simulations are a heat transfer and stress analyses. ► The main results are the expected ranges of fatigue loadings. - Abstract: The extension of operational licenses of second generation pressurized water reactor (PWR) nuclear power plants depends to a large extent on the analyses of fatigue usage of the reactor coolant pressure boundary. The reliable estimation of the fatigue usage requires detailed thermal and stress analyses of the affected components. Analyses, based upon the in-service transient loads should be compared to the loads analyzed at the design stage. The thermal and stress transients can be efficiently analyzed using the finite element method. This requires that a 3-D solid model of a given system is discretized with finite elements (FE). The FE mesh density is crucial for both the accuracy and the cost of the analysis. The main goal of the paper is to propose a set of computational tools which assist a user in a deployment of modular spatial FE model of main components of a typical reactor coolant system, e.g., pipes, pressure vessels and pumps. The modularity ensures that the components can be analyzed individually or in a system. Also, individual components can be meshed with different mesh densities, as required by the specifics of the particular transient studied. For optimal accuracy, all components are meshed with hexahedral elements with quadratic interpolation. The performance of the model is demonstrated with simulations performed with a complete two-loop PWR coolant system (RCS). Heat transfer analysis and stress analysis for a complete loading and unloading cycle of the RCS are performed. The main results include expected ranges of fatigue loading for the pipe lines and coolant pump components under the given conditions.

  14. Study of motor-pump operating as generator groups in hydroelectric micro hydropower plants; Estudo de grupos moto-bomba operando como grupos geradores em microcentrais hidreletricas

    Ricardo, Mateus

    2007-07-01

    The present work aims to evaluate the in site use of pumps working as turbines (PATs) and induction engines operating as generators (MIGs), for the replacement of conventional generating groups in micro hydropower plants (MHPs), focusing on PATs. To achieve this goal, a case study was accomplished on a micro hydropower plant installed in Mantiqueira Mountain that, after been rebuilt, received a PAT/MIG generating group. It's a pioneer research, being the first documented case of implementation of PATs and MIGs in a real MHP in Brazil. After tests for defining the efficiency, and the comparison with theoretical and laboratory results, it was concluded that PAT/MIG generating groups can be used effectively in electricity generation, by the application of appropriate selection methodologies and adjustments to optimize its operation in reverse. Besides the technical aspects, it was also addressed the technology's economic aspects. Through prices research and costs composition, it was possible to quantify the economic benefit of PAT/MIG groups over conventional generating groups, noting that the studied technology reduces the micro hydropower plants deployment costs in a significant way . (author)

  15. Study of motor-pump operating as generator groups in hydroelectric micro hydropower plants; Estudo de grupos moto-bomba operando como grupos geradores em microcentrais hidreletricas

    Ricardo, Mateus

    2007-07-01

    The present work aims to evaluate the in site use of pumps working as turbines (PATs) and induction engines operating as generators (MIGs), for the replacement of conventional generating groups in micro hydropower plants (MHPs), focusing on PATs. To achieve this goal, a case study was accomplished on a micro hydropower plant installed in Mantiqueira Mountain that, after been rebuilt, received a PAT/MIG generating group. It's a pioneer research, being the first documented case of implementation of PATs and MIGs in a real MHP in Brazil. After tests for defining the efficiency, and the comparison with theoretical and laboratory results, it was concluded that PAT/MIG generating groups can be used effectively in electricity generation, by the application of appropriate selection methodologies and adjustments to optimize its operation in reverse. Besides the technical aspects, it was also addressed the technology's economic aspects. Through prices research and costs composition, it was possible to quantify the economic benefit of PAT/MIG groups over conventional generating groups, noting that the studied technology reduces the micro hydropower plants deployment costs in a significant way . (author)

  16. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 3: nonseismic stress analysis. Final report

    Chan, A.L.; Curtis, D.J.; Rybicki, E.F.; Lu, S.C.

    1981-08-01

    This volume describes the analyses used to evaluate stresses due to loads other than seismic excitations in the primary coolant loop piping of a selected four-loop pressurized water reactor nuclear power station. The results of the analyses are used as input to a simulation procedure for predicting the probability of pipe fracture in the primary coolant system. Sources of stresses considered in the analyses are pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, and mechanical vibrations. Pressure and thermal transients arising from plant operations are best estimates and are based on actual plant operation records supplemented by specified plant design conditions. Stresses due to dead weight and thermal expansion are computed from a three-dimensional finite element model that uses a combination of pipe, truss, and beam elements to represent the reactor coolant loop piping, reactor pressure vessel, reactor coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients are obtained by closed-form solutions. Calculations of residual stresses account for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation are estimated by a dynamic analysis using existing measurements of pump vibrations

  17. The role of gastroesophageal reflux in relation to symptom onset in patients with proton pump inhibitor-refractory nonerosive reflux disease accompanied by an underlying esophageal motor disorder.

    Izawa, Shinya; Funaki, Yasushi; Iida, Akihito; Tokudome, Kentaro; Tamura, Yasuhiro; Ogasawara, Naotaka; Sasaki, Makoto; Kasugai, Kunio

    2014-01-01

    The symptom improvement rate is low with proton pump inhibitors (PPIs) in nonerosive reflux disease (NERD). The underlying pathogenic mechanism is complex. Esophageal motility disorders (EMDs) are thought to be a factor, but their prevalence, type, symptoms and the role played by gastroesophageal reflux (GER) in symptom onset have not been fully investigated. To investigate the role of GER in symptom onset in PPI-refractory NERD patients with EMDs. This study comprised 76 patients with PPI-refractory NERD. Manometry was performed during PPI treatment and patients were divided into an EMD group and normal motility (non-EMD) group. Then, multichannel intraluminal impedance-pH monitoring was performed and medical interviews were conducted. Nineteen patients (25%) had an EMD. Data were compared between 17 patients, excluding 2 with achalasia and 57 non-EMD patients. No significant differences were observed between groups in 24-hour intraesophageal pH <4 holding time (HT), mean number of GER episodes or mean number of proximal reflux episodes. The reflux-related symptom index (≥50%) showed a relationship between reflux and symptoms in 70.5% of EMD patients and 75% of non-EMD patients. In the EMD group, the score for FSSG (Frequency Scale for the Symptoms of GERD) question (Q)10 was significantly correlated with the number of GER episodes (r = 0.58, p = 0.02) and the number of proximal reflux episodes (r = 0.63, p = 0.02). In addition, the score for Q9 tended to be correlated with the number of GER episodes (r = 0.44, p = 0.06). Our results suggest that some PPI-refractory NERD patients have EMDs, and that GER plays a role in symptom onset.

  18. Analysis of the core reflooding of a PWR reactor under a loss-of-coolant postulated accident

    Austregesilo Filho, H.

    1978-12-01

    The main purpose of this work is to analyse the termohydraulic behaviour of emergency cooling water, during reflooding of a PWR core submitted to a postulated loss-of-coolant accident, with the scope of giving the boundary conditions needed to verify fuel element and containment integrity. The analytical model presented was applied to the simulation of Angra I core reflooding phase, after a double-ended break between pressure vessel and discharge of one of the main coolant pumps. For this accident, with a discharge coefficient of C sub(D) = 0.4, the highest peak cladding temperature is expected. (author) [pt

  19. Nuclear reactor of pressurized liquid coolant type

    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

  20. ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

    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.

  1. Iron crud supply device to reactor coolant

    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.)

  2. Limits to fuel/coolant mixing

    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

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

    S. K. Pavlov

    2014-01-01

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

  4. 49 CFR 178.337-15 - Pumps and compressors.

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Pumps and compressors. 178.337-15 Section 178.337... PACKAGINGS Specifications for Containers for Motor Vehicle Transportation § 178.337-15 Pumps and compressors. (a) Liquid pumps or gas compressors, if used, must be of suitable design, adequately protected...

  5. 49 CFR 178.338-17 - Pumps and compressors.

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Pumps and compressors. 178.338-17 Section 178.338... PACKAGINGS Specifications for Containers for Motor Vehicle Transportation § 178.338-17 Pumps and compressors. (a) Liquid pumps and gas compressors, if used, must be of suitable design, adequately protected...

  6. Liquid metals pumping

    Le Frere, J.P.

    1984-01-01

    Pumps used to pump liquid metals depend on the liquid metal and on the type of application concerned. One deals more particularly with electromagnetic pumps, the main pumps used with mechanical pumps. To pump sodium in the nuclear field, these two types of pumps are used; the pumps of different circuits of Super Phenix are presented and described [fr

  7. Electrokinetic pump

    Patel, Kamlesh D.

    2007-11-20

    A method for altering the surface properties of a particle bed. In application, the method pertains particularly to an electrokinetic pump configuration where nanoparticles are bonded to the surface of the stationary phase to alter the surface properties of the stationary phase including the surface area and/or the zeta potential and thus improve the efficiency and operating range of these pumps. By functionalizing the nanoparticles to change the zeta potential the electrokinetic pump is rendered capable of operating with working fluids having pH values that can range from 2-10 generally and acidic working fluids in particular. For applications in which the pump is intended to handle highly acidic solutions latex nanoparticles that are quaternary amine functionalized can be used.

  8. Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

    2003-01-01

    In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

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

    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)

  10. On-Line Coolant Chemistry Analysis

    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

  11. Leak detection device for reactor coolant

    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.)

  12. A dynamic model of the reactor coolant system flow for KMRR plant simulation

    Rhee, B.W.; Noh, T.W.; Park, C.; Sim, B.S.; Oh, S.K.

    1990-01-01

    To support computer simulation studies for reactor control system design and performance evaluation, a dynamic model of the reactor coolant system (RCS) and reflector cooling system has been developed. This model is composed of the reactor coolant loop momentum equation, RCS pump dynamic equation, RCS pump characteristic equation, and the energy equation for the coolant inside the various components and piping. The model is versatile enough to simulate the normal steady-state conditions as well as most of the anticipated flow transients without pipe rupture. This model has been successfully implemented as the plant simulation code KMRRSIM for the Korea Multi-purpose Research Reactor and is now under extensive validation testing. The initial stage of validation has been comparison of its result with that of already validated, more detailed reactor system transient codes such as RELAP5. The results, as compared to the predictions by RELAP5 simulation, have been generally found to be very encouraging and the model is judged to be accurate enough to fulfill its intended purpose. However, this model will continue to be validated against other plant's data and eventually will be assessed by test data from KMRR

  13. An evaluation of debris mobility within a PWR reactor coolant system during the recirculation mode

    Andreychek, T.S.

    1987-01-01

    To provide for the long-term cooling of the nuclear core of a Pressurized Water Rector (PWR) following a hypothetical Loss-of-Coolant Accidnet (LOCA), water is drawn from the containment sump and pumped into the reactor coolant system (RCS). It has been postulated that debris from the containment, such as dirt, sand, and paint from containment walls and in-containment equipment, could be carried into the containment sump due to the action of the RCS coolant that escapes from the breach in the piping and then flows to the sump. Once in the sump, this debris could be pumped into the Safety Injection System (SIS) and ultimately the RCS itself, causing the performance of the SIS to be degraded. Of particular interest is the potential for core blockage that may occur due to debris transport into the core region by the recirculating flow. This paper presents a method of evaluating the potential for debris from the sump to form core blockages under recirculating flow conditions following a hypothetical LOCA for a PWR

  14. Evaluation of Coolant Injection Procedure in the Severe Accident Management Strategy of APR1400

    Cho, Yongjin; Lim, Kukhee; Song, Sungchu; Lee, Sukho; Hwang, Taesuk

    2013-01-01

    A coolant injection strategy in the severe accident management guideline (SAMG) of APR1400 relates to immediate coolant injection into RCS (Reactor Coolant System) or injection following the recovery of secondary coolant inventory. This strategy could play important role in accident mitigation and radiological consequences. In this study, appropriateness of the strategy was evaluated using MELCOR1.8.6 and several sensitivity studies of the key parameters were performed. Analysis for APR1400 using MELCOR 1.8.6 was performed to evaluate the effectiveness of accident management strategies and the following conclusions were identified. Sequential operation of secondary and RCS injection may not be the best strategy and the simultaneous injection of secondary and RCS injection could be more preferable. At least, the RCS injection should start before complete drainage of water in the safety injection tank using mobile pumps. In this study, the effectiveness of timing of operator action has been examined and the amount of injection flowrate needs to be studied in the future

  15. Phenomena occuring in the reactor coolant system during severe core damage accidents

    Malinauskas, A.P.

    1990-01-01

    The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. This paper discusses, how in the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. The physical and chemical processes occurring within the RCS during normal operation of the reactor are relatively uncomplicated and are reasonably well understood. When the flow of coolant is properly adjusted, the thermal energy resulting from nuclear fission (or, in the shutdown mode, from radioactive decay processes) and secondary inputs, such as pumps, are exactly balanced by thermal losses through the RCS boundaries and to the various heat sinks that are employed to effect the conversion of heat to electrical energy. Because all of the heat and mass fluxes remain sensibly constant with time, mathematical descriptions of the thermophysical processes are relatively straightforward, even for boiling water reactor (BWR) systems. Although the coolant in a BWR does undergo phase changes, the phase boundaries remain well-defined and time-invariant

  16. Conceptual design of the integral test loop (I): Reactor coolant system and secondary system

    Song, Chul Hwa; Lee, Seong Je; Kwon, Tae Soon; Moon, Sang Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    This report describes the conceptual design of the primary coolant system and the secondary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the primary coolant system and secondary system is as follows ; Reference plant: Korean Standard Nuclear Plant (KSNP), Height ratio : 1/1, Volume ratio : 1/200, Power scale : Max. 15% of the scaled nominal power, Temperature, Pressure : Real plant conditions. The primary coolant system includes a reactor vessel, which contains a core simulator, a steam generator, a reactor coolant pump simulator, a pressurizer and piping, which consists of two hot legs, four cold legs and four intermediate legs. The secondary system consists of s steam discharge system, a feedwater supply system and a steam condensing system. This conceptual design report describes general configuration of the reference plant, and major function and operation of each system of the plant. Also described is the design philosophy of each component and system of the ITL, and specified are the design criteria and technical specifications of each component and system of the ITL in the report. 17 refs., 43 figs., 51 tabs. (Author)

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

    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)

  18. Reciprocating Linear Electric Motor

    Goldowsky, M. P.

    1984-01-01

    Features include structural simplicity and good force/displacement characteristics. Reciprocating motor has simple, rugged construction, relatively low reciprocating weight, improved power delivery, and improved force control. Wear reduced by use of magnetic bearings. Intended to provide drivers for long-lived Stirling-cycle cryogenic refrigerators, concept has less exotic applications, such as fuel pumps.

  19. 49 CFR 178.348-3 - Pumps, piping, hoses and connections.

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Pumps, piping, hoses and connections. 178.348-3... FOR PACKAGINGS Specifications for Containers for Motor Vehicle Transportation § 178.348-3 Pumps, piping, hoses and connections. Each pump and all piping, hoses and connections on each cargo tank motor...

  20. Pump system characterization and reliability enhancement

    Staunton, R.H.

    1998-01-01

    Pump characterization studies were performed at the Oak Ridge National Laboratory (ORNL) to review and analyze six years (1990-1995) of data from pump systems at domestic nuclear plants. The studies considered not only pumps and pump motors but also pump-related circuit breakers and turbine drives (i.e., the pump system). One significant finding was that the number of 'significant' failures of the pump circuit breaker exceeds the number of significant failures of the pump itself. The study also shows how regulatory code testing was designed for the pump only and therefore did not lead to the discovery of other significant pump system failures. Potential diagnostic technologies, both experimental and mature, suitable for on-line and off-line pump testing were identified. The study does not select or recommend technologies but proposes diagnostic technologies and monitoring techniques that should be further evaluated/developed for making meaningful and critically-needed improvements in the reliability of the pump system. (author)

  1. Pumps for nuclear power stations

    Ogura, Shiro

    1979-01-01

    16 nuclear power plants are in commercial operation in Japan, and nuclear power generation holds the most important position among various substitute energies. Hereafter also, it is expected that the construction of nuclear power stations will continue because other advantageous energy sources are not found. In this paper, the outline of the pumps used for BWR plants is described. Nuclear power stations tend to be large scale to reduce the construction cost per unit power output, therefore the pumps used are those of large capacity. The conditions to be taken in consideration are high temperature, high pressure, radioactive fluids, high reliability, hydrodynamic performances, aseismatic design, relevant laws and regulations, and quality assurance. Pumps are used for reactor recirculation system, control rod driving hydraulic system, boric acid solution injecting system, reactor coolant purifying system, fuel pool cooling and purifying system, residual heat removing system, low pressure and high pressure core spraying systems, and reactor isolation cooling system, for condensate, feed water, drain and circulating water systems of turbines, for fresh water, sea water, make-up water and fire fighting services, and for radioactive waste treating system. The problems of the pumps used for nuclear power stations are described, for example, the requirement of high reliability, the measures to radioactivity and the aseismatic design. (Kako, I.)

  2. Pumps and pump facilities. 2. ed.

    Bohl, W.; Bauerfeind, H.; Gutmann, G.; Leuschner, G.; Matthias, H.B.; Mengele, R.; Neumaier, R.; Vetter, G.; Wagner, W.

    1981-01-01

    This book deals with the common fundamental aspects of liquid pumps and gives an exemplary choice of the most important kinds of pumps. The scientific matter is dealt with by means of practical mathematical examples among other ways of presenting the matter. Survey of contents: Division on main operational data of pumps - pipe characteristics - pump characteristics - suction behaviour of the pumps - projecting and operation of rotary pumps - boiler feed pumps - reactor feed pumps - oscillating positive-displacement pumps - eccentric spiral pumps. (orig./GL) [de

  3. Four-quadrant characteristics of Psb-VVER pumps

    Lipatov, I.A.; Elkin, I.V.; Antonova, A.I.; Dremin, G.I.; Kapustin, A.V.; Nikonov, S.M.; Rovnov, A.A.; Gudkov, V.I.

    2005-01-01

    This paper represents description of determination of Tunis-1620 pump head and torque characteristics of the integral thermophysical test facility Psb-Ver, obtained for single-phase coolant. Test procedure and main results obtained are described in the paper. (author)

  4. Features of electric drive sucker rod pumps for oil production

    Gizatullin, F. A.; Khakimyanov, M. I.; Khusainov, F. F.

    2018-01-01

    This article is about modes of operation of electric drives of downhole sucker rod pumps. Downhole oil production processes are very energy intensive. Oil fields contain many oil wells; many of them operate in inefficient modes with significant additional losses. Authors propose technical solutions to improve energy performance of a pump unit drives: counterweight balancing, reducing of electric motor power, replacing induction motors with permanent magnet motors, replacing balancer drives with chain drives, using of variable frequency drives.

  5. Development and Optimized Design of Propeller Pump System & Structure with VFD in Low-head Pumping Station

    Rentian, Zhang; Honggeng, Zhu; Arnold, Jaap; Linbi, Yao

    2010-06-01

    Compared with vertical-installed pumps, the propeller (bulb tubular) pump systems can achieve higher hydraulic efficiencies, which are particularly suitable for low-head pumping stations. More than four propeller pumping stations are being, or will be built in the first stage of the S-to-N Water Diversion Project in China, diverting water from Yangtze River to the northern part of China to alleviate water-shortage problems and develop the economy. New structures of propeller pump have been developed for specified pumping stations in Jiangsu and Shandong Provinces respectively and Variable Frequency Drives (VFDs) are used in those pumping stations to regulate operating conditions. Based on the Navier-Stokes equations and the standard k-e turbulent model, numerical simulations of the flow field and performance prediction in the propeller pump system were conducted on the platform of commercial software CFX by using the SIMPLEC algorithm. Through optimal design of bulb dimensions and diffuser channel shape, the hydraulic system efficiency has improved evidently. Furthermore, the structures of propeller pumps have been optimized to for the introduction of conventional as well as permanent magnet motors. In order to improve the hydraulic efficiency of pumping systems, both the pump discharge and the motor diameter were optimized respectively. If a conventional motor is used, the diameter of the pump casing has to be increased to accommodate the motor installed inside. If using a permanent magnet motor, the diameter of motor casing can be decreased effectively without decreasing its output power, thus the cross-sectional area is enlarged and the velocity of flowing water decreased favorably to reduce hydraulic loss of discharge channel and thereby raising the pumping system efficiency. Witness model tests were conducted after numerical optimization on specific propeller pump systems, indicating that the model system hydraulic efficiencies can be improved by 0.5%˜3.7% in

  6. Fission product release into the primary coolant

    Apperson, C.E.

    1977-01-01

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

  7. FY 1986 Report on research and development of super heat pump energy accumulation system. Part 1. Development of elementary techniques; 1986 nendo super heat pump energy shuseki system no kenkyu kaihatsu seika hokokusho. 1. Yoso gijutsu no kenkyu

    NONE

    1987-04-01

    Summarized in detail herein are R and D results of the super high performance heat pumps and elementary equipment and working fluids, for R and D of the super heat pump energy accumulation system. For R and D of the super high performance compression heat pumps, the R and D efforts are directed to development of new working fluids, high-performance heat exchangers, closed motors and so on for the highly efficient type (for heating only); to researches on mixed coolants, high-efficiency screw compressors and so on for the highly efficient type (for cooling and heating); to development of tooth shape of the screw compression section, surveys on thermal stability of the working fluids for heating and so on for the high temperature type (utilizing low temperature heat source); and to R and D of the high-speed reciprocating compressors and steam superchargers for the high temperature type (utilizing high temperature heat source). For R and D of the elementary equipment and working fluids, researches are conducted on evaporators for mixed working fluids, condensers utilizing the EHD effect, stainless steel plate fin type heat exchangers, heat exchangers for the chemical heat accumulation unit, and so on. The R and D efforts are also directed to the working fluids (alcohol-based and nonalcohol-based). (NEDO)

  8. Wind pumps for agriculture: Cost and environmental benefits (comparisons with electric and combustion engine driven pumps)

    Piccoli, F.

    1991-01-01

    After describing initial and running costs of a group of wind-pumps, the author calculates and compares, as far as agricultural and zootechnical purposes are concerned, the costs for each cubic meter of water extracted through wind-powered, electric and internal-combustion engines. The comparisons clearly show, under adequate wind conditions, that wind-pumps are economically more suitable than electric and motor pumps with similar delivery heads

  9. Interfacing systems loss of coolant accident (ISLOCA) pressure capacity methodology and Davis-Besse results

    Wesley, D.A.

    1991-01-01

    A loss of coolant accident resulting from the overpressurization by reactor coolant fluid of a system designed for low-pressure, low-temperature service has been identified as a potential contributor to nuclear power plant risk. In this paper, the methodology developed to assess the probability of failure as a function of internal pressure is presented, and sample results developed for the controlling failure modes and locations of four fluid systems at the Davis-Besse Plant are shown. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The variability in the probability of failure is included, and the estimated leak rates or leak areas are given for the controlling modes of failure. For this evaluation, all failures are based on quasistatic pressures since the probability of dynamic effects resulting from such causes as water hammer have been initially judged to be negligible for the Davis-Besse plant ISLOCA

  10. Numerical investigation of the coolant mixing during fast deboration transients for VVER-440 type reactors

    Hoehne, T.; Rhode, U.

    2000-01-01

    The VVER-440 (440 MW) V-230 was considered for analyzing the flow field and mixing processes. The V-230 has no elliptical sieve plate in the lower plenum. Previously, the 3D flow distribution in the downcomer and the lower plenum of the VVER-440 reactor have been calculated by means of the CFD code CFX-4 for operational conditions. The CFX-calculations were compared with the experimental data and the analytical mixing model. In this paper, CFD calculations for the start-up of the first main coolant pump in a VVER-440 type reactor are reported about. This scenario is important in case that there is a plug of lower borated water in one of the primary coolant loops. (orig.)

  11. Analysis Of Primary Coolant Suction Side Pressure In The Delay Chamber Of The RSG-GAS

    Dibyo, Sukmanto

    2000-01-01

    Delay chamber is a tank to delay flow that located in the primary cooling suction side of RSG-GAS. A void occurred when operation reactor caused by too high the delta P at inlet suction pump. The condition may be avoided by using one line mode of the cooling flow. The analysis show that void volume in the delay chamber is occurred because the coolant negative pressure lowers the saturation pressure should be avoided though decreasing the delta P until about 0.1 bar at about 45 exp 0 C. Solution suggested are to use bypass flow from the spent fuel to the delay chamber. Coolant temperature can be also decreased by decreasing the power level of the reactor as well as improving the heat exchanger and cooling tower performances

  12. Single failure effects of reactor coolant system large bore hydraulic snubbers for Korean Standard Nuclear Power Plant

    Choi, T.S.; Park, S.H.; Sung, K.K.; Kim, T.W.; Jheon, J.H.

    1996-01-01

    A potential snubber single failure is one of the safety significances identified in General Safety Issue 113 for Large Bore Hydraulic Snubber (LBHS) dynamic qualification. This paper investigates dynamic structural effects of single failures of the steam generator and reactor coolant pump snubbers in Korean Standard Nuclear Power Plant by performing the time history dynamic analyses for the reactor coolant system under seismic and postulated pipe break events. The seismic input motions considered are the synthesized ground time histories conforming to SRP 3.7.1, and he postulated pipe break input loadings result from steam generator main seam line and feedwater line pipe breaks which govern pipe breaks remaining after applying LBB to the main coolant line and primary side ranch lines equal to and greater than 12 inch nominal pipe size

  13. Swimming bath pumps. Ranges of application, selection, installation, energy efficiency; Schwimmbadpumpen. Einsatzbereiche, Auswahl, Aufbau, Energieeffizienz

    Korupp, Sascha

    2012-07-01

    The booklet under consideration reports on pumps being used in swimming baths. The range of application of the individual sorts of pump is described. The knowledge on the selection of suitable pumps is given. Subsequently, the design of a pump with the main components, the installation of a pump as well as the most important issues and difficulties of the pump material are illustrated. The total costs of a pump within its life cycle are determined. These life cycle costs can be reduced by propulsion components increasing the energy efficiency such as frequency converter, motors with permanent magnets or heat exchanger motors.

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

    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...

  15. Design and experimental characterization of an EM pump

    Kim, Hee Reyoung; Hong, Sang Hee

    1999-01-01

    Generally, an EM (electromagnetic) pump is been employed to circulate electrically conducting liquids by using the Lorentz force. Especially, at the liquid metal reactor (LMR), which uses liquid sodium with high electrical conductivity as a coolant, an EM pump is needed due to its advantages over a mechanical pump, such as no rotating parts, no noise, and simplicity. In this research, an EM pump of a pilot annular linear induction type with a flow rate of 200 l/min was designed by using the electrical equivalent-circuit method. The pump was designed and manufactured by considering material and environmental (high temperature and liquid sodium) requirements. The pump performance was experimentally characterized based on input currents, voltage, power, and frequency. Also, the theoretical prediction was compared with the experimental result

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

    Siamidis, John; Mason, Lee

    2006-01-01

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

  17. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

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

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

  18. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

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

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

  19. Fluidic pumps

    Priestman, G.H.

    1990-01-01

    A fluidic pump has primary and secondary vessels connected by a pipe, a displacement vessel having liquid to be delivered through a pipe via a rectifier provided with a feed tank. A drive unit delivers pressure fluid to a line to raise liquid and compress trapped gas or liquid in the space, including the pipe between the liquids in the two vessels and thus drive liquid out of the displacement vessel. The driving gas is therefore separated by the barrier liquid and the trapped gas or liquid from the liquid to be pumped which liquid could be e.g. radioactive. (author)

  20. Pumped storage

    Strauss, P.L.

    1991-01-01

    The privately financed 1,000 MW Rocky Point Pumped Storage Project located in central Colorado, USA, will be one of the world's highest head, 2,350 feet reversible pump/turbine projects. The project will offer an economical supply of peaking power and spinning reserve power to Colorado and other southwestern states. This paper describes how the project will be made compatible with the environmental conditions in the project area and the type of terrestrial mitigation measures that are being proposed for those situations where the project impacts the environment, either temporarily or permanently

  1. Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

    Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-11-15

    The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

  2. Estimation of pump operational state with model-based methods

    Ahonen, Tero; Tamminen, Jussi; Ahola, Jero; Viholainen, Juha; Aranto, Niina; Kestilae, Juha

    2010-01-01

    Pumps are widely used in industry, and they account for 20% of the industrial electricity consumption. Since the speed variation is often the most energy-efficient method to control the head and flow rate of a centrifugal pump, frequency converters are used with induction motor-driven pumps. Although a frequency converter can estimate the operational state of an induction motor without external measurements, the state of a centrifugal pump or other load machine is not typically considered. The pump is, however, usually controlled on the basis of the required flow rate or output pressure. As the pump operational state can be estimated with a general model having adjustable parameters, external flow rate or pressure measurements are not necessary to determine the pump flow rate or output pressure. Hence, external measurements could be replaced with an adjustable model for the pump that uses estimates of the motor operational state. Besides control purposes, modelling the pump operation can provide useful information for energy auditing and optimization purposes. In this paper, two model-based methods for pump operation estimation are presented. Factors affecting the accuracy of the estimation methods are analyzed. The applicability of the methods is verified by laboratory measurements and tests in two pilot installations. Test results indicate that the estimation methods can be applied to the analysis and control of pump operation. The accuracy of the methods is sufficient for auditing purposes, and the methods can inform the user if the pump is driven inefficiently.

  3. Liquid sodium pumps

    Allen, H.G.

    1985-01-01

    The pump for use in a nuclear reactor cooling system comprises a booster stage impeller for drawing the liquid through the inlet. A diffuser is affixedly disposed within the pump housing to convert the kinetic pressure imparted to the liquid into increased static pressure. A main stage impeller is rotatively driven by a pump motor at a relatively high speed to impart a relatively high static pressure to the liquid and for discharging the liquid at a relatively high static pressure. A hydraulic coupling is disposed remotely from the liquid path for hydraulically coupling the main stage impeller and the booster stage impeller to rotate the booster stage impeller at a relatively low speed to maintain the low net positive suction pressure applied to the liquid at the inlet greater than the vapor pressure of the liquid and to ensure that the low net positive suction heat, as established by the main stage impeller exceeds the vapor pressure. The coupling comprises a grooved drum which rotates between inner and outer drag coupling members. In a modification the coupling comprises a torque converter. (author)

  4. Electrocentrifugal pumping; Bombeo electrocentrifugo

    Rodriguez Perez, Guillermo; Medellin Otero, Hector [Instituto Mexicano del Peroleo (Mexico)

    1996-07-01

    The exploitation of isolated oil deposits, in losing their own energy, enter a phase of secondary recovery. One of the technologies of new development in Mexico is the one of electrocentrifugal pumping , which consists of introducing the motor-pump as an integral part of the production pipe down to the well bottom and pumping directly up to central complexes, from where it is sent inland. In the present paper is intended to explain what this type of secondary recovery consists of. [Spanish] La explotacion de yacimientos aislados de petroleo, al perder su energia propia, entran en una fase de recuperacion secundaria. Una de las tecnologias de nuevo desarrollo en Mexico es la de bombeo electrocentrifugo, la cual consiste en introducir la motobomba como parte integral de la tuberia de produccion hasta el fondo del pozo y bombearlo directamente hasta los complejos centrales, de donde se envia a tierra. En el presente trabajo se pretende explicar en que consiste este tipo de recuperacion secundaria.

  5. Design and construction of a two-stage centrifugal pump | Nordiana ...

    Centrifugal pumps are widely used in moving liquids from one location to another in homes, offices and industries. Due to the ever increasing demand for centrifugal pumps it became necessary to design and construction of a two-stage centrifugal pump. The pump consisted of an electric motor, a shaft, two rotating impellers ...

  6. Modeling the spatial distribution of the parameters of the coolant in the reactor volume

    Nikonov, S.P.

    2011-01-01

    In this paper the approach to the question about the spatial distribution of the parameters of the coolant in-reactor volume. To describe the in-core space is used specially developed preprocessor. When the work of the preprocessor in the first place, is recreated on the basis of available information (mostly-the original drawings) with high accuracy three-dimensional description of the structures of the reactor volume and, secondly, are prepared on this basis blocks input to the nodal system code improved estimate ATHLET, allows to take into account the hydrodynamic interaction between the spatial control volumes. As an example the special case of solutions of international standard problem on the reconstruction of the transition process in the third unit of the Kalinin nuclear power plant, due to the shutdown of one of the four Main Coolant Pumps in operation at the rated capacity (first download). Model-core area consists of approximately 58 000 control volumes and spatial relationships. It shows the influence of certain structural units of the core to the distribution of the mass floe rate of its height. It is detected a strong cross-flow coolant in the area over the baffle. Moreover, we study the distribution of the coolant temperature at the assembly head of WWER-1000 reactor. It is shown that in the region of the top of the assembly head, where we have installation of thermocouples, the flow coolant for internal assemblies core is formed by only from guide channel Reactor control and protected system Control rod flow, or a mixture of the guide channel flow and flow from the area in front of top grid head assembly (the peripheral assemblies). It is shown that the magnitude of the flow guide channels affects not only the position of control rods, but also the presence of a particular type of measuring channels (Self powered neutron detector sensors or Temperature control sensors) in the cassette. (Author)

  7. Pump failure leads to alternative vertical pump condition monitoring technique

    DeVilliers, Adriaan; Glandon, Kevin

    2011-01-01

    Condition monitoring and detecting early signs of potential failure mechanisms present particular problems in vertical pumps. Most often, the majority of the pump assembly is not readily accessible for visual or audible inspection or conventional vibration monitoring techniques using accelerometers and/or proximity sensors. The root cause failure analysis of a 2-stage vertical centrifugal service-water pump at a nuclear power generating facility in the USA is presented, highlighting this long standing challenge in condition monitoring of vertical pumps. This paper will summarize the major findings of the root cause analysis (RCA), highlight the limitations of traditional monitoring techniques, and present an expanded application of motor current monitoring as a means to gain insight into the mechanical performance and condition of a pump. The 'real-world' example of failure, monitoring and correlation of the monitoring technique to a detailed pump disassembly inspection is also presented. This paper will explain some of the reasons behind well known design principles requiring natural frequency separation from known forcing frequencies, as well as explore an unexpected submerged brittle fracture failure mechanism, and how such issues may be avoided. (author)

  8. Revised Mark 22 coolant temperature coefficients

    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

  9. Freeform Deposition Method for Coolant Channel Closeout

    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.

  10. CAREM-25: considerations about primary coolant chemistry

    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)

  11. Recovery studies for plutonium machining oil coolant

    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

  12. Reactor coolant pressure boundary leakage detection system

    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

  13. Reactor coolant pressure boundary leakage detection system

    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

  14. Reactor coolant pressure boundary leakage detection system

    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.)

  15. Pump Coastdown with the Submerged Flywheel

    Yoon, Hyun-Gi; Seo, KyoungWoo; Kim, Seong Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Many research reactors are generally designed as open pool types in consideration of the heat removal of the nuclear fuels, reactor operation and accessibility. Reactor structure assembly is generally placed at the pool bottom as shown in Fig. 1. Primary cooling system pump circulates the coolant from the reactor structure to the heat exchanger in order to continuously remove the heat generated from the reactor core in the research reactor as shown in Fig. 1. The secondary cooling system releases the transferred heat to the atmosphere by the cooling tower. Coastdown flow rate of the primary cooling system pump with the submerged flywheel are calculated analytically in case of the accident situation. Coastdown flow rate is maintained until almost 80 sec when the pump stops normally. But, coastdown flow rate is rapidly decreased when the flywheel is submerged because of the friction load on the flywheel surface.

  16. Enhancing resistance to burnout via coolant chemistry

    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.

  17. THE PROBLEM OF ENERGY EFFICIENCY OF THE GEOTHERMAL CIRCULATION SYSTEM IN DIFFERENT MODES OF REINJECTION OF THE COOLANT

    D. K. Djavatov; A. A. Azizov

    2017-01-01

    Aim. Advanced technologies are crucial for widespread use of geothermal energy to ensure its competitiveness with conventional forms of energy. To date, the basis for the development of geothermal energy is the technology of extracting the heat transfer fluids from the subsoil. There are the following ways to extract the coolant: freeflow; pumping and circular methods. Of greatest interest is the technology to harness the geothermal energy based on geothermal circulatory system (GCS). There i...

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

    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. Minimizing secondary coolant blowdown in HANARO

    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

  20. Use of microPCM fluids as enhanced liquid coolants in automotive EV and HEV vehicles. Final report

    Mulligan, James C.; Gould, Richard D.

    2001-10-31

    Proof-of-concept experiments using a specific microPCM fluid that potentially can have an impact on the thermal management of automotive EV and HEV systems have been conducted. Samples of nominally 20-micron diameter microencapsulated octacosane and glycol/water coolant were prepared for testing. The melting/freezing characteristics of the fluid, as well as the viscosity, were determined. A bench scale pumped-loop thermal system was used to determine heat transfer coefficients and wall temperatures in the source heat exchanged. Comparisons were made which illustrate the enhancements of thermal performance, reductions of pumping power, and increases of heat transfer which occur with the microPCM fluid.

  1. High-Tc superconducting electric motors

    Schiferl, R.; Stein, J.

    1992-01-01

    In this paper, the advantages and limitations of using superconductors in motors are discussed. A synchronous motor with a high temperature superconducting field winding for pump and fan drive applications is described and some of its unique design features are identified. A 10,000 horsepower superconducting motor design is presented. The critical field and current density requirements for high temperature superconducting wire in motors is discussed. Finally, recent progress in superconducting wire performance is presented

  2. CANDU with supercritical water coolant: conceptual design features

    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)

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

    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.)

  4. Development of Coolant Radioactivity Interpretation Code

    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

  5. Continuous improvement of pump seals

    Wong, W.; Eyvindson, A.; Rhodes, D.B.

    2003-01-01

    Pump seal reliability continues to be an area needing improvement and ongoing vigilance. Methods have been developed for identifying and assessing factors relating to seal performance, selecting the most relevant ones for a specific station, and then focusing on the most significant aspects and how to improve. Discussion invariably addresses maintenance practices, seal design, monitoring capabilities, operating conditions, transients, and pump and motor design. Success in reliability improvement requires ongoing dialogue among the station operators, pump manufacturers and seal designers. AECL CAN-seals lead the nuclear industry in reliability and seal life. They effectively save operators millions of dollars in outage time and person-rem. This paper describes some of the significant developments in AECL's ongoing program in seal R and D, as well as recent new installations following the most demanding seal qualification programs to date. (author)

  6. Solar Pump

    Pique, Charles

    1987-01-01

    Proposed pump moves liquid by action of bubbles formed by heat of sun. Tube of liquid having boiling point of 100 to 200 degrees F placed at focal axis of cylindrical reflector. Concentrated sunlight boils liquid at focus, and bubbles of vapor rise in tube, carrying liquid along with them. Pressure difference in hot tube sufficient to produce flow in large loop. Used with conventional flat solar heating panel in completely solar-powered heat-storage system.

  7. Miniature Scroll Pumps Fabricated by LIGA

    Wiberg, Dean; Shcheglov, Kirill; White, Victor; Bae, Sam

    2009-01-01

    Miniature scroll pumps have been proposed as roughing pumps (low - vacuum pumps) for miniature scientific instruments (e.g., portable mass spectrometers and gas analyzers) that depend on vacuum. The larger scroll pumps used as roughing pumps in some older vacuum systems are fabricated by conventional machining. Typically, such an older scroll pump includes (1) an electric motor with an eccentric shaft to generate orbital motion of a scroll and (2) conventional bearings to restrict the orbital motion to a circle. The proposed miniature scroll pumps would differ from the prior, larger ones in both design and fabrication. A miniature scroll pump would include two scrolls: one mounted on a stationary baseplate and one on a flexure stage (see figure). An electromagnetic actuator in the form of two pairs of voice coils in a push-pull configuration would make the flexure stage move in the desired circular orbit. The capacitance between the scrolls would be monitored to provide position (gap) feedback to a control system that would adjust the drive signals applied to the voice coils to maintain the circular orbit as needed for precise sealing of the scrolls. To minimize power consumption and maximize precision of control, the flexure stage would be driven at the frequency of its mechanical resonance. The miniaturization of these pumps would entail both operational and manufacturing tolerances of pump components. In addition, the vibrations of conventional motors and ball bearings exceed these tight tolerances by an order of magnitude. Therefore, the proposed pumps would be fabricated by the microfabrication method known by the German acronym LIGA ( lithographie, galvanoformung, abformung, which means lithography, electroforming, molding) because LIGA has been shown to be capable of providing the required tolerances at large aspect ratios.

  8. Vessel coolant mass depletion during a 5% SBLOCA in the Semiscale Mod-2C facility

    Shaw, R.A.; Loomis, G.G.

    1985-01-01

    Experimental results are presented from two 5% small-break loss-of-coolant accident (SBLOCA) simulations in the Semiscale Mod-2C facility. In performing the simulated 5% SBLOCAs, boundary conditions scaled from a pressurized water reactor (PWR) were used. The experiment was run with initial conditions typical of a PWR (15.6 MPa pressure and 35 K core differential temperature). The Mod-2C facility represents the state-of-the-art in small facilities scaled from PWRs. Phenomena which occurred during the transient included: primary fluid saturation (change from subcooled to saturated blowdown), break uncovery (a centerline break was simulated), condensation-induced liquid hold-up in the steam generator primary tubes, pump suction liquid seal formation and core level depression with resulting core rod temperature excursion, pump suction liquid seal clearance, loop fluid mass redistribution, and gradual core rewet. The influence of core bypass flow is also discussed. 11 refs., 13 figs

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

    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

  10. High aspect ratio, remote controlled pumping assembly

    Brown, Steve B.; Milanovich, Fred P.

    1995-01-01

    A miniature dual syringe-type pump assembly which has a high aspect ratio and which is remotely controlled, for use such as in a small diameter penetrometer cone or well packer used in water contamination applications. The pump assembly may be used to supply and remove a reagent to a water contamination sensor, for example, and includes a motor, gearhead and motor encoder assembly for turning a drive screw for an actuator which provides pushing on one syringe and pulling on the other syringe for injecting new reagent and withdrawing used reagent from an associated sensor.

  11. Regulating vacuum pump speed with feedback control

    Ludington, D.C.; Aneshansley, D.J.; Pellerin, R.; Guo, F.

    1992-01-01

    Considerable energy is wasted by the vacuum pump/motor on dairy farms. The output capacity (m 3 /min or cfm) of the vacuum pump always exceeds the capacity needed to milk cows and wash pipelines. Vacuum pumps run at full speed and load regardless of actual need for air. Excess air is admitted through a controller. Energy can be saved from electrical demand reduced by regulating vacuum pump speed according to air based on air usage. An adjustable speed drive (ASD) on the motor and controlled based upon air usage, can reduce the energy used by the vacuum pump. However, the ASD unit tested could not maintain vacuum levels within generally accepted guidelines when air usage changed. Adding a high vacuum reserve and a dual vacuum controller between the vacuum pump and the milking pipeline brought vacuum stability within guidelines. The ASD/dual vacuum system can reduce energy consumption and demand by at least 50 percent during milking and provide better vacuum stability than conventional systems. Tests were not run during washing cycles. Using 1990 costs and only the energy saved during milking, the simple payback on investment in new equipment for a 5 hp motor, speed controller and vacuum regulator would be about 5 years

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

    Finn, P.A.

    1991-03-01

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

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

    Finn, P.A.

    1991-03-01

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

  14. Design of the coolant system for the Large Coil Test Facility pulse coils

    Bridgman, C.; Ryan, T.L.

    1983-01-01

    The pulse coils will be a part of the Large Coil Test Facility in Oak Ridge, Tennessee, which is designed to test six large tokamak-type superconducting coils. The pulse coil set consists of two resistive coaxial solenoid coils, mounted so that their magnetic axis is perpendicular to the toroidal field lines of the test coil. The pulse coils provide transient vertical fields at test coil locations to simulate the pulsed vertical fields present in tokamak devices. The pulse coils are designed to be pulsed for 30 s every 150 s, which results in a Joule heating of 116 kW per coil. In order to provide this capability, the pulse coil coolant system is required to deliver 6.3 L/s (100 gpm) of subcooled liquid nitrogen at 10-atm absolute pressure. The coolant system can also cool down each pulse coil from room temperature to liquid nitrogen temperature. This paper provides details of the pumping and heat exchange equipment designed for the coolant system and of the associated instrumentation and controls

  15. Pumping behavior of sputter ion pumps

    Chou, T.S.; McCafferty, D.

    The ultrahigh vacuum requirements of ISABELLE is obtained by distributed pumping stations. Each pumping station consists of 1000 l/s titanium sublimation pump for active gases (N 2 , H 2 , O 2 , CO, etc.), and a 20 l/s sputter ion pump for inert gases (methane, noble gases like He, etc.). The combination of the alarming production rate of methane from titanium sublimation pumps (TSP) and the decreasing pumping speed of sputter ion pumps (SIP) in the ultrahigh vacuum region (UHV) leads us to investigate this problem. In this paper, we first describe the essential physics and chemistry of the SIP in a very clean condition, followed by a discussion of our measuring techniques. Finally measured methane, argon and helium pumping speeds are presented for three different ion pumps in the range of 10 -6 to 10 -11 Torr. The virtues of the best pump are also discussed

  16. Tests of cooling water pumps at Dukovany nuclear power plant

    Travnicek, J.

    1986-01-01

    Tests were performed to examine the operating conditions of the 1600 BQDV cooling pumps of the main coolant circuit of unit 1 of the Dukovany nuclear power plant. For the pumps, the performance was tested in the permissible operating range, points were measured below this range and the guaranteed operating point was verified. Pump efficiency was calculated from the measured values. The discussion of the measurement of parameters has not yet been finished because the obtained values of the amount delivered and thus of the pump efficiency were not up to expectation in all detail. It was also found that for obtaining the guaranteed flow the pump impeller had to be opened to 5deg -5.5deg instead of the declared 3deg. Also tested were pump transients, including the start of the pump, its stop, the operation and failure of one of the two pumps. In these tests, pressures were also measured at the inlet and the outlet of the inner part of the TG 11 turbine condenser. It was shown that the time course and the pressure course of the processes were acceptable. In addition to these tests, pressure losses in the condenser and the cooling water flow through the feed pump electromotor cooler wre tested for the case of a failure of one of the two pumps. (E.S.)

  17. High Power Density Motors

    Kascak, Daniel J.

    2004-01-01

    With the growing concerns of global warming, the need for pollution-free vehicles is ever increasing. Pollution-free flight is one of NASA's goals for the 21" Century. , One method of approaching that goal is hydrogen-fueled aircraft that use fuel cells or turbo- generators to develop electric power that can drive electric motors that turn the aircraft's propulsive fans or propellers. Hydrogen fuel would likely be carried as a liquid, stored in tanks at its boiling point of 20.5 K (-422.5 F). Conventional electric motors, however, are far too heavy (for a given horsepower) to use on aircraft. Fortunately the liquid hydrogen fuel can provide essentially free refrigeration that can be used to cool the windings of motors before the hydrogen is used for fuel. Either High Temperature Superconductors (HTS) or high purity metals such as copper or aluminum may be used in the motor windings. Superconductors have essentially zero electrical resistance to steady current. The electrical resistance of high purity aluminum or copper near liquid hydrogen temperature can be l/lOO* or less of the room temperature resistance. These conductors could provide higher motor efficiency than normal room-temperature motors achieve. But much more importantly, these conductors can carry ten to a hundred times more current than copper conductors do in normal motors operating at room temperature. This is a consequence of the low electrical resistance and of good heat transfer coefficients in boiling LH2. Thus the conductors can produce higher magnetic field strengths and consequently higher motor torque and power. Designs, analysis and actual cryogenic motor tests show that such cryogenic motors could produce three or more times as much power per unit weight as turbine engines can, whereas conventional motors produce only 1/5 as much power per weight as turbine engines. This summer work has been done with Litz wire to maximize the current density. The current is limited by the amount of heat it

  18. Coolant Void Reactivity Analysis of CANDU Lattice

    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.

  19. EDF PWRs primary coolant purification strategies

    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)

  20. Coolant degassing device for PWR type reactors

    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.)

  1. Microstructural characterization of primary coolant pipe steel

    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

  2. Calorimetric and reactor coolant system flow uncertainty

    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

  3. Reactor feedwater pump control device

    Nishiyama, Hiroyuki.

    1990-01-01

    An amount of feedwater necessary for ensuring reactor inventory after scram is ensured automatically based on the reactor output before scram of a BWR type reactor. That is, if scram should occur, a feedwater flow rate just before the scram is stored by reactor output signals. Further, the amount of feedwater required after the scram is determined based on the output of the memory. The reactor power after the scram based on a feedwater flow rate and a main steam flow rate is inputted to an integrator, to calculate and output the amount of the feedwater flow rate (1) injected after the scram for the inventory. A coast down flowrate (2) in a case of pump trip is forecast by the output signals. Automatic trip is outputted to all turbine driving feedwater pumps when the sum of (1) and (2) exceeds a necessary and sufficient amount of feedwater required for ensuring inventory. For motor driving feedwater pumps, only a portion, for example, one of the pumps is automatically started while other pumps are stopped their operation, only in this case, to prevent excess water feeding. (I.S.)

  4. Replacement Saltwell Pumping System Document Bibliography

    BELLOMY, J.R.

    2000-01-01

    This document bibliography is prepared to identify engineering documentation developed during the design of the Replacement Saltwell Pumping System. The bibliography includes all engineering supporting documents and correspondence prepared prior to the deployment of the system in the field. All documents referenced are available electronically through the Records Management Information System (RMIS). Major components of the Replacement Saltwell Pumping System include the Sundyne Canned Motor Pump, the Water Filter Skid, the Injection Water Skid and the Backflow Preventer Assembly. Drawing H-14-104498 provides an index of drawings (fabrication details, PandIDs, etc.) prepared to support development of the Replacement Saltwell Pumping System. Specific information pertaining to new equipment can be found in Certified Vendor Information (CVI) File 50124. This CVI file has been established specifically for new equipment associated with the Replacement Saltwell Pumping System

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

    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)

  6. Circulating water pumps for nuclear power stations

    Satoh, Hiroshi; Ohmori, Tsuneaki

    1979-01-01

    Shortly, the nuclear power station with unit power output of 1100 MW will begin the operation, and the circulating water pumps manufactured recently are those of 2.4 to 4 m bore, 840 to 2170 m 3 /min discharge and 2100 to 5100 kW driving power. The circulating water pumps are one of important auxiliary machines, because if they fail, power generation capacity lowers immediately. Enormous quantity of cooling water is required to cool condensers, therefore in Japan, sea water is usually used. As siphon is formed in circulating water pipes, the total head of the pumps is not very high. The discharge of the pumps is determined so as to keep the temperature rise of discharged water lower than 7 deg. C. The quantity of cooling water for nuclear power generation is about 50% more as compared with thermal power generation because of the difference in steam conditions. The total head of the pumps is normally from 8 to 15 m. The circulating water pumps rarely stop after they started the operation, therefore it is economical to determine the motor power so that it can withstand 10% overload for a short period, instead of large power. At present, vertical shaft, oblique flow circulating water pumps are usually employed. Recently, movable blade pumps are adopted. The installation, construction and materials of the pumps and the problems are described. (Kako, I.)

  7. Browns Ferry Nuclear Plant: variation in test intervals for high-pressure coolant injection (HPCI) system

    Christie, R.F.; Stetkar, J.W.

    1985-01-01

    The change in availability of the high-pressure coolant injection system (HPCIS) due to a change in pump and valve test interval from monthly to quarterly was analyzed. This analysis started by using the HPCIS base line evaluation produced as part of the Browns Ferry Nuclear Plant (BFN) Probabilistic Risk Assessment (PRA). The base line evaluation showed that the dominant contributors to the unavailability of the HPCI system are hardware failures and the resultant downtime for unscheduled maintenance. The effect of changing the pump and valve test interval from monthly to quarterly was analyzed by considering the system unavailability due to hardware failures, the unavailability due to testing, and the unavailability due to human errors that potentially could occur during testing. The magnitude of the changes in unavailability affected by the change in test interval are discussed. The analysis showed a small increase in the availability of the HPCIS to respond to loss of coolant accidents (LOCAs) and a small decrease in the availability of the HPCIS to respond to transients which require HPCIS actuation. In summary, the increase in test interval from monthly to quarterly does not significantly impact the overall HPCIS availability

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

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

    1997-04-01

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

  9. Natural circulation in reactor coolant system

    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

  10. CFD analyses of coolant channel flowfields

    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.

  11. Efficiency of water coolant for DEMO divertor

    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.

  12. Chemistry of liquid metal coolants and sensors

    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

  13. Efficiency of water coolant for DEMO divertor

    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. Requirements of coolants in nuclear reactors

    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)

  15. Polar representation of centrifugal pump homologous curves

    Veloso, Marcelo Antonio; Mattos, Joao Roberto Loureiro de

    2008-01-01

    Essential for any mathematical model designed to simulate flow transient events caused by pump operations is the pump performance data. The performance of a centrifugal pump is characterized by four basic parameters: the rotational speed, the volumetric flow rate, the dynamic head, and the hydraulic torque. Any one of these quantities can be expressed as a function of any two others. The curves showing the relationships between these four variables are called the pump characteristic curves, also referred to as four-quadrant curves. The characteristic curves are empirically developed by the pump manufacturer and uniquely describe head and torque as functions of volumetric flow rate and rotation speed. Because of comprising a large amount of points, the four-quadrant configuration is not suitable for computational purposes. However, it can be converted to a simpler form by the development of the homologous curves, in which dynamic head and hydraulic torque ratios are expressed as functions of volumetric flow and rotation speed ratios. The numerical use of the complete set of homologous curves requires specification of sixteen partial curves, being eight for the dynamic head and eight for the hydraulic torque. As a consequence, the handling of homologous curves is still somewhat complicated. In solving flow transient problems that require the pump characteristic data for all the operation zones, the polar form appears as the simplest way to represent the homologous curves. In the polar method, the complete characteristics of a pump can be described by only two closed curves, one for the dynamic head and other for the hydraulic torque, both in function of a single angular coordinate defined adequately in terms of the quotient between volumetric flow ratio and rotation speed ratio. The usefulness and advantages of this alternative method are demonstrated through a practical example in which the homologous curves for a pump of the type used in the main coolant loops of a

  16. Parametric representation of centrifugal pump homologous curves

    Veloso, Marcelo A.; Mattos, Joao R.L. de

    2015-01-01

    Essential for any mathematical model designed to simulate flow transient events caused by pump operations is the pump performance data. The performance of a centrifugal pump is characterized by four basic quantities: the rotational speed, the volumetric flow rate, the dynamic head, and the hydraulic torque. The curves showing the relationships between these four variables are called the pump characteristic curves. The characteristic curves are empirically developed by the pump manufacturer and uniquely describe head and torque as functions of volumetric flow rate and rotation speed. Because of comprising a large amount of points, this configuration is not suitable for computational purposes. However, it can be converted to a simpler form by the development of the homologous curves, in which dynamic head and hydraulic torque ratios are expressed as functions of volumetric flow and rotation speed ratios. The numerical use of the complete set of homologous curves requires specification of sixteen partial curves, being eight for the dynamic head and eight for the hydraulic torque. As a consequence, the handling of homologous curves is still somewhat complicated. In solving flow transient problems that require the pump characteristic data for all the operation zones, the parametric form appears as the simplest way to deal with the homologous curves. In this approach, the complete characteristics of a pump can be described by only two closed curves, one for the dynamic head and other for the hydraulic torque, both in function of a single angular coordinate defined adequately in terms of the quotient between volumetric flow ratio and rotation speed ratio. The usefulness and advantages of this alternative method are demonstrated through a practical example in which the homologous curves for a pump of the type used in the main coolant loops of a pressurized water reactor (PWR) are transformed to the parametric form. (author)

  17. Duodopa pump treatment in patients with advanced Parkinson's disease

    Karlsborg, Merete; Korbo, Lise; Regeur, Lisbeth

    2010-01-01

    Patients with advanced Parkinson's disease (PD) often develop motor complications including fluctuations and involuntary movements (dyskinesias). In Denmark, treatment has comprised Deep Brain Stimulation (DBS) since the late 1990s, and as from 2002 use of a subcutaneous apomorphine pump. Monothe......Patients with advanced Parkinson's disease (PD) often develop motor complications including fluctuations and involuntary movements (dyskinesias). In Denmark, treatment has comprised Deep Brain Stimulation (DBS) since the late 1990s, and as from 2002 use of a subcutaneous apomorphine pump...

  18. Upper internals of PWR with coolant flow separator

    Chevereau, G.; Heuze, A.

    1989-01-01

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

  19. Coolant make-up device for BWR type reactor

    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.)

  20. Motor Neurons

    Hounsgaard, Jorn

    2017-01-01

    Motor neurons translate synaptic input from widely distributed premotor networks into patterns of action potentials that orchestrate motor unit force and motor behavior. Intercalated between the CNS and muscles, motor neurons add to and adjust the final motor command. The identity and functional...... in in vitro preparations is far from complete. Nevertheless, a foundation has been provided for pursuing functional significance of intrinsic response properties in motoneurons in vivo during motor behavior at levels from molecules to systems....