WorldWideScience

Sample records for reactor pump final

  1. Fusion reactor pumped laser

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1988-01-01

    A nuclear pumped laser is described comprising: a toroidal fusion reactor, the reactor generating energetic neutrons; an annular gas cell disposed around the outer periphery of the reactor, the cell including an annular reflecting mirror disposed at the bottom of the cell and an annular output window disposed at the top of the cell; a gas lasing medium disposed within the annular cell for generating output laser radiation; neutron reflector material means disposed around the annular cell for reflecting neutrons incident thereon back into the gas cell; neutron moderator material means disposed between the reactor and the gas cell and between the gas cell and the neutron reflector material for moderating the energy of energetic neutrons from the reactor; converting means for converting energy from the moderated neutrons to energy pumping means for pumping the gas lasing medium; and beam compactor means for receiving output laser radiation from the annular output window and generating a single output laser beam therefrom

  2. Pumps for German pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Reactor recirculation pump test loop

    International Nuclear Information System (INIS)

    Taka, Shusei; Kato, Hiroyuki

    1979-01-01

    A test loop for a reactor primary loop recirculation pumps (PLR pumps) has been constructed at Ebara's Haneda Plant in preparation for production of PLR pumps under license from Byron Jackson Pump Division of Borg-Warner Corporation. This loop can simulate operating conditions for test PLR pumps with 130 per cent of the capacity of pumps for a 1100 MWe BWR plant. A main loop, primary cooling system, water demineralizer, secondary cooling system, instrumentation and control equipment and an electric power supply system make up the test loop. This article describes the test loop itself and test results of two PLR pumps for Fukushima No. 2 N.P.S. Unit 1 and one main circulation pump for HAZ Demonstration Test Facility. (author)

  4. Reactor coolant pumps for nuclear reactors

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  5. Fast breeder reactor electromagnetic pump

    International Nuclear Information System (INIS)

    Araseki, Hideo; Murakami, Takahiro

    2008-01-01

    Main pumps circulating sodium in the FBR type reactor have been mechanical types, not electromagnetic pumps. Electromagnetic pump of 1-2 m 3 /min has been used as an auxiliary pump. Large sized electromagnetic pumps such as several hundred m 3 /min have not been commercialized due to technical difficulties with electromagnetic instability and pressure pulsations. This article explained electromagnetic and fluid equations and magnetic Reynolds number related with electromagnetic pumps and numerical analysis of instability characteristics and pressure pulsations and then described applications of the results to FBR system. Magnetic Reynolds number must be chosen less than one with appropriate operating frequency and optimum slip of 0.2-0.4. (T. Tanaka)

  6. Reactor feedwater pump control device

    International Nuclear Information System (INIS)

    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.)

  7. Reactor coolant pump transportation incident

    International Nuclear Information System (INIS)

    Noce, D.

    1992-01-01

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

  8. Vertical reactor coolant pump instabilities

    International Nuclear Information System (INIS)

    Jones, R.M.

    1985-01-01

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

  9. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

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

  10. Fusion reactor high vacuum pumping

    International Nuclear Information System (INIS)

    Sedgley, D.W.; Walthers, C.R.; Jenkins, E.M.

    1992-01-01

    This paper reports on recent experiments which have shown the practicality of using activated carbon (coconut charcoal) at 4K to pump helium and hydrogen isotopes for a fusion reactor. Both speed and capacity for deuterium/helium and tritium/helium-3 mixtures were satisfactory. The long-term effects of tritium on the charcoal/cement system developed by Grumman and LLNL was now known; therefore a program was undertaken to see what, if any, effect long-term tritium exposure has on the cryosorber. Several charcoal on aluminum test samples were subjected to six months exposure of tritium at approximately 77 K. The tritium was scanned several times with a residual gas analyzer and the speed-capacity performance of the samples was measured before, approximately one-third way through, and after the exposure. Modest effects were noted which would not seriously restrict the use of charcoal as a cryosorber for fusion reactor high-vacuum pumping applications

  11. Reactor having coolant recycling pump

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Tsutsui, Toshiaki

    1979-01-01

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

  13. Mathematical model of the reactor coolant pump

    International Nuclear Information System (INIS)

    Kozuh, M.

    1989-01-01

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

  14. Vacuum pumping for controlled thermonuclear reactors

    International Nuclear Information System (INIS)

    Watson, J.S.; Fisher, P.W.

    1976-01-01

    Thermonuclear reactors impose unique vacuum pumping problems involving very high pumping speeds, handling of hazardous materials (tritium), extreme cleanliness requirements, and quantitative recovery of pumped materials. Two principal pumping systems are required for a fusion reactor, a main vacuum system for evacuating the torus and a vacuum system for removing unaccelerated deuterium from neutral beam injectors. The first system must pump hydrogen isotopes and helium while the neutral beam system can operate by pumping only hydrogen isotopes (perhaps only deuterium). The most promising pumping techniques for both systems appear to be cryopumps, but different cryopumping techniques can be considered for each system. The main vacuum system will have to include cryosorption pumps cooled to 4.2 0 K to pump helium, but the unburned deuterium-tritium and other impurities could be pumped with cryocondensation panels (4.2 0 K) or cryosorption panels at higher temperatures. Since pumping speeds will be limited by conductance through the ducts and thermal shields, the pumping performance for both systems will be similar, and other factors such as refrigeration costs are likely to determine the choice. The vacuum pumping system for neutral beam injectors probably will not need to pump helium, and either condensation or higher temperature sorption pumps can be used

  15. Liquid metal pump for nuclear reactors

    International Nuclear Information System (INIS)

    Allen, H.G.; Maloney, J.R.

    1975-01-01

    A pump for use in pumping high temperature liquids at high pressures, particularly liquid metals used to cool nuclear reactors is described. It is of the type in which the rotor is submerged in a sump but is fed by an inlet duct which bypasses the sump. A chamber, kept full of fluid, surrounds the pump casing into which fluid is bled from the pump discharge and from which fluid is fed to the rotor bearings and hence to the sump. This equalizes pressure inside and outside the pump casing and reduces or eliminates the thermal shock to the bearings and sump tank

  16. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Pothier, N.E.; Metcalfe, R.

    1986-06-01

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

  17. Reactor primary pumps dynamic balancing test

    International Nuclear Information System (INIS)

    Lu Qunxian

    2002-01-01

    Reactor primary Pump is the important equipment in the primary circuit, its working quality would directly influence the safety and operation of nuclear power plant. The author describes that the primary pump vibration status, vibration fault diagnosis and dynamic balancing process on site have been performed since commercial operation of DA YA BAY Nuclear Power plant

  18. Reliability of reactor plant water cleanup pumps

    International Nuclear Information System (INIS)

    Pearson, J.L.

    1979-01-01

    Carolina Power and Light Company's Brunswick 2 nuclear plant experienced a high reactor water cleanup pump-failure rate until inlet temperature and flow were reduced and mechanical modifications were implemented. Failures have been zero for about one year, and water cleanup efficiency has increased

  19. Inducer pumps for liquid metal reactor plants

    International Nuclear Information System (INIS)

    Jackson, E.D.

    2002-01-01

    Pumps proposed for liquid metal reactor plants typically use centrifugal impellers as the rotating element and are required to maintain a relatively low speed to keep the suction specific speed low enough to operate at the available net positive suction head (HPSH) and to avoid cavitation damage. These low speeds of operation require that the pump diameter increase and/or multiple stages be used to achieve the design head. This frequently results in a large, heavy, complex pump design. In addition, the low speed results in a larger drive motor size so that the resultant penalty to the plant designer is multiplied. The heavier pump can also result in further complications as, e.g., the difficulty in maintaining the first critical speed sufficiently above the pump operating range to provide margin for rotor dynamic stability. To overcome some of these disadvantages, it was proposed the use of inducer pumps for Liquid Metal Fast Breeder Reactor (LMFBR) plants. This paper discusses some of the advantages of the inducer pump and the development history of designing and testing these pumps both in water and sodium. The inducer pump is seen to be a sound concept with a strong technology base derived from the aerospace and ship propulsion industries. The superior suction performance capability of the inducer offers significant system design advantages, primarily a smaller, lighter weight, less complex pump design with resulting saving in cost. Extensive testing of these pumps has been conducted in both sodium and water to demonstrate the long-life capability with no cavitation damage occurring in those designs based on Rockwell's current design criteria. These tests have utilized multiple inspection and measurement approaches to accurately assess and identify any potential for cavitation damage, and these approaches have all concluded that no damage is occurring. Therefore, it is concluded that inducer pumps can be safely designed for long life operation in sodium with

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  1. Fission reactor recycling pump handling device

    International Nuclear Information System (INIS)

    Togasawa, Hiroshi; Komita, Hideo; Susuki, Shoji; Endo, Takio; Yamamoto, Tetsuzo; Takahashi, Hideaki; Saito, Noboru.

    1991-01-01

    This invention provides a device for handling a recycling pump in a nuclear reactor upon periodical inspections in a BWR type power plant. That is, in a handling device comprising a support for supporting components of a recycling pump, and a lifter for vertically moving the support below a motor case disposed passing through a reactor pressure vessel, a weight is disposed below the support. Then, the center of gravity of the components, the support and the entire weight is substantially aligned with the position for the support. With such a constitution, the components can be moved vertically to the motor case extremely safely, to remarkably suppress vibrations. Further, the operation safety can remarkably be improved by preventing turning down upon occurrence of earthquakes. Further, since vibration-proof jigs as in a prior art can be saved, operation efficiency can be improved. (I.S.)

  2. Fission reactor recycling pump handling device

    Energy Technology Data Exchange (ETDEWEB)

    Togasawa, Hiroshi; Komita, Hideo; Susuki, Shoji; Endo, Takio; Yamamoto, Tetsuzo; Takahashi, Hideaki; Saito, Noboru

    1991-06-24

    This invention provides a device for handling a recycling pump in a nuclear reactor upon periodical inspections in a BWR type power plant. That is, in a handling device comprising a support for supporting components of a recycling pump, and a lifter for vertically moving the support below a motor case disposed passing through a reactor pressure vessel, a weight is disposed below the support. Then, the center of gravity of the components, the support and the entire weight is substantially aligned with the position for the support. With such a constitution, the components can be moved vertically to the motor case extremely safely, to remarkably suppress vibrations. Further, the operation safety can remarkably be improved by preventing turning down upon occurrence of earthquakes. Further, since vibration-proof jigs as in a prior art can be saved, operation efficiency can be improved. (I.S.).

  3. Study and application of boiling water reactor jet pump characteristic

    International Nuclear Information System (INIS)

    Liao Lihyih

    1992-01-01

    RELAP5/MOD2 is an advanced thermal-hydraulic computer code used to analyze plant response to postulated transient and loss-of-coolant accidents in light water nuclear reactors. Since this computer code was originally developed for pressurized water reactor transient analysis, some of its capabilities are questioned when the methods are applied to a boiling water reactor. One of the areas which requires careful assessment is the jet pump model. In this paper, the jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From an investigation of the momentum equations, it is found that the jet pump models of these codes are not exactly the same. However, the effects of the jet pump models on the M-N characteristic curve are negligible. In this study, it is found that the relationship between the flow ratio, M, and the head ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions, the M-N characteristic curve will not change with power, level, recirculation pump speed or loop flow rate. When the effects of wall friction and gravitational head are included, the shape of the M-N curve will change. For certain conditions, the slope of the M-N curve can even change from negative to positive. The changes in the M-N curve caused by the separate effects of the wall friction and gravitational head will be presented. Sensitivity studies on the drive flow nozzle form loss coefficients, K d , the suction flow junction form loss coefficients, K s , the diffuser form loss coefficient, K c , and the ratio of different flow areas in the jet pump are performed. Finally, useful guidelines will be presented for plants without a plant specific M-N curve. (orig.)

  4. Reactor coolant pump seal leakage monitoring

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  5. Reactor coolant pump monitoring and diagnostic system

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  6. Reactor operations for nuclear pumping of lasers

    Energy Technology Data Exchange (ETDEWEB)

    Beck, G; Cooper, G [University of Illinois (United States)

    1974-07-01

    Experiments involving the measurement of gas parameters that are related to lasing, and lasing of various gas mixtures have comprised a major part of the utilization of the University of Illinois Advanced TRIGA Reactor since the upgrading of the facility was completed in 1969. A thru beam port, which was added during upgrading, has been the facility used for these measurements. The laser cell is placed in the port adjacent to the core. Alignment is then accomplished by using both ends of the port or by a mirror placed at the back side of the apparatus. The reactor has been operated in all modes (pulsing, square wave, and steady state) for the experiments although pulsing is the primary mode that is used. Laser enhancement has been obtained in several cases, but efforts toward direct pumping from the radiation alone have not as yet succeeded. Improved laser operation from direct pumping has been suggested with an emphasis on high-powered systems where the basic input energy is to be derived from a nuclear reactor.

  7. Power supplyer for reactor coolant recycling pump

    International Nuclear Information System (INIS)

    Nara, Hiroshi; Okinaka, Yo.

    1991-01-01

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

  8. Automated surveillance of reactor coolant pump performance

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  9. Utilization of large electromagnetic pumps in the fast breeder reactors

    International Nuclear Information System (INIS)

    Deverge, C.; Lefrere, J.P.; Peturaud, P.; Sauvage, M.

    1984-04-01

    After an overview concerning the induction annular electromagnetic pumps and the dimensioning methods usually utilized, development of these components for a fast breeder integrated reactor is considered: - utilization of cooled EMP in the intermediate circuit, - utilization of immersed pumps, coupled with the intermediate exchanger, for the primary pumping; dimensioning, energetic aspects, and effects on the power plant geometrical configurations [fr

  10. Method of starting internal pumps of a nuclear reactor

    International Nuclear Information System (INIS)

    Kumagami, Shoji.

    1985-01-01

    Purpose: To reduce the noise effects by decreasing the invading current into the main line upon starting an internal pump type nuclear reactor adapted to forcively recycle the reactor water by a plurality of internal pumps. Method: A plurality of internal pumps are divided into several groups and, upon starting pumps belonging to the individual unit group, the starting instances for the respective pumps are deviated to reduce the surges applied to the main line and suppress the invading current lower to reduce the earth noises. As a result, effects caused to other devices or equipments can be moderated to improve the reliability. Furthermore, by actuating the respective pumps on every group units in a starting pattern along the orthogonal line, flow rate distribution in the reactor can be balanced. Then, the instability region during low rotation of pumps, that is, instability of the flow rate near the resonance frequency can be decreased. (Kawakami, Y.)

  11. Monitoring for shaft cracks on reactor recirculation pumps

    International Nuclear Information System (INIS)

    Kowal, M.G.; O'Brien, J.T. Jr.

    1989-01-01

    The article discusses the vibration characteristics associated with a boiling water reactor (BWR) recirculation pump. It also describes the application of diagnostic techniques and shaft crack theory to an on-line diagnostic monitoring system for reactor recirculation pumps employed at Philadelphia Electric Company's Peach Bottom Atomic Power Station. Specific emphasis is placed on the unique monitoring techniques associated with these variable speed vertical pumps

  12. Maintenance of reactor recirculation pumps [Paper No.: II-1

    International Nuclear Information System (INIS)

    Ansari, M.A.; Bhat, K.P.

    1981-01-01

    At Tarapur Atomic Power Station (TAPS), two reactor recirculation pumps are provided, one each for the two reactor units. The performance of pumps has been uniformly good; however, leakage through the cartridge type, two stage, mechanical seals which are installed on these pumps was encountered on few occasions. The paper describes the leakage problems, identification of certain design deficiencies and rectification carried out at TAPS for overcoming these problems. (author)

  13. LMR [liquid metal reactor] centrifugal pump coastdowns

    International Nuclear Information System (INIS)

    Dunn, F.E.; Malloy, D.J.

    1987-01-01

    A centrifugal pump model which describes the interrelationships of the pump discharge flowrate, pump speed, shaft torque and dynamic head has been implemented based upon existing models. Specifically, the pump model is based upon the dimensionless-homologous pump theory of Wylie and Streeter. Given data from a representative pump, homologous theory allows one to predict the transient characteristics of similarly sized pumps. This homologous pump model has been implemented into both the one-dimensional SASSYS-1 systems analysis code and the three-dimensional COMMIX-1A code. Comparisons have been made both against other pump models (CRBR) and actual pump coastdown data (EBR-II and FFTF). Agreement with this homologous pump model has been excellent. Additionally, these comparisons indicate the validity of applying the medium size pump data of Wylie and Streeter to a range of typical LMR centrifugal pumps

  14. An optimized design of rectangle pumping cell for nuclear reactor pumped laser

    International Nuclear Information System (INIS)

    Wan, J.-S.; Chen, L.-X.; Zhao, Z.-M.; Pan, X.-B.; Jing, C.-Y.; Zhao, X.-Q.; Liu, F.-H.

    2003-01-01

    Basing on our research of energy deposition in RPL (Reactor Pumped Laser) pumping cell and the laser power efficiency, a RPL test device on Pulsed Reactor has been designed. In addition, the laser beam power of the RPL test device is estimated in the paper. (author)

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

    International Nuclear Information System (INIS)

    Grandia, M.R.

    1976-03-01

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

  16. Development of ceramic vacuum pumps for fusion reactors

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    To achieve the magnetic field resistance and tritium resistance which are required for vacuum pumps for fusion reactors, a vacuum pump consisting of middle-ceramic turbo molecular pump (TMP), using ceramic rotor and ceramic turbo roughing pump was developed. In colaboration with the Japan Atomic Energy Research Institute, performance tests on pumping speed, compression ratio of middle-ceramic TMP and both of pumping characteristics were carried out. Sufficient performances were obtained. It was showed that middle-ceramic TMP had pumping speed of more than 500 l/s, and could achieve the pressure below 4 x 10 -7 Pa. Ceramic turbo roughing pump could vacuum from atmospheric pressure. It is concluded that complete oil-free ceramic vacuum pump can be put into practical use (K.S.)

  17. Cryosorption vacuum pumping under fusion reactor conditions

    International Nuclear Information System (INIS)

    Watson, J.S.; Fisher, P.W.

    1977-01-01

    Experiments are in progress on cryosorption pumping using a double-chevron pump with a molecular sieve pumping surface. Studies have been made with hydrogen, deuterium, helium, and deuterium-helium at 10 -7 to 3 x 10 -3 torr. Steady pumping speeds were observed for deuterium; above 10 -4 torr the speed increased with pressure until runaway occurred. At less than 10 -6 torr and low panel loading, hydrogen pumping speeds resemble those for deuterium. At higher pressures, the pump can function by condensation or sorption, and unsteady speeds are observed. Helium pumping is always by sorption, but regions of instability have been observed and defined. Deuterium-helium pumping tests showed that deuterium condensation on the panels prevents cryosorption of helium; however, compound pumps with separate panels for helium and hydrogen will be satisfactory

  18. Electromagnetic pumping of liquid lithium in inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Baker, R.S.; Blink, J.A.; Tessier, M.J.

    1983-01-01

    The basic operating principles and geometries of ten electromagnetic pumps are described. Two candidate pumps, the annular-linear-induction pump and the helical-rotor electromagnetic pump, are compared for possible use in a full-scale liquid-lithium inertial confinement fusion reactor. A parametric design study completed for the helical-rotor pump is shown to be valid when applied to an experimental sodium pump. Based upon the preliminary HYLIFE requirements for a lithium flow rate per pump of 8.08 m 3 /s at a head of 82.5 kPa, a complete set of 70 variables are specified for a helical-rotor pump with either a normally conducting or a superconducting winding. The two alternative designs are expected to perform with efficiencies of 50 and 60%, respectively

  19. The operating reliability of the reactor coolant pump

    International Nuclear Information System (INIS)

    Grancy, W.

    1996-01-01

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

  20. Final Stage Development of Reactor Console Simulator

    International Nuclear Information System (INIS)

    Mohamad Idris Taib; Ridzuan Abdul Mutalib; Zareen Khan Abdul Jalil Khan; Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Nurfarhana Ayuni Joha

    2013-01-01

    The Reactor Console Simulator PUSPATI TRIGA Reactor was developed since end of 2011 and now in the final stage of development. It is will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behavior and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of human system interface (HSI) is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate and estimated reactor console parameters. The capabilities in user interface, reactor physics and thermal-hydraulics can be expanded and explored to simulation as well as modeling for New Reactor Console, Research Reactor and Nuclear Power Plant. (author)

  1. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-15

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

  2. Pressurizer pump reliability analysis high flux isotope reactor

    International Nuclear Information System (INIS)

    Merryman, L.; Christie, B.

    1993-01-01

    During a prolonged outage from November 1986 to May 1990, numerous changes were made at the High Flux Isotope Reactor (HFIR). Some of these changes involved the pressurizer pumps. An analysis was performed to calculate the impact of these changes on the pressurizer system availability. The analysis showed that the availability of the pressurizer system dropped from essentially 100% to approximately 96%. The primary reason for the decrease in availability comes because off-site power grid disturbances sometimes result in a reactor trip with the present pressurizer pump configuration. Changes are being made to the present pressurizer pump configuration to regain some of the lost availability

  3. Self-sustaining nuclear pumped laser-fusion reactor experiment

    International Nuclear Information System (INIS)

    Boody, F.P.; Choi, C.K.; Miley, G.H.

    1977-01-01

    The features of a neutron feedback nuclear pumped (NFNP) laser-fusion reactor equipment were studied with the intention of establishing the feasibility of the concept. The NFNP laser-fusion concept is compared schematically to electrically pumped laser fusion. The study showed that, once a method of energy storage has been demonstrated, a self-sustaining fusion-fission hybrid reactor with a ''blanket multiplication'' of two would be feasible using nuclear pumped Xe F* excimer lasers having efficiencies of 1 to 2 percent and D-D-T pellets with gains of 50 to 100

  4. Energy efficient ammonia heat pump. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Madsen, Claus; Pijnenburg, B.; Schumann Grindorf, H. [Danish Technological Institute, Aarhus (Denmark); Christensen, Rolf [Alfa Laval, Lund (Sweden); Rasmussen, Bjarne D. [Grundfos, Bjerringbro (Denmark); Gram, S.; Fredborg Jakobsen, D. [Svedan Industri Koeleanlaeg, Greve (Denmark)

    2013-09-15

    The report describes the development of a highly effective ammonia heat pump. Heat pumps play an increasingly important role in the search for more effective use of energy in our society. Highly efficient heat pumps can contribute to reduced energy consumption and improved economy of the systems which they are a part of. An ammonia heat pump with high pressure reciprocating compressor and a novel split condenser was developed to prove potential for efficiency optimization. The split of the condenser in two parts can be utilized to obtain smaller temperature approaches and, thereby, improved heat pump efficiency at an equal heat exchanger area, when compared to the traditional solution with separate condenser and de-superheater. The split condenser design can also be exploited for heating a significant share of the total heating capacity to a temperature far above the condensing temperature. Furthermore, the prototype heat pump was equipped with a plate type evaporator combined with a U-turn separator with a minimum liquid height and a liquid pump with the purpose of creating optimum liquid circulation ratio for the highest possible heat transfer coefficients at the lowest possible pressure drop. The test results successfully confirmed the highest possible efficiency; a COP of 4.3 was obtained when heating water from 40 deg. C to 80 deg. C while operating with evaporating/condensing temperatures of +20 deg C/+73 deg C. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. Nuclear reactor with coolant circulation pumps

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  7. Solid State Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Mays, G.T.

    2004-03-10

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

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

    International Nuclear Information System (INIS)

    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)

  9. Trends and experiences in reactor coolant pump motors

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

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

  10. Upgraded prototype-reactor internal pump for ABWR

    International Nuclear Information System (INIS)

    Kumagai, Mikio; Amemori, Shiro; Saito, Takehiko

    1988-01-01

    In 1983, Toshiba, using their own technology, manufactured a commercial grade reactor internal pump (RIP). Recently, however, a licensing agreement with KSB of West Germany covering the RIP technology, has combined the know-how of KSB with Toshiba's technology to produce a truly high-quality prototype RIP. The pump produces the required coreflow for ABWR at low speed and with high efficiency, and simply by increasing the pump speed to the prior level, the coreflow can be further increased for such advantages as improved fuel cycle economy. Here, the advanced features and test results of the RIP are summarized. (author)

  11. Acoustic monitoring of the BOR-60 reactor circulating pump state

    International Nuclear Information System (INIS)

    Efimov, V.N.; Myntsov, A.A.

    1988-01-01

    Diagnostics methods for circulation pumps of the experimental BOR-60 fast reactor are described. The results of signal processing during a microcompain, as well as detected anomalies in pump operation in the earth stage are presented. Analysis carried out for an acoustic signal envelope has shown high efficiency of the method. When oscillations of a mechanical shaft are present, the envelope level increases 1.5 times. More detailed investigation is carried out by the analysis of the spectrum of the pump acoustic signal envelope. During abnormal operation there are peaks, corresponding to the circulation frequency, and harmonics multiple of it, in the spectrum. 6 figs

  12. Pump/heat exchanger assembly for pool-type reactor

    International Nuclear Information System (INIS)

    Nathenson, R.D.; Slepian, R.M.

    1987-01-01

    A heat exchanger and pump assembly comprising a heat exchanger including a housing for defining an annularly shaped cavity and supporting therein a plurality of heat transfer tubes. A pump is disposed beneath the heat exchanger and is comprised of a plurality of flow couplers disposed in a circular array. Each flow coupler is comprised of a pump duct for receiving a first electrically conductive fluid, i.e. the primary liquid metal, from a pool thereof, and a generator duct for receiving a second electrically conductive fluid, i.e. the intermediate liquid metal. The primary liquid metal is introduced from the reactor pool into the top, inlet ends of the tubes, flowing downward therethrough to be discharged from the tubes' bottom ends directly into the reactor pool. The primary liquid metal is variously introduced into the pump ducts directly from the reactor pool, either from the bottom or top end of the flow coupler. The intermediate fluid introduced into the generator ducts via the inlet duct and inlet plenum and after leaving the generator ducts passes through the annular cavity of the exchanger to cool the primary liquid in the tubes. The annular magnetic field of the pump is produced by a circular array of electromagnets having hollow windings cooled by a flow of the intermediate metal. (author)

  13. NCSU reactor sharing program. Final technical report

    International Nuclear Information System (INIS)

    Perez, P.B.

    1997-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities. This report is the Final Technical Report for the DOE award reference number DE-FG05-95NE38136 which covers the period September 30, 1995 through September 30, 1996

  14. Final-Independent Confirmatory Survey Report For The Reactor Building, Hot Laboratory, Primary Pump House, And Land Areas At The Plum Brook Reactor Facility, Sandusky, Ohio DCN:2036-SR-01-10

    International Nuclear Information System (INIS)

    Bailey, Erika N.

    2011-01-01

    In 1941, the War Department acquired approximately 9,000 acres of land near Sandusky, Ohio and constructed a munitions plant. The Plum Brook Ordnance Works Plant produced munitions, such as TNT, until the end of World War II. Following the war, the land remained idle until the National Advisory Committee for Aeronautics later called the National Aeronautics and Space Administration (NASA) obtained 500 acres to construct a nuclear research reactor designed to study the effects of radiation on materials used in space flight. The research reactor was put into operation in 1961 and was the first of fifteen test facilities eventually built by NASA at the Plum Brook Station. By 1963, NASA had acquired the remaining land at Plum Brook for these additional test facilities

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  16. Axis vibration detection device for reactor recycling pump

    International Nuclear Information System (INIS)

    Ide, Katsuki.

    1995-01-01

    The present invention provides a device for detecting, in a contactless manner, vibrations of a recycling pump shaft disposed in a reactor pressure vessel of a BWR type reactor. Namely, the vibration detector comprises an eddy current type displacement gauge having a sensing portion at one end of a linear tube type metal holder. It also comprises a rotational member made of an electroconductive material rotating integrally with a rotational pump shaft. The vibration detector is inserted into an attaching hole passing through a pump casing at a position where the sensing portion faces the outer circumference of the rotational member. The attaching hole is closed by a holder of the oscillation detector and a metal cap integrated to one end of the holder. A high pressure hermetic seal connector is disposed at a position outer side of the attaching hole of the vibration detector for electrically connecting the inside and the outside thereof. The device of the present invention can directly detect the vibration of the pump shaft. As a result, an abnormality, if should occur, in the recycling pump can be found in an early stage. Since the vibration detector is covered with a metal and shielded by the high pressure hermetic seal connector, it can sufficiently ensure pressure resistance. (I.S.)

  17. Reactor coolant pump shaft seal behavior during blackout conditions

    International Nuclear Information System (INIS)

    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

  18. Vacuum pumping of tritium in fusion power reactors

    International Nuclear Information System (INIS)

    Coffin, D.O.; Walthers, C.R.

    1979-01-01

    Compound cryopumps of three different designs will be tested with deuterium-tritium (DT) mixtures under simulated fusion reactor conditions at the Tritium Systems Test Assembly (TSTA) now being constructed at the Los Alamos Scientific Laboratory (LASL). The first of these pumps is already in operation, and its preliminary performance is presented. The supporting vacuum facility necessary to regenerate these fusion facility cryopumps is also described. The next generation of fusion system vacuum pumps may include non-cryogenic or conventional-cryogenic hybrid systems, several of which are discussed

  19. Design of Reactor Coolant Pump Seal Online Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Design of Reactor Coolant Pump Seal Online Monitoring System

    International Nuclear Information System (INIS)

    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

  1. Development of inertia-increased reactor internal pump

    International Nuclear Information System (INIS)

    Tanaka, Masaaki; Matsumura, Seiichi; Kikushima, Jun; Kawamura, Shinichi; Yamashita, Norimichi; Kurosaki, Toshikazu; Kondo, Takahisa

    2000-01-01

    The Reactor Internal Pump (RIP) was adopted for the Reactor Recirculation System (RRS) of Advanced Boiling Water Reactor (ABWR) plants, and ten RIPs are located at the bottom of the reactor pressure vessel. In order to simplify the power supply system for the RIPs, a new inertia-increased RIP was developed, which allows to eliminate the Motor-Generator (M-G) sets. The rotating inertia was increased approximately 2.5 times of current RIP inertia by addition of flywheel on its main shaft. A full scale proving test of the inertia-increased RIP under actual plant operating conditions using full scale test loop was performed to evaluate vibration characteristics and coast down characteristics. From the results of this proving test, the validity of the new inertia-increased RIP and its power supply system (without M-G sets) was confirmed. (author)

  2. Thermodynamic cycle calculations for a pumped gaseous core fission reactor

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Van Dam, H.

    1991-01-01

    Finite and 'infinitesimal' thermodynamic cycle calculations have been performed for a 'solid piston' model of a pumped Gaseous Core Fission Reactor with dissociating reactor gas, consisting of Uranium, Carbon and Fluorine ('UCF'). In the finite cycle calculations the influence has been investigated of several parameters on the thermodynamics of the system, especially on the attainable direct (nuclear to electrical) energy conversion efficiency. In order to facilitate the investigation of the influence of dissociation, a model gas, 'Modelium', was developed, which approximates, in a simplified, analytical way, the dissociation behaviour of the 'real' reactor gas. Comparison of the finite cycle calculation results with those of a so-called infinitesimal Otto cycle calculation leads to the conclusion that the conversion efficiency of a finite cycle can be predicted, without actually performing the finite cycle calculation, with reasonable accuracy, from the so-called 'infinitesimal efficiency factor', which is determined only by the thermodynamic properties of the reactor gas used. (author)

  3. Sodium pumping: pump problems

    International Nuclear Information System (INIS)

    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

  4. Aerosol core nuclear reactor for space-based high energy/power nuclear-pumped lasers

    International Nuclear Information System (INIS)

    Prelas, M.A.; Boody, F.P.; Zediker, M.S.

    1987-01-01

    An aerosol core reactor concept can overcome the efficiency and/or chemical activity problems of other fuel-reactant interface concepts. In the design of a laser using the nuclear energy for a photon-intermediate pumping scheme, several features of the aerosol core reactor concept are attractive. First, the photon-intermediate pumping concept coupled with photon concentration methods and the aerosol fuel can provide the high power densities required to drive high energy/power lasers efficiently (about 25 to 100 kW/cu cm). Secondly, the intermediate photons should have relatively large mean free paths in the aerosol fuel which will allow the concept to scale more favorably. Finally, the aerosol core reactor concept can use materials which should allow the system to operate at high temperatures. An excimer laser pumped by the photons created in the fluorescer driven by a self-critical aerosol core reactor would have reasonable dimensions (finite cylinder of height 245 cm and radius of 245 cm), reasonable laser energy (1 MJ in approximately a 1 millisecond pulse), and reasonable mass (21 kg uranium, 8280 kg moderator, 460 kg fluorescer, 450 kg laser medium, and 3233 kg reflector). 12 references

  5. Reactor coolant pump shaft seal behavior during station blackout

    International Nuclear Information System (INIS)

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

    1985-04-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  8. Innovative approaches to inertial confinement fusion reactors: Final report

    International Nuclear Information System (INIS)

    Bourque, R.F.; Schultz, K.R.

    1986-11-01

    Three areas of innovative approaches to inertial confinement fusion (ICF) reactor design are given. First, issues pertaining to the Cascade reactor concept are discussed. Then, several innovative concepts are presented which attempt to directly recover the blast energy from a fusion target. Finally, the Turbostar concept for direct recovery of that energy is evaluated. The Cascade issues discussed are combustion of the carbon granules in the event of air ingress, the use of alternate granule materials, and the effect of changes in carbon flow on details of the heat exchanger. Carbon combustion turns out to be a minor problem. Four ICF innovative concepts were considered: a turbine with ablating surfaces, a liquid piston system, a wave generator, and a resonating pump. In the final analysis, none show any real promise. The Turbostar concept of direct recovery is a very interesting idea and appeared technically viable. However, it shows no efficiency gain or any decrease in capital cost compared to reactors with conventional thermal conversion systems. Attempts to improve it by placing a close-in lithium sphere around the target to increase gas generation increased efficiency only slightly. It is concluded that these direct conversion techniques require thermalization of the x-ray and debris energy, and are Carnot limited. They therefore offer no advantage over existing and proposed methods of thermal energy conversion or direct electrical conversion

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

    International Nuclear Information System (INIS)

    Deng Shaowen

    1998-01-01

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

  10. In-operation diagnostic system for reactor coolant pump

    International Nuclear Information System (INIS)

    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)

  11. Reactor coolant pump shaft seal stability during station blackout

    International Nuclear Information System (INIS)

    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

  12. Reactor coolant pump shaft seal stability during station blackout

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. Modeling Results For the ITER Cryogenic Fore Pump. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Pfotenhauer, John M. [University of Wisconsin, Madison, WI (United States); Zhang, Dongsheng [University of Wisconsin, Madison, WI (United States)

    2014-03-31

    A numerical model characterizing the operation of a cryogenic fore-pump (CFP) for ITER has been developed at the University of Wisconsin – Madison during the period from March 15, 2011 through June 30, 2014. The purpose of the ITER-CFP is to separate hydrogen isotopes from helium gas, both making up the exhaust components from the ITER reactor. The model explicitly determines the amount of hydrogen that is captured by the supercritical-helium-cooled pump as a function of the inlet temperature of the supercritical helium, its flow rate, and the inlet conditions of the hydrogen gas flow. Furthermore the model computes the location and amount of hydrogen captured in the pump as a function of time. Throughout the model’s development, and as a calibration check for its results, it has been extensively compared with the measurements of a CFP prototype tested at Oak Ridge National Lab. The results of the model demonstrate that the quantity of captured hydrogen is very sensitive to the inlet temperature of the helium coolant on the outside of the cryopump. Furthermore, the model can be utilized to refine those tests, and suggests methods that could be incorporated in the testing to enhance the usefulness of the measured data.

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

    International Nuclear Information System (INIS)

    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

  15. Reactor coolant pump seal response to loss of cooling

    International Nuclear Information System (INIS)

    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)

  16. Station blackout with reactor coolant pump seal leakage

    International Nuclear Information System (INIS)

    Evinay, A.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  18. Reactor Coolant Pump Motor Maintenance Experience in Krsko NPP

    International Nuclear Information System (INIS)

    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).

  19. Replacement of the Pumps for Fuel Channel Cooling Circuit of the Maria Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krzysztoszek, G.; Mieleszczenko, W.; Moldysz, A. [National Centre for Nuclear Research, Otwock–Świerk (Poland)

    2014-08-15

    The high flux Maria research reactor is operated by the National Centre for Nuclear Research in Świerk. It is a pool type reactor with pressurized fuel channels located in the beryllium matrix. According to the Global Threat Reduction Initiative programme our goal is to convert the Maria reactor from HEU to LEU fuel. Hydraulic losses in the new LEU fuel produced by CERCA are about 30% higher than the existing HEU fuel of type MR-6. For the MR-6 fuel were installed four two speed pumps. These pumps performed the function of the main circulations pumps during reactor operation with residual pumping power provided by emergency pumps. In the new system four main pumps will be used for circulating coolant while the reactor is operation with three auxiliary pumps for decay heat removal after reactor shutdown, meaning that the conversion of Maria research reactor will be possible after increasing flow in the primary cooling circuit of the fuel channels. The technical design of replacement of the pumps in the primary fuel channel cooling circuit was finished in April 2011 and accepted by the Safety Committee. After delivery of the new pumps we are planning to upgrade the primary fuel channel cooling circuit during October–November 2012. (author)

  20. Iowa Hill Pumped Storage Project Investigations - Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, David [Sacramento Municipal Unitlity District, Sacramento, CA (United States)

    2016-07-01

    This Final Technical Report is a summary of the activities and outcome of the Department of Energy (DOE) Assistance Agreement DE-EE0005414 with the Sacramento Municipal Utility District (SMUD). The Assistance Agreement was created in 2012 to support investigations into the Iowa Hill Pumped-storage Project (Project), a new development that would add an additional 400 MW of capacity to SMUD’s existing 688MW Upper American River Hydroelectric Project (UARP) in the Sierra Nevada mountains east of Sacramento, California.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core. 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. The paper 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 motorgenerator set.

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

    International Nuclear Information System (INIS)

    Loenhout, Gerard van; Hurni, Juerg

    2014-01-01

    A modern reactor recirculation pump circulates a large volume of high temperature, very pure water from the reactor pressure vessel back to the core. 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. The paper 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 motorgenerator set.

  5. Leak rate analysis of the Westinghouse Reactor Coolant Pump

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Earthquake-proof support structures for the recycling pump in FBR type reactors

    International Nuclear Information System (INIS)

    Nakagawa, Masaki; Shigeta, Masayuki.

    1984-01-01

    Purpose: To improve the earthquake proofness of the recycling pump for use in FBR type reactors upon earthquake by reducing the vibration response of the pump. Constitution: The outer casing of a recycle pump suspended into liquid sodium is extended to the portion that penetrates a reactor core support structures. Support structures surrounding the outer side of the recycling pump are disposed with a gap not restraining the free thermal deformations of the recycling pump to the inside of the partition wall structures and the portion of the recycling pump penetrating the reator core support structures, to integrate the support structures with the reactor core support structures. Accordingly, there are no interferences between the recycling pump and the support structures with respect to the thermal deformations that change gradually with time. Upon vibrating under the rapidly changing external forces of earthquakes, however, the pressure resulted to the liquid in the gap due to the vibrations of the recycling pump is transmitted with no escape to the support structures, the recycling pump and the support structures integrally resist the vibrations thereby enabling to reduce the vibrations in the recycling pumps. (Horiuchi, T.)

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

    International Nuclear Information System (INIS)

    McNulty, A.L.; Cheshire, A.

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  12. A drift-pump coil design for a Tandem Mirror Reactor

    International Nuclear Information System (INIS)

    Neef, W.S.; Logan, B.

    1983-01-01

    This paper describes both the theory and mechanical design behind a new concept for trapped ion removal from tandem mirror end plugs. The design has been developed for the Mirror Advanced Reactor Study (MARS). The new drift-pump coils replace charge exchange pump beams. Pump beams consume large amounts of power and seriously reduce reactor performance. Drift-pump coils consume only a few megawatts of power and introduce no added burden to the reactor vacuum pumps. In addition, they are easy to replace. The coils are similar in shape to a paper clip and are located at two positions in each end plug. The coils between the transition coil and the first anchor yinyang serve to remove ions trapped in the magnetic well just outboard of the high field choke coil. The coils located between the anchor coil set and the plug coil set remove sloshing ions and trapped cold ions from the plug region

  13. Biodenitrification in Sequencing Batch Reactors. Final report

    International Nuclear Information System (INIS)

    Silverstein, J.

    1996-01-01

    One plan for stabilization of the Solar Pond waters and sludges at Rocky Flats Plant (RFP), is evaporation and cement solidification of the salts to stabilize heavy metals and radionuclides for land disposal as low-level mixed waste. It has been reported that nitrate (NO 3- ) salts may interfere with cement stabilization of heavy metals and radionuclides. Therefore, biological nitrate removal (denitrification) may be an important pretreatment for the Solar Pond wastewaters at RFP, improving the stability of the cement final waste form, reducing the requirement for cement (or pozzolan) additives and reducing the volume of cemented low-level mixed waste requiring ultimate disposal. A laboratory investigation of the performance of the Sequencing Batch Reactor (SBR) activated sludge process developed for nitrate removal from a synthetic brine typical of the high-nitrate and high-salinity wastewaters in the Solar Ponds at Rocky Flats Plant was carried out at the Environmental Engineering labs at the University of Colorado, Boulder, between May 1, 1994 and October 1, 1995

  14. The development of a cavitation free sodium pump for the breeder reactor

    International Nuclear Information System (INIS)

    Baladi, J.Y.; Nyilas, C.P.

    1986-12-01

    The sodium pumps for a liquid metal fast breeder reactor must be designed for exceptionally high reliability and long life. The principal adverse factor which tends to limit the primary pump life is cavitation which becomes potentially severe under off-design flow conditions caused by the requirement of two loop operations which resulted in a large operating flow range. This problem prompted an extensive study which included experimental investigations of scaled down and full size pumps. The investigations involved visual observations, acoustic signature recordings, and physical characteristic measurements of the model and full size impellers. The blade configuration of the model was modified several times. After each modification intensive testing was conducted with feedback to established design criteria. The results obtained from the final configuration showed excellent cavitation performance. This configuration was then machined on the full scale impeller and tested. The results confirmed acceptable performance in the entire range of operating conditions. This paper describes the test facilities erected for this study, discusses the experimental techniques employed, and presents the experimental techniques employed, and presents a sample of the experimental results

  15. Final project report: TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1993-02-01

    This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project including, Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 M 3 of mixed waste

  16. Final project report, TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1992-01-01

    This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project, including Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 m 3 of low-level solid radioactive waste and 5 m 3 of mixed waste

  17. Evaluation of a nonevaporable getter pump for tritium handling in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Singleton, M.F.; Griffith, C.M.

    1978-01-01

    Lawrence Livermore Laboratory has tested and evaluated a commercially available getter pump for use with tritium in the Tokamak Fusion Test Reactor (TFTR). The pump contains Zr(84%)--Al in cartridge form with a concentric heating unit. It performed well in all tests, except for frequent heater failures

  18. Pump selection and application in a pressurized water reactor electric generating plant

    International Nuclear Information System (INIS)

    Kitch, D.M.

    1985-01-01

    Various pump applications utilized in a nuclear pressurized water reactor electric generating plant are described. Emphasis is on pumps installed in the auxiliary systems of the primary nuclear steam supply system. Hydraulic and mechanical details, the ASME Code (Nuclear Design), materials, mechanical seals, shaft design, seismic qualification, and testing are addressed

  19. First domestic primary loop recircuration pump for boiling water reactor

    International Nuclear Information System (INIS)

    Fukuda, Minoru; Taka, Shusei; Kato, Hiroyuki

    1981-01-01

    Two primary loop recirculation (PLR) pumps for the second unit of the Fukushima No. 2 Nuclear Power Station of the Tokyo Electric Power Co., Inc., have been manufactured by Ebara Corporation. They are the first domestically produced pumps for commercial power plants and were manufactured under license from Byron Jackson Pump Division of Borg Warner Corporation. This article describes the special features of pump design and stress analysis, and the results of the 700 hours of factory loop tests, which are all essential for the PLR pump. (author)

  20. 241-SY-101 mixer pump lifetime expectancy. Final report

    International Nuclear Information System (INIS)

    Shaw, C.P.

    1995-01-01

    The purpose of WHC-SD-WM-TI-726, Rev. 0 241-SY-101 Mixer Pump Lifetime Expectancy is to determine a best estimate of the mean lifetime of non-repairable (located in the waste) essential features of the hydrogen mitigation mixer pump presently installed in 101-SY. The estimated mean lifetime is 9.1 years. This report does not demonstrate operation of the entire pump assembly within the Tank Farm ''safety envelope''. It was recognized by the Defense Nuclear Facilities Safety Board (DNFSB) this test pump was not specifically designed for long term service in tank 101-SY. In June 95 the DNFSB visited Hanford and ask the question, ''how long will this test pump last and how will the essential features fail?'' During the 2 day meeting with the DNFSB it was discussed and defined within the meeting just exactly what essential features of the pump must operate. These essential features would allow the pump to operate for the purpose of extending the window for replacement. Operating with only essential features would definitely be outside the operating safety envelope and would require a waiver. There are three essential features: 1. The pump itself (i.e. the impeller and motor) must operate 2. Nozzles and discharges leg must remain unplugged 3. The pump can be re-aimed, new waste targeted, even if manually

  1. Pump

    International Nuclear Information System (INIS)

    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)

  2. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Reactor vessel decommissioning project. Final report

    International Nuclear Information System (INIS)

    Schoonen, D.H.

    1984-09-01

    This report describes a reactor vessel decommissioning project; it documents and explains the project objectives, scope, performance results, and sodium removal process. The project was successfully completed in FY-1983, within budget and without significant problems or adverse impact on the environment. Waste generated by the operation included the reactor vessel, drained sodium, and liquid, solid, and gaseous wastes which were significantly less than project estimates. Personnel radiation exposures were minimized, such that the project total was one-half the predicted exposure level. Except for the sodium removed, the material remaining in the reactor vessel is essentially the same as when the vessel arrived for processing

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

    International Nuclear Information System (INIS)

    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

  5. Earthquake resistance of residual heat removed (RHR) pump for pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    Uga, Takeo; Shiraki, K.; Honma, T.; Matsubayashi, H.; Inazuka, H.

    1980-01-01

    The present paper deals with the earthquake resistance of the residual heat removed (RHR) pump of single stage double volute type, which is one of the structurally simplest pumps used for pressurized water reactors (PWR). The results of the study can be summarized as follows: (1) Any trouble which can give effect on the functions of the pump at earthquake does not become a problem so long as each part of the pump is of aseismatically rigid structure. (2) Aseismatic tolerance test in the pump's operating condition has shown that the earthquake resistance of the pump at its location has a tolerance about five times the dynamic design acceleration. (3) The pump is provided with an impeller-casing wear ring at the pressure boundary between the suction side pressure and discharge side pressure. This wear ring acts as an underwater bearing when the pump is in operation, and improves the vibration characteristics, particularly damping ratio, of the pump shaft to a great extent to make the pump more aseismatic. (4) In the evaluation of the underwater bearing characteristics of the wear ring, the evaluation accuracy of the vibration characteristics of the pump shaft can be improved by taking into consideration the pressure loss in the wear ring part from the head of the single stage of the pump due to the rotation of the impeller. (author)

  6. About a fuel for burnup reactor of periodical pulsed nuclear pumped laser

    International Nuclear Information System (INIS)

    Volkov, A.I.; Lukin, A.V.; Magda, L.E.; Magda, E.P.; Pogrebov, I.S.; Putnikov, I.S.; Khmelnitsky, D.V.; Scherbakov, A.P.

    1998-01-01

    A physical scheme of burnup reactor for a Periodic Pulsed Nuclear Pumped Laser was supposed. Calculations of its neutron physical parameters were made. The general layout and construction of basic elements of the reactor are discussed. The requirements for the fuel and fuel elements are established. (author)

  7. Design and Selection of Innovative Primary Circulation Pumps for GEN-IV Lead Fast Reactors

    Directory of Open Access Journals (Sweden)

    Walter Borreani

    2017-12-01

    Full Text Available Although Lead-cooled Fast Reactor (LFR is not a new concept, it continues to be an example of innovation in the nuclear field. Recently, there has been strong interest in liquid lead (Pb or liquid lead–bismuth eutectic (LBE both critical and subcritical systems in a relevant number of Countries, including studies performed in the frame of GENERATION-IV initiative. In this paper, the theoretical and computational findings for three different designs of Primary Circulation Pump (PCP evolving liquid lead (namely the jet pump, the Archimedean pump and the blade pump are presented with reference to the ALFRED (Advanced Lead Fast Reactor European Demonstrator design. The pumps are first analyzed from the theoretical point of view and then modeled with a 3D CFD code. Required design performance of the pumps are approximatively around an effective head of 2 bar with a mass flow rate of 5000 kg/s. Taking into account the geometrical constraints of the reactor and the fluid dynamics characteristics of the molten lead, the maximum design velocity for molten lead fluid flow of 2 m/s may be exceeded giving rise to unacceptable erosion phenomena of the blade or rotating component of the primary pumping system. For this reason a deep investigation of non-conventional axial pumps has been performed. The results presented shows that the design of the jet pump looks like beyond the current technological feasibility while, once the mechanical challenges of the Archimedean (screw pump and the fluid-dynamic issues of the blade pump will be addressed, both could represent viable solutions as PCP for ALFRED. Particularly, the blade pump shows the best performance in terms of pressure head generated in normal operation conditions as well as pressure drop in locked rotor conditions. Further optimizations (mainly for what the geometrical configuration is concerned are still necessary.

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

    International Nuclear Information System (INIS)

    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

  9. Main boiler feed pump for fast breeder test reactor. Failure analysis and remedial measures

    International Nuclear Information System (INIS)

    Iyer, M.A.K.; Chande, S.K.; Raghuvir, A.D.; Baskar, S.; Kale, R.D.

    1994-01-01

    A small capacity ten stage 670 kw feed water pump is used for supplying feed water at a temperature of 190 deg C to a once through steam generator in the Fast Breeder Test Reactor at Kalpakkam. During preparatory heating up stage to commission the steam generator the pump suffered a severe loss of suction which resulted in failure of hydrostatic journal bearings and extensive damage to pump internals. This paper discusses the detailed mechanism of loss of suction, details of damage to the pump and various modifications carried out to prevent recurrence of the problem. (author). 4 refs., 3 figs., 2 tabs

  10. Prometheus Project Reactor Module Final Report, For Naval Reactors Information

    International Nuclear Information System (INIS)

    MJ Wollman; MJ Zika

    2006-01-01

    The Naval Reactors Prime Contractor Team (NRPCT) led the development of a power plant for a civilian nuclear electric propulsion (NEP) system concept as part of the Prometheus Project. This report provides a summary of the facts, technical insights, and programmatic perspectives gained from this two-year program. The Prometheus Project experience has been extensively documented to better position the US for future space reactor development. Major Technological and engineering challenges exist to develop a system that provides useful electric power from a nuclear fission heat source operating in deep space. General issues include meeting mission requirements in a system that has a mass low enough to launch from earth while assuring public safety and remaining safely shutdown during credible launch accidents. These challenges may be overcome in the future if there is a space mission with a compelling need for nuclear power to drive development. Past experience and notional mission requirements indicate that any useful space reactor system will be unlike past space reactors and existing terrestrial reactors

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

    International Nuclear Information System (INIS)

    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.)

  12. TMI-2 reactor vessel plenum final lift

    International Nuclear Information System (INIS)

    Wilson, D.C.

    1986-01-01

    Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

  13. Fusion reactor remote maintenance study. Final report

    International Nuclear Information System (INIS)

    Sniderman, M.

    1979-04-01

    An analysis of a major maintenance operation, the remote replacement of a modular sector of a tokamak reactor, was performed in substantial detail. Specific assumptions were developed which included concepts from various existing designs so that the operation which was studied includes some design features generic to any fusion reactor design. Based on the work performed in this study, the principal conclusions are: (1) It appears feasible to design a tokamak fusion reactor plant with availability comparable to existing fossil and fission plants, but this will require diligence and comprehensive planning during the complete design phase. (2) Since the total fusion program is paced by the success of each device, maintenance considerations must be incorporated into each device during design, even if the device is an experimental unit. (3) Innovative approaches, such as automatic computer controlled operations, should be developed so that large step reductions in planned maintenance times can be achieved

  14. Reactor kinetics methods development. Final report

    International Nuclear Information System (INIS)

    Hansen, K.F.; Henry, A.F.

    1978-01-01

    This report is a qualitative summary of research conducted at MIT from 1967 to 1977 in the area of reactor kinetics methods. The objectives of the research were to find methods of integration of various mathematical models of nuclear reactor transients. From the beginning the work was aimed at numerical integration methods. Specific areas of research, discussed in more detail following, included: integration of multigroup diffusion theory models by finite difference and finite element methods; response matrix and nodal methods; coarse-mesh homogenization; and special treatment of boundary conditions

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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. Heat pumps for geothermal applications: availability and performance. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Reistad, G.M.; Means, P.

    1980-05-01

    A study of the performance and availability of water-source heat pumps was carried out. The primary purposes were to obtain the necessary basic information required for proper evaluation of the role of water-source heat pumps in geothermal energy utilization and/or to identify the research needed to provide this information. The Search of Relevant Literature considers the historical background, applications, achieved and projected performance evaluations and performance improvement techniques. The commercial water-source heat pump industry is considered in regard to both the present and projected availability and performance of units. Performance evaluations are made for units that use standard components but are redesigned for use in geothermal heating.

  19. Internal fluid flow management analysis for Clinch River Breeder Reactor Plant sodium pumps

    International Nuclear Information System (INIS)

    Cho, S.M.; Zury, H.L.; Cook, M.E.; Fair, C.E.

    1978-12-01

    The Clinch River Breeder Reactor Plant (CRBRP) sodium pumps are currently being designed and the prototype unit is being fabricated. In the design of these large-scale pumps for elevated temperature Liquid Metal Fast Breeder Reactor (LMFBR) service, one major design consideration is the response of the critical parts to severe thermal transients. A detailed internal fluid flow distribution analysis has been performed using a computer code HAFMAT, which solves a network of fluid flow paths. The results of the analytical approach are then compared to the test data obtained on a half-scale pump model which was tested in water. The details are presented of pump internal hydraulic analysis, and test and evaluation of the half-scale model test results

  20. Rock bed storage with heat pump. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Remmers, H.E.; Mills, G.L.

    1979-05-01

    The study, Rock Bed Storage with Heat Pump, established the feasibility of mating a heat pump to a rock bed storage to effect optimal performance at the lowest cost in single family residences. The operating characteristics of off-the-shelf components of heat pump/rock bed storage systems were studied, and the results were used to formulate configurations of representative systems. These systems were modeled and subsequently analyzed using the TRNSYS computer program and a life cycle cost analysis program called LCCA. A detailed load model of a baseline house was formulated as part of the TRNSYS analysis. Results of the analysis involved the development of a technique to confine the range of heat pump/rock bed storage systems to those systems which are economical for a specific location and set of economic conditions. Additionally, the results included a comparison of the detailed load model with simple UA models such as the ASHRAE bin method. Several modifications and additions were made to the TRNSYS and LCCA computer programs during the course of the study.

  1. Emergency cooling system with hot-water jet pumps for nuclear reactors

    International Nuclear Information System (INIS)

    Reinsch, A.O.W.

    1977-01-01

    The ECCS for a PWR or BWR uses hot-water jet pumps to remove the thermal energy generated in the reactor vessel and stored in the water. The hot water expands in the nozzle part (Laval nozzle) of the jet pump and sucks in coolant (borated water) coming from a storage tank containing subcooled water. This water is mixing with the hot water/steam mixture from the Laval nozzle. The steam is condensed. The kinetic energy of the water is converted into a pressure increase which is sufficient to feed the water into the reactor vessel. The emergency cooling may further be helped by a jet condenser also operating according to the principle of a jet pump and condensing the steam generated in the reactor vessel. (DG) [de

  2. Integrated main coolant pumps for pressurized-water reactors

    International Nuclear Information System (INIS)

    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

  3. Pump/heat exchanger assembly for pool-type reactor

    International Nuclear Information System (INIS)

    Nathenson, R.D.; Slepian, R.M.

    1989-01-01

    This patent describes a heat exchanger and pump assembly for transferring thermal energy from a heated, first electrically conductive fluid to a pumped, second electrically conductive fluid and for transferring internal energy from the pumped, second electrically conductive fluid to the first electrically conductive fluid, the assembly adapted to be disposed within a pool of the first electrically conductive fluid and comprising: a heat exchanger comprising means for defining a first annularly shaped cavity for receiving a flow of the second electrically conductive fluid and a plurality of tubes disposed within the cavity, whereby the second electrically conductive fluid in the cavity is heated, each of the tubes having an input and an output end. The input ends being disposed at the top of the heat exchanger for receiving from the pool a flow of the first electrically conductive fluid therein. The output ends being disposed at the bottom of and free of the cavity defining means for discharging the first electrically conductive fluid directly into the pool; a pump disposed beneath the heat exchanger and comprised of a plurality of flow couplers disposed in a circular array, each flow coupler comprised of a pump duct for receiving the first electrically conductive fluid and a generator duct for receiving the second electrically conductive fluid

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

    International Nuclear Information System (INIS)

    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.

  5. Thermionic conversion reactor technology assessment. Final report

    International Nuclear Information System (INIS)

    1984-02-01

    The in-core thermionic space nuclear power supply may be the only identified reactor-power concept that can meet the SP-100 size functional requirements with demonstrated state-of-the-art reactor system and space-qualified power system component temperatures. The SP-100 configuration limits provide a net 40 m 2 of primary non-deployed radiator area. If a reasonable 7-year degradation allowance of 15% to 20% is provided then the beginning of life (BOL) net power output requirement is about 120 kWe. Consequently, the SP-100 power system must produce a P/A of 2.7 kWe/m 2 . This non-deployed radiator area power density performance can only be reasonably achieved by the thermionic in-core convertr system, the potassium Rankine turbine system and the Stirling engine system. The purpose of this study is to examine past and current tests and data, and to assess the potential for successful development of suitable fueled-thermionic converters that will meet SP-100 and growth requirements. The basis for the assessment will be provided and the recommended key developments plan set forth

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  10. Optimization of reactor power by taking into consideration temperature increase in a reactor pumped 3He-Xe laser

    International Nuclear Information System (INIS)

    Cetin, Fuesun

    2009-01-01

    In nuclear pumped lasers, gas parameters are optimized in a manner such that output power is increased for constructing a high power laser. Since output power increases with the increase of energy deposited in the gas, high output power requires high pumping power. However, the high energy loading results in elevated gas temperature. Temperature increase of this magnitude can detrimentally influence the laser gain and efficiency, since it negatively impacts several important laser kinetic.processes. This fact may cause laser output to abruptly terminate before the peak of the pump pulse [1-3]. A nuclear pumped laser using a volumetric energy source through the 3 He(n, p) 3 H reaction has here been considered. It is assumed that TRIGA Mark II Reactor at Istanbul Technical University is used for nuclear pumping as the neutron source. In the previous papers, the optimal parameters for improving both output power and optical homogeneity were determined [4-5]. Spatial and temporal variations of gas temperature during pumping pulse for maximum peak power (1200 MW) were determined for various operating pressures in Ref. [6]. It was seen that gas temperature reaches up to 1000 0 K near the peak of the pumping pulse for the initial pressures of 1-4 atm. This means that laser output may terminate before the peak of the pump pulse due to overheating of laser gas. Under these conditions, a question arises about a further optimisation taking into consideration gas temperature. This question has been examined in this study. Experimental results (Batyrbekov et al, 1989) showed that temperature rise up to 650 C had no influence on Xe laser characteristics [ 7]. Therefore, It has here been assumed that the lasing will terminate when gas temperature reaches 1000 0 K for a Xe-laser with 3 He buffer gas. Under these conditions optimum reactor power is investigated by taking into consideration lasing duration also. (orig.)

  11. Constructional features of an integrated centrifugal pump in BWR reactors

    International Nuclear Information System (INIS)

    Mueller, H.D.; Schneider, W.

    1981-01-01

    The pump construction described here has been produced as a prototype and tested at full load. From trial runs with hot and cold water, important operating data were obtained, e.g. throughput, lift, vibrations, loads on the radial bearings, cooling water temperatures, etc. (orig./GL) [de

  12. Steady-state and loss-of-pumping accident analyses of the Savannah River new production reactor representative design

    International Nuclear Information System (INIS)

    Pryor, R.J.; Maloney, K.J.

    1990-10-01

    This document contains the steady-state and loss-of-pumping accident analysis of the representative design for the Savannah River heavy water new production reactor. A description of the reactor system and computer input model, the results of the steady-state analysis, and the results of four loss-of-pumping accident calculations are presented. 5 refs., 37 figs., 4 tabs

  13. Study of an electromagnetic pump in a sodium cooled reactor. Design study of secondary sodium main pumps (Joint research)

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Kisohara, Naoyuki; Hishida, Masahiko; Fujii, Tadashi; Konomura, Mamoru; Ara, Kuniaki; Hori, Toru; Uchida, Akihito; Nishiguchi, Youhei; Nibe, Nobuaki

    2006-07-01

    In the feasibility study on commercialized fast breeder cycle system, a medium scale sodium cooled reactor with 750 MW electricity has been designed. In this study, EMPs are applied to the secondary sodium main pump. The EMPs type is selected to be an annular linear induction pump (ALIP) type with double stators which is used in the 160 m 3 /min EMP demonstration test. The inner structure and electromagnetic features are decided reviewing the 160 m 3 /min EMP. Two dimensional electromagnetic fluid analyses by EAGLE code show that Rms (magnetic Reynolds number times slip) is evaluated to be 1.08 which is less than the stability limit 1.4 confirmed by the 160 m 3 /min EMP test, and the instability of the pump head is evaluated to be 3% of the normal operating pump head. Since the EMP stators are cooled by contacting coolant sodium duct, reliability of the inner structures are confirmed by temperature distribution and stator-duct contact pressure analyses. Besides, a power supply system, maintenance and repair feature and R and D plan of EMP are reported. (author)

  14. Low-power photolytically pumped lasers: Final technical report

    International Nuclear Information System (INIS)

    Messing, I.; Lorents, D.C.; Eckstrom, D.J.

    1987-08-01

    We have carried out an extensive series of measurements of the time-resolved Xe 2 * emission spectra following optical pumping by a short-pulse F 2 laser at 157.6 nm. Most measurements were performed using a gated Optical Multichannel Analyzer detector; we also made measurements using a scanning monochromator fitted with a photomultiplier and using a boxcar integrator for time resolution. The two sets of results agree well and show that both the singlet and triplet emission bands are broader than expected and have center wavelengths closer together than expected. Measurements were performed both at room temperature and at elevated (140 0 C) and reduced (-27 0 C) temperatures. The broad bandwidth of the individual spectral bands was unexpected and conflicted with a previous spectral measurement using optical pumping by the Xe* resonance line from a microwave discharge lamp. Therefore, we also performed a series of spectral measurements using this type of optical pumping. We achieved good agreement with some previous results in the literature, but not with the result in question. We conclude that the present results are reliable. The results presented in this report provide the first definitive measurement of the individual excimer emissions from each of the Xe 2 (0/sub u/ + ) and Xe 2 (1/sub u/) states. From these measurements and the known ground state potential, we derived a 1/sub u/ potential that reproduces the emission band very well. However, the 1/sub u/ potential is in substantial disagreement with the recent 1/sub u/ potential derived by the Toronto group. 13 refs., 32 figs., 3 tabs

  15. Strategies for reactor safety. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, K

    1997-12-01

    The NKS/RAK-1 project formed part of a four-year nuclear research program (1994-1997) in the Nordic countries, the NKS Programme. The project aims were to investigate and evaluate the safety work, to increase realism and reliability of the safety analysis, and to give ideas for how safety can be improved in selected areas. An evaluation of the safety work in nuclear installations in Finland and Sweden was made, and a special effort was devoted to plant modernisation and to see how modern safety standards can be met up with. A combination of more resources and higher efficiency is recommended to meet requirements from plant modernisation and plant renovations. Both the utilities and the safety authorities are recommended to actively follow the evolving safety standards for new reactors. Various approaches to estimating LOCA frequencies have been explored. In particular, a probabilistic model for pipe ruptures due to intergranular stress corrosion has been developed. A survey has been done over methodologies for integrated sequence analysis (ISA), and different approaches have been developed and tested on four sequences. Structured frameworks for integration between PSA and behavioural sciences have been developed, which e.g. have improved PSA. The status of maintenance strategies in Finland and Sweden has been studied and a new maintenance data information system has been developed. (au) 41 refs.

  16. Strategies for reactor safety. Final report

    International Nuclear Information System (INIS)

    Andersson, K.

    1997-12-01

    The NKS/RAK-1 project formed part of a four-year nuclear research program (1994-1997) in the Nordic countries, the NKS Programme. The project aims were to investigate and evaluate the safety work, to increase realism and reliability of the safety analysis, and to give ideas for how safety can be improved in selected areas. An evaluation of the safety work in nuclear installations in Finland and Sweden was made, and a special effort was devoted to plant modernisation and to see how modern safety standards can be met up with. A combination of more resources and higher efficiency is recommended to meet requirements from plant modernisation and plant renovations. Both the utilities and the safety authorities are recommended to actively follow the evolving safety standards for new reactors. Various approaches to estimating LOCA frequencies have been explored. In particular, a probabilistic model for pipe ruptures due to intergranular stress corrosion has been developed. A survey has been done over methodologies for integrated sequence analysis (ISA), and different approaches have been developed and tested on four sequences. Structured frameworks for integration between PSA and behavioural sciences have been developed, which e.g. have improved PSA. The status of maintenance strategies in Finland and Sweden has been studied and a new maintenance data information system has been developed. (au)

  17. Final Remediation Report for the K-Area Bingham Pump Outage Pit (643-1G); FINAL

    International Nuclear Information System (INIS)

    Morganstern, M.

    2002-01-01

    The K-Area Bingham Pump Outage Pit (K BPOP) Building Number 643-1G, is situated immediately south and outside the K-Reactor fence line and is approximately 400 feet in length and 60 feet in width. For the K BPOP operable unit, the Land Use Control (LUC) objectives are to prevent contact, removal, or excavation of buried waste in the area and to preclude residential use of the area

  18. Housing maintenance of primary cooling pump I of Kartini reactor

    International Nuclear Information System (INIS)

    Agung Nugroho; Wahyu Imam W

    2013-01-01

    Housing maintenance of Primary Cooling Pump have been done with purpose to enhance capability of fluid block and stopping leakage. The procedures of modification are follow: Replace mechanical seal type Nock 560 – 1 ½'' by mechanical seal type Nock 560 - 38, modification gland plate/housing with to lathe the diameter of gland plate/housing from diameter of 54 mm to 60 mm, in size setting-up mechanical seal, alignment and then function test. The result of the modification are: mechanical seal has been installed, housing has been modified, and the leakage of primary cooling water has been repaired and operated properly. Conclusion of the maintenance are the primary cooling water pump is working well, because the primary cooling water is not leakage any more. (author)

  19. Study on operation of a research reactor during one PCS pump failure accident

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Kyoung Woo; Yoon, Hyu Ngi; Kim, Seong Hoon; Chi, Dae Young; Yoon, Juh Yeon [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The Primary cooling system (PCS) of a research reactor is designed to provide adequate cooling to the reactor core with a reasonable margin during all operation modes. The PCS consists of pumps, heat exchangers, and all necessary interconnecting pipes, valves, and instruments. The number of pumps is determined from a safety and economic point of view. As the number of pump trains increase, the cost increases according to the increase in safety class equipment. However, it is impossible to install one pump for a PCS because a zero flow can instantaneously occur during a pump failure such as a pump seizure. Thus, a PCS frequently consists of two parallel 50% capacity pumps and heat exchangers. In addition, check valves are generally installed to prevent a reversal flow when multiple pumps are designed to operate. However, if a swing type check valve is used, it should be estimated whether the slam due to instantaneous closing of the valve affects the system vibration. To reduce the vibration by a slam phenomenon, additional equipment such as a damper will be installed in the valve. The purpose of the check valve in PCS is to prevent the flow path when a reverse flow occurs. The installation of additional equipment will make it difficult to perform this function. In this study, it is estimated whether the PCS can operate without check valves. First, a flow analysis using Flowmaster was compared and verified by the calculation employing a empirical correlation. Second, the simulation for a one pump failure accident was performed and analyzed.

  20. Study on operation of a research reactor during one PCS pump failure accident

    International Nuclear Information System (INIS)

    Seo, Kyoung Woo; Yoon, Hyu Ngi; Kim, Seong Hoon; Chi, Dae Young; Yoon, Juh Yeon

    2012-01-01

    The Primary cooling system (PCS) of a research reactor is designed to provide adequate cooling to the reactor core with a reasonable margin during all operation modes. The PCS consists of pumps, heat exchangers, and all necessary interconnecting pipes, valves, and instruments. The number of pumps is determined from a safety and economic point of view. As the number of pump trains increase, the cost increases according to the increase in safety class equipment. However, it is impossible to install one pump for a PCS because a zero flow can instantaneously occur during a pump failure such as a pump seizure. Thus, a PCS frequently consists of two parallel 50% capacity pumps and heat exchangers. In addition, check valves are generally installed to prevent a reversal flow when multiple pumps are designed to operate. However, if a swing type check valve is used, it should be estimated whether the slam due to instantaneous closing of the valve affects the system vibration. To reduce the vibration by a slam phenomenon, additional equipment such as a damper will be installed in the valve. The purpose of the check valve in PCS is to prevent the flow path when a reverse flow occurs. The installation of additional equipment will make it difficult to perform this function. In this study, it is estimated whether the PCS can operate without check valves. First, a flow analysis using Flowmaster was compared and verified by the calculation employing a empirical correlation. Second, the simulation for a one pump failure accident was performed and analyzed

  1. Pumps of molten metal based on magnetohydrodynamicprinciple for cooling high-temperature nuclear reactors

    Czech Academy of Sciences Publication Activity Database

    Doležel, Ivo; Donátová, M.; Karban, P.; Ulrych, B.

    2009-01-01

    Roč. 85, č. 4 (2009), s. 13-15 ISSN 0033-2097 R&D Projects: GA ČR(CZ) GA102/07/0496 Institutional research plan: CEZ:AV0Z20570509 Keywords : pumps of molten metal * magnetohydrodynamic principle * nuclear reactors Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 0.196, year: 2009

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

    International Nuclear Information System (INIS)

    Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

    2015-01-01

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

  3. Feedback control of primary circulation pump of PIUS-Type reactor

    International Nuclear Information System (INIS)

    Fujii, Mikiya; Anoda, Yoshinari; Murata, Hideo; Yonomoto, Taisuke; Kukita, Yutaka; Tasaka, Kanji.

    1991-05-01

    In operating the PIUS-Type reactor, it is required to keep stationary density interfaces between the primary loop hot water and the poison tank cold, borated water by maintaining pressure balance between the primary-loop and the poison-tank. The authors have developed a primary circulation pump speed control system and tested it in small-scale experiments. This control system regulates the pump speed based on measurements of the density lock differential pressure which is proportional to the elevation of the interface in the density lock. This pump speed control facilitated the normal plant operation which included core power changes. However, the elevation of the density interface indicated oscillatory behavior when the pump speed was regulated as a linear function of the density lock differential pressure. The mechanism responsible for such oscillatory behavior was found to be manometric oscillations that could be eliminated by adding a damping term to compensate for the mechanical delay of the primary pump speed. The passive shutdown function of the reactor was retained by setting an upper limit to the pump speed. This was confirmed in a loss-of-feedwater abnormal transient test. (author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    OpenAIRE

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

    2012-01-01

    In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspec...

  7. Commercialization of PV-powered pumping systems for use in utility PV service programs. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    The project described in this report was a commercialization effort focused on cost-effective remote water pumping systems for use in utility-based photovoltaic (PV) service programs. The project combined a commercialization strategy tailored specifically for electric utilities with the development of a PV-powered pumping system that operates conventional ac pumps rather than relying on the more expensive and less reliable PV pumps on the market. By combining these two attributes, a project goal was established of creating sustained utility purchases of 250 PV-powered water pumping systems per year. The results of each of these tasks are presented in two parts contained in this Final Summary Report. The first part summarizes the results of the Photovoltaic Services Network (PSN) as a new business venture, while the second part summarizes the results of the Golden Photon system installations. Specifically, results and photographs from each of the system installations are presented in this latter part.

  8. WEXA: exergy analysis for increasing the efficiency of air/water heat pumps - Final report

    Energy Technology Data Exchange (ETDEWEB)

    Gasser, L.; Wellig, B.; Hilfiker, K.

    2008-04-15

    This comprehensive final report for the Swiss Federal Office of Energy (SFOE) presents the results of a study at the made by the Engineering and Architecture department at the Lucerne University of Applied Sciences and Arts. The subject of the WEXA study (Waermepumpen-Exergie-Analyse - heat pump exergy analysis) is the analysis of the operation of air/water heat-pumps using exergy analysis methods. The basic thermodynamics of heating systems using heat-pumps is discussed. The exergy analyses and exergy balances for the various components and processes of an air/water heat-pump are presented and discussed. Comparisons are presented for heat-pumps with on/off and continuous control systems for their compressors and fans. The paper is concluded with a collection of appendices on the subject.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

    1989-03-01

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

  12. Inadvertent raising of levels in the FFTF primary sodium pumps. Final unusual occurrence report, HEDL 79-34 (FFTF-58)

    International Nuclear Information System (INIS)

    Kuechle, J.D.

    1981-01-01

    The final unusual occurrence report describes the inadvertent raising of the sodium level in the FFTF primary sodium pumps during system testing. This event is now judged to have caused permanent deformation of the primary pump shaft on loop 1 during a period when pump rotation was stopped and sodium level in the pump tank was inadvertently increased. The shaft was subsequently removed, straightened, and returned to service in the spare FFTF pump

  13. Pumps modelling of a sodium fast reactor design and analysis of hydrodynamic behavior - 15294

    International Nuclear Information System (INIS)

    Ordonez, J.; Lazaro, A.; Martorell, S.

    2015-01-01

    One of the goals of Generation IV reactors is to increase safety from those of previous generations. Different research platforms have identified the need to improve the reliability of the simulation tools to ensure the capability of the plant to accommodate the design basis transients established in preliminary safety studies. The paper describes the modeling of recirculation pumps in advanced sodium cooled reactors using the TRACE code. Following the implementation of the models, the results obtained in the analysis of different design basis transients are compared with the simplifying approximations used in reference models. The paper shows the process to obtain a consistent pump model of the ESFR (European Sodium Fast Reactor) design and the analysis of loss of flow transients triggered by pumps coast-down analyzing the thermal hydraulic neutronic coupled system response. A sensitivity analysis of the system pressure drops effect and the other relevant parameters that influence the natural convection after the pumps coast-down is also included. (authors)

  14. Pumps modelling of a sodium fast reactor design and analysis of hydrodynamic behavior

    Directory of Open Access Journals (Sweden)

    Ordóñez Ródenas José

    2016-01-01

    Full Text Available One of the goals of Generation IV reactors is to increase safety from those of previous generations. Different research platforms have been identified the need to improve the reliability of the simulation tools to ensure the capability of the plant to accommodate the design basis transients established in preliminary safety studies. The paper describes the modelling of primary pumps in advanced sodium cooled reactors using the TRACE code. Following the implementation of the models, the results obtained in the analysis of different design basis transients are compared with the simplifying approximations used in reference models. The paper shows the process to obtain a consistent pump model of the ESFR (European Sodium Fast Reactor design and the analysis of loss of flow transients triggered by pumps coast–down analyzing the thermal hydraulic neutronic coupled system response. A sensitivity analysis of the system pressure drops effect and the other relevant parameters that influence the natural convection after the pumps coast–down is also included.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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. Reactor coolant pump motors manufacturing capability and references

    Energy Technology Data Exchange (ETDEWEB)

    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

  20. Reactor coolant pump motors manufacturing capability and references

    International Nuclear Information System (INIS)

    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

  1. Design of an isopropanol–acetone–hydrogen chemical heat pump with exothermic reactors in series

    International Nuclear Information System (INIS)

    Xu, Min; Duan, Yanjun; Xin, Fang; Huai, Xiulan; Li, Xunfeng

    2014-01-01

    The isopropanol–acetone–hydrogen chemical heat pump system with a series of exothermic reactors in which the reaction temperatures decrease successively is proposed. This system shows the better energy performances as compared with the traditional system with a single exothermic reactor, especially when the higher upgraded temperature is need. At the same amounts of the heat released, the work input of the compressor and the heater are both reduced notably. The results indicate that the advantages of the IAH-CHP system with exothermic reactors in series are obvious. - Highlights: • We propose the IAH-CHP system with exothermic reactors in series. • The COP and exergy efficiency of the system increase by 7.6% and 10.3% respectively. • The work input of the system is reduced notably at the same quantity of heat released

  2. Development of sputter ion pump based SG leak detection system for Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Babu, B.; Sureshkumar, K.V.; Srinivasan, G.

    2013-01-01

    Highlights: ► Development and commissioning of SG leak detection system for FBTR. ► Development of Robust method of using sputter ion pump based system. ► Modifications for improving reliability and availability. ► On line injection of hydrogen in sodium during reactor operation. ► Triplication of the SG leak detection system. - Abstract: The Fast Breeder Test Reactor (FBTR) is a 40 MWt, loop type sodium cooled fast reactor built at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam as a fore-runner to the second stage of Indian nuclear power programme. The reactor design is based on the French reactor Rapsodie with several modifications which include the provision of a steam-water circuit and turbo-generator. FBTR uses sodium as the coolant in the main heat transport medium to transfer heat from the reactor core to the feed water in the tertiary loop for producing superheated steam, which drives the turbo-generator. Sodium and water flow in shell and tube side respectively, separated by thin-walls of the ferritic steel tubes of the once-through steam generator (SG). Material defects in these tubes can lead to leakage of water into sodium, resulting in sodium water reactions leading to undesirable consequences. Early detection of water or steam leaks into sodium in the steam generator units of liquid metal fast breeder reactors (LMFBR) is an important requirement from safety and economic considerations. The SG leak in FBTR is detected by Sputter Ion Pump (SIP) based Steam Generator Leak Detection (SGLD) system and Thermal Conductivity Detector (TCD) based Hydrogen in Argon Detection (HAD) system. Many modifications were carried out in the SGLD system for the reactor operation to improve the reliability and availability. This paper details the development and the acquired experience of SIP based SGLD system instrumentation for real time hydrogen detection in sodium for FBTR.

  3. The Influence of RSG-GAS Primary Pump Operation Concerning the Rise Water Level of Reactor Pool in 15 MW Reactor Power

    International Nuclear Information System (INIS)

    Djunaidi

    2004-01-01

    The expansion of air volume in the delay chamber shows in rise water level of reactor pool during the operation. The rises of water level in the reactor pool is not quite from the expansion of air volume in the delay chamber, but some influence the primary pump operation. The purpose evaluated of influence primary pump is to know the influence primary pump power concerning the rise water level during the reactor operation. From the data collection during 15 MW power operation in the last core 42 the influence of primary pump operation concerning the rise water level in the reactor pool is 34.48 % from the total increased after operation during 12 days. (author)

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

    International Nuclear Information System (INIS)

    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)

  5. Nuclear reactor pump diagnostics via noise analysis/artificial neural networks

    International Nuclear Information System (INIS)

    Keyvan, S.; Rabelo, L.C.

    1991-01-01

    A feasibility study is performed on the utilization of artificial neural networks as a tool for reactor diagnostics. Reactor pump signals utilized in a wear-out monitoring system developed for early detection of degradation of pump shaft are analyzed as a semi-benchmark test to study the feasibility of neural networks for pattern recognition. The Adaptive Resonance Theory (ART 2) paradigm of artificial neural networks is applied in this study. The signals are collected signals as well as generated signals simulating the wear progress. The wear-out monitoring system applies noise analysis techniques, and is capable of distinguishing between these signals and providing a measure of the progress of the degradation. This paper presents the results of the analysis of these data via the ART 2 paradigm

  6. Development of electro-magnetic pump for the ASTRID Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Suzuki, Tetsu; Aizawa, Rie; Wakasaki, Shingo; Dechelette, Frank; Benoit, Fabrice

    2017-01-01

    In the framework of the SFR (Sodium-cooled Fast Reactor) prototype called ASTRID (Advance Sodium Technological Reactor for Industrial Demonstration), the large capacity Electro-Magnetic Pumps (EMP) as main circulating pumps on the intermediate sodium circuits has been considered instead of mechanical pumps by CEA. The use of EMP has several decisive technological merits compared with mechanical pump in the reactor design, operation and maintenance. Nevertheless, some theoretical and technological developments have to be carried out in order to validate the design tools which take Magneto Hydro Dynamic (MHD) phenomena into account and the applicability of the EMP to the steady state and transient operating conditions of ASTRID. To move forward to developments, a collaboration agreement between the CEA and TOSHIBA Corporation was made and entered into to carry out a joint work program on the EMP for ASTRID design and development. CEA performed the theoretical analysis, and the EMP experimental model is constructed by CEA to support these theoretical developments. This model consists of a middle-size annular EMP for the liquid metal sodium. The various testing program using this model has been started in 2016. TOSHIBA performed the examination of design specification for ASTRID, an electromagnetic design, a structural design and various analyses. The structure design has been examined the placement of the sodium boundary and the withstand pressure, etc. And, if the thicknesses of the structure increase for withstanding pressure, the pump efficiency falls because the loss of the electromagnetic force increases. Therefore the balance between withstanding pressure and the efficiency has been considered by an electromagnetism design. This paper presents the design studies and experimental activities for the EMP development in the framework of the CEA-TOSHIBA collaborations. (author)

  7. The results of studies of the thermohydraulics of the primary pumps in PWR reactors

    International Nuclear Information System (INIS)

    Grison, P.; Laura, J.F.

    1982-01-01

    In the context of its nuclear-safety programme for pressurized-water reactors, E.D.F. has engaged in theoretical and experimental studies in order to gain better knowledge of the behaviour of a pump under accident conditions, with passage to diphasic regime. The results of these studies are presented here, both from the experimental and theoretical points of view. They show in particular that the behaviour of the pump is essentially dictated by the interfacial friction for small flows, and by the appearance of a critical flow rate for large flows. The outline of the theoretical model describing the operation of the pump in the first three quadrants (positive and negative flow, positive and negative rotation) is described, as are certain special applications of the model, such as the determination of racing velocities in diphasic conditions [fr

  8. Results of studies of the thermohydraulics of the primary pumps in PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grison, P; Laura, J F [E.D.F., Chatou (France)

    1982-01-01

    In the context of its nuclear-safety program for pressurized-water reactors, E.D.F. has engaged in theoretical and experimental studies in order to gain better knowledge of the behaviour of a pump under accident conditions, with passage to diphasic regime. The results of these studies are presented here, both from the experimental and theoretical points of view. They show in particular that the behaviour of the pump is essentially dictated by the interfacial friction for small flows, and by the appearance of a critical flow rate for large flows. The outline of the theoretical model describing the operation of the pump in the first three quadrants (positive and negative flow, positive and negative rotation) is described, as are certain special applications of the model, such as the determination of racing velocities in diphasic conditions.

  9. The development of a cavitation free sodium pump for the breeder reactor

    International Nuclear Information System (INIS)

    Baladi, J.Y.; Nyilas, C.P.

    1986-01-01

    The sodium pumps for a liquid metal fast breeder reactor must be designed for exceptionally high reliability and long life. The principal adverse factor which tends to limit the primary pump life is cavitation which becomes potentially severe under off-design flow conditions caused by the requirement of two loop operations which resulted in a large operating flow range. This problem prompted an extensive study which included experimental investigations of scaled down and full size pumps. The investigations involved visual observations, acoustic signature recordings, and physical characteristic measurements of the model and full size impellers. The blade configuration of the model was modified several times. This paper describes the test facilities erected for this study, discusses the experimental techniques employed, and presents a sample of the experimental results

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Cardenas V, J.; Filio L, C.

    2016-09-01

    This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

  13. Development and computational simulation of thermoelectric electromagnetic pumps for controlling the fluid flow in liquid metal cooled space nuclear reactors

    International Nuclear Information System (INIS)

    Borges, E.M.

    1991-01-01

    Thermoelectric Electromagnetic (TEEM) Pumps can be used for controlling the fluid flow in the primary and secondary circuits of liquid metal cooled space nuclear reactor. In order to simulate and to evaluate the pumps performance, in steady-state, the computer program BEMTE has been developed to study the main operational parameters and to determine the system actuation point, for a given reactor operating power. The results for each stage of the program were satisfactory, compared to experimental data. The program shows to be adequate for the design and simulating of direct current electromagnetic pumps. (author)

  14. Use mobile pumps and liquid chilling water units to provide chilled water for nuclear reactor during nuclear power plant accident

    International Nuclear Information System (INIS)

    Zhang Guobin; Feng Jiaxuan

    2012-01-01

    From the nuclear accident in Japan Fuksuhima in March this year, despite a shut down of the reactor, the residue heat inside the reactor was not able to remove due to the failure of the cooling system and it finally caused the catastrophe. It was observed that when the failure of the cooling system after an earthquake of magnitude 9 and a tsunami of 28 meters height, the containment vessel for the reactor core was still able to maintain its integrity in the first 24 hours before the first explosion was happened. A backup emergency heat removal system for nuclear power plants using mo- bile pumps and liquid chilling units has been proposed 20 years ago by Cheung [Ref. 1]. Due to the fact that there are more than 400 nuclear power plants around the world and 10% of them are located in earthquake active zone, together with the aging of some of the power plants which were built more than 30 years ago, the risk of another nuclear accident becomes high. An emergency safety measure has to be designed in order to deal with the unforeseen scenario. This re- port re-visits the proposal again; to re-design to the suit the need and to integrate with the current situation of the nuclear industry. (authors)

  15. PUMPS

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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

  17. Design features and operation experience of the main circulating pumps for the ''Loviisa'' NPP with the WWER-440 reactor

    International Nuclear Information System (INIS)

    Iofs, D.; Kujyala, I.; Timperi, I.; Shlejfer, G.; Vistbakka, V.; Prudovskij, A.M.; Turetskij, L.I.; Vorona, P.N.

    1980-01-01

    Technical characteristics and the operation of main circulating pumps (MCP) designed and mounted at the ''Loviisa'' NPP by Finnish firms ''Alstrem'' and ''Stremberg'' are described. The above MCP have specific advantages over similar pumps mounted at other NPP with pressurized water cooled reactors. This is a possibility of substitution of potentially most damaged units (bearing and pump shaft sealing) for several hours, without MCP disassembly as a whole as well as using rolling bearings together with the original electromagnetic unloading system from the axial force instead of usually employed in similar MCP radial thrust slip bearings. The two year operation experience has confirmed the efficiency and reliability of ''Loviisa'' NPP main circulating pumps

  18. Calcium oxide/carbon dioxide reactivity in a packed bed reactor of a chemical heat pump for high-temperature gas reactors

    International Nuclear Information System (INIS)

    Kato, Yukitaka; Yamada, Mitsuteru; Kanie, Toshihiro; Yoshizawa, Yoshio

    2001-01-01

    The thermal performance of a chemical heat pump that uses a calcium oxide/carbon dioxide reaction system was discussed as a heat storage system for utilizing heat output from high temperature gas reactors (HTGR). Calcium oxide/carbon dioxide reactivity for the heat pump was measured using a packed bed reactor containing 1.0 kg of reactant. The reactor was capable of storing heat at 900 deg. C by decarbonation of calcium carbonate and generating up to 997 deg. C by carbonation of calcium oxide. The amount of stored heat in the reactor was 800-900 kJ kg -1 . The output temperature of the reactor could be controlled by regulating the carbonation pressure. The thermal storage performance of the reactor was superior to that of conventional sensible heat storage systems. A heat pump using this CaO/CO 2 reactor is expected to contribute to thermal load leveling and to realize highly efficient utilization of HTGR output due to the high heat storage density and high-quality temperature output of the heat pump

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

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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. RA Reactor operation and maintenance (I-IX), part V, Task 3.08/04-06, Refurbishment of the heavy water pumps

    International Nuclear Information System (INIS)

    Zecevic, V.; Nikolic, M.; Milic, J.

    1963-12-01

    In addition to detailed instructions for maintenance and repair of the heavy water pumps at the RA reactor this document includes nine annexes. They are as follows: cleaning the heavy water pump Avala with distilled water; instructions for repair of the pump CEN-132 (two annexes); list of operating characteristics of the pumps before repair; conclusions of the experts concerning the worn out bearings of the heavy water pump Avala, with the analysis of the stellite layer; report on the completed repair actions on the pumps Avala and CEN-132; report on the measurements done on the pump Avala; and the certificate concerning inspection of the pump

  3. Final Physics Report for the Engineering Test Reactor

    International Nuclear Information System (INIS)

    Wolfe, I. B.

    1956-01-01

    order to reduce as much as possible the cost of the physics analysis and in order to obtain information otherwise unavailable, considerable use was made of perturbation techniques. These perturbation computations were initially performed using the one-dimensional approximation and were extended to two dimensions in stages 3 and 4. To do this a method of obtaining adjoint fluxes; using available reactor computer codes, was developed. Only the physics which bears on the final design of the ETR is summarized. This volume, together with the Physics Progress Report, represents a complete account of the studies undertaken,. methods used, and results obtained in the physics work on the ETR

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    Peyrouty, P.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Peyrouty, P.

    1996-12-01

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

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

    International Nuclear Information System (INIS)

    Peyrouty, P.

    1996-01-01

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

  8. Tritium system for a tokamak reactor with a self-pumped limiter

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Sze, D.K.

    1986-01-01

    The self-pumping concept was proposed as a means of simplifying the impurity control system in a fusion reactor. The idea is to remove helium in-situ by trapping in freshly deposited metal surface layers of a limiter or divertor. Trapping material is added to the plasma scrape-off or edge region where it is transported to the wall. Some of the key issues for this concept are the tritium inventory in the trapping material and the permeation of protium and recycling of tritium. These quantities are shown to be acceptable for the reference design. The tritium issues for a helium-cooled solid breeder reactor design with vanadium alloy as a structural material are also examined. Models are presented for tritium permeation and inventory calculation for structure materials with the effect of a thin layer of coating material

  9. Centrifugal pump assembly for use in nuclear reactor plants or the like

    International Nuclear Information System (INIS)

    Honold, E.; Ruepp, M.

    1975-01-01

    A description is given of a vertical centrifugal pump assembly for use as a primary recirculating pump in a nuclear reactor plant which has an uppermost shaft which is driven by a motor, a lowermost shaft which drives the impeller of the pump, and an intermediate shaft which is movable axially between the uppermost and lowermost shafts and carries a housing for radial and/or thrust bearings. A rigid coupling between the lowermost shaft and the intermediate shaft is disengaged when the intermediate shaft is lifted to thereby afford access to certain parts of the lowermost shaft. The uppermost shaft can drive the intermediate shaft through the medium of a flexible coupling having coupling elements mounted on the lower end of the uppermost shaft and the upper end of the intermediate shaft, and a distancing sleeve whose internal threads are in permanent mesh with external threads of the two coupling elements. A shroud which surrounds the flexible coupling has a maximum-diameter tubular section rigid with the housing for the bearings, a minimum-diameter tubular section rigid with but detachable from the casing of the motor, and a median tubular section which penetrates into the maximum-diameter section when the intermediate shaft is lifted. When the minimum-diameter section is detached from the casing of the motor, it can be slid, together with the intermediate section, into the maximum-diameter section to thereby afford access to the flexible coupling. The rigid coupling has two coaxial coupling elements which are mounted at the upper end of the pump shaft and the lower end of the intermediate shaft and may have teeth which mate when the intermediate shaft assumes its lower end position in which it can drive the pump

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

    International Nuclear Information System (INIS)

    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)

  11. IRIS International Reactor Innovative and Secure Final Technical Progress Report

    International Nuclear Information System (INIS)

    Carelli, M.D.

    2003-01-01

    OAK-B135 This NERI project, originally started as the Secure Transportable Autonomous Light Water Reactor (STAR-LW) and currently known as the International Reactor Innovative and Secure (IRIS) project, had the objective of investigating a novel type of water-cooled reactor to satisfy the Generation IV goals: fuel cycle sustainability, enhanced reliability and safety, and improved economics. The research objectives over the three-year (1999-2002) program were as follows: First year: Assess various design alternatives and establish main characteristics of a point design; Second year: Perform feasibility and engineering assessment of the selected design solutions; Third year: Complete reactor design and performance evaluation, including cost assessment These objectives were fully attained and actually they served to launch IRIS as a full fledged project for eventual commercial deployment. The program did not terminate in 2002 at the end of the NERI program, and has just entered in its fifth year. This has been made possible by the IRIS project participants which have grown from the original four member, two-countries team to the current twenty members, nine countries consortium. All the consortium members work under their own funding and it is estimated that the value of their in-kind contributions over the life of the project has been of the order of $30M. Currently, approximately 100 people worldwide are involved in the project. A very important constituency of the IRIS project is the academia: 7 universities from four countries are members of the consortium and five more US universities are associated via parallel NERI programs. To date, 97 students have worked or are working on IRIS; 59 IRIS-related graduate theses have been prepared or are in preparation, and 41 of these students have already graduated with M.S. (33) or Ph.D. (8) degrees. This ''final'' report (final only as far as the NERI program is concerned) summarizes the work performed in the first four

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  15. Basic concept of common reactor physics code systems. Final report of working party on common reactor physics code systems (CCS)

    International Nuclear Information System (INIS)

    2004-03-01

    A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)

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

    International Nuclear Information System (INIS)

    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)

  17. Research Project 'RB research nuclear reactor' (operation and maintenance), Final report

    International Nuclear Information System (INIS)

    1985-01-01

    This final report covers operation and maintenance activities at the RB reactor during period from 1981-1985. First part covers the RB reactor operation, detailed description of reactor components, fuel, heavy water, reactor vessel, cooling system, equipment and instrumentation, auxiliary systems. It contains data concerned with dosimetry and radiation protection, reactor staff, and financial data. Second part deals maintenance, regular control and testing of reactor equipment and instrumentation. Third part is devoted to basic experimental options and utilization of the RB reactor including training

  18. Development and validation of three-dimensional CFD techniques for reactor safety applications. Final report

    International Nuclear Information System (INIS)

    Buchholz, Sebastian; Palazzo, Simone; Papukchiev, Angel; Scheurer Martina

    2016-12-01

    The overall goal of the project RS 1506 ''Development and Validation of Three Dimensional CFD Methods for Reactor Safety Applications'' is the validation of Computational Fluid Dynamics (CFD) software for the simulation of three -dimensional thermo-hydraulic heat and fluid flow phenomena in nuclear reactors. For this purpose a wide spectrum of validation and test cases was selected covering fluid flow and heat transfer phenomena in the downcomer and in the core of pressurized water reactors. In addition, the coupling of the system code ATHLET with the CFD code ANSYS CFX was further developed and validated. The first choice were UPTF experiments where turbulent single- and two-phase flows were investigated in a 1:1 scaled model of a German KONVOI reactor. The scope of the CFD calculations covers thermal mixing and stratification including condensation in single- and two-phase flows. In the complex core region, the flow in a fuel assembly with spacer grid was simulated as defined in the OECD/NEA Benchmark MATIS-H. Good agreement are achieved when the geometrical and physical boundary conditions were reproduced as realistic as possible. This includes, in particular, the consideration of heat transfer to walls. The influence of wall modelling on CFD results was investigated on the TALL-3D T01 experiment. In this case, the dynamic three dimensional fluid flow and heat transfer phenomena were simulated in a Generation IV liquid metal cooled reactor. Concurrently to the validation work, the coupling of the system code ATHLET with the ANSYS CFX software was optimized and expanded for two-phase flows. Different coupling approaches were investigated, in order to overcome the large difference between CPU-time requirements of system and CFD codes. Finally, the coupled simulation system was validated by applying it to the simulation of the PSI double T-junction experiment, the LBE-flow in the MYRRA Spallation experiment and a demonstration test case simulating a pump trip

  19. Research and development on super heat pump energy accumulation system. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-06-01

    This is the final report on research and development of super heat pump energy accumulation system, which has been carried out from FY 1985 to 1992. It describes outline of the research and development program, R and D results, final evaluation methodology, evaluation of the R and D, proposals for the commercialization, and so on. The super high performance compression heat pumps are technically evaluated for highly efficient type (for heating, and cooling and heating), high temperature type (utilizing high temperature heat source, and low temperature heat source), working fluids (alcohol-based and nonalcohol-based), stainless steel plate fin type heat exchanger, EHD heat exchanger, and so on. The other techniques evaluated include those for chemical heat storage, combined systems, plant simulation, and systemization. The evaluation works are also directed to the economic and environmental aspects. Finally, the R and D themes are proposed to leap over various hurdles, e.g., reliability and economic viability, for the eventual commercialization of the energy accumulation system. (NEDO)

  20. Characterization of nuclear reactor containment penetrations. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Shackelford, M.H.; Bump, T.R.; Seidensticker, R.W.

    1985-02-01

    This report concludes a preliminary report prepared by ANL for Sandia, published as NUREG/CR-3855, in June 1984. The preliminary report, NUREG/CR-3855, presented the results of a survey of nuclear reactor containment penetrations, covering the number of plants surveyed at that time (22 total). Since that time, an additional 26 plants have been included in the survey. This final report serves two purposes: (1) to add the summary data sheets and penetration details for the additional plants now included in the survey; and (2) to confirm, revise, or add to analyses and discussions presented in the first report which, of course, were based solely on the earlier sample of 22 plants. This final report follows the outline and format of the preliminary survey report. In general, changes and additions to the preliminary report are implied, rather than stated as such to avoid repeated reference to that report. If no changes have been made in a section the title of the section of the previous report is simply repeated followed by ''No Changes''. Some repetition is used for continuity and clarity.

  1. Assessment of feasibility of helium ash exhaust and heat removal by pumped-limiter in tokamak fusion reactor

    International Nuclear Information System (INIS)

    Hitoki, Shigehisa; Sugihara, Masayoshi; Saito, Seiji; Fujisawa, Noboru

    1985-01-01

    A detailed calculation of the behavior of fuel and He particles in tokamak reactor with pumped-limiter is performed by one-dimensional tokamak transport code. Energy of neutral particles flowing back from limiter chamber is calculated by two-dimensional Monte Carlo neutral code. Feasibility of He ash exhaust and heat removal by the pumped-limiter are analyzed. Following features of the pumped-limiter are clarified: (1) Electron temperature decays rapidly in radial direction in scrape-off layer, while density profile is broader than that of temperature. (2) Helium accumulation in main plasma can be kept at desired level by rather short limiter and moderate pumping system. (3) Minimum amount of tritium pumped out little depends on limiter length. (4) Although high temperature plasma in scrape-off layer could be realized by large pumping and ideal pellet injection, it is not sufficiently high to reduce the erosion of the limiter surface and the leading edge. In conclusion, He ash exhaust may be possible by the pumped-limiter, while the heat load and erosion will be so high that the pumped-limiter may not be applicable unless the boundary plasma is cooled by radiation or by some other means. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  4. Study of the modifications on the synchronous generators, heavy water pumps and condenser batteries of the RA reactor - Annex 17

    International Nuclear Information System (INIS)

    Milosevic, M.

    1964-01-01

    Modifications done on the synchronous generators are related to the emergency power supply system, meaning one of the most important devices responsible for reactor safety. Without reducing the efficiency of the heavy water pumps the improved stability of generators operation was achieved by reducing the possibility of errors and simplifying manipulation. Condensator batteries were improved in order to decrease the leakage currents

  5. Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report

    International Nuclear Information System (INIS)

    Rohde, Ulrich; Pivovarov, Valeri; Matveev, Yurij

    2010-12-01

    The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

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

    International Nuclear Information System (INIS)

    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

  7. A Takagi-Sugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Yue [Institute of Nuclear and New Energy Technology, Tsinghua University, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing (China); Coble, Jamie [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

    2017-08-15

    Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional–integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR) design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi–Sugeno (T–S) fuzzy logic-based power distribution system. Two T–S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T–S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

  8. A Takagi–Sugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation

    Directory of Open Access Journals (Sweden)

    Yue Yuan

    2017-08-01

    Full Text Available Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional–integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi–Sugeno (T–S fuzzy logic-based power distribution system. Two T–S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T–S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

  9. Apparatus for servicing a jet pump hold down beam in a nuclear reactor

    International Nuclear Information System (INIS)

    Howell, D.A.; Hydeman, J.E.; Slater, J.L.; Bodnar, R.J.; Golick, L.R.; Sckera, R.S.; Roth, C.H. Jr.

    1991-01-01

    This patent describes an apparatus for replacing the hold down beam of a fluid circulating jet pump mounted in a nuclear reactor, the hold down beam having a beam body, a pair of opposed beam tabs and a pair of opposed beam positioning trunnions extending outwardly from the beam body. It comprises a housing having a lower surface configured to be positionable over the body of the hold down beam; means coupled to the housing for engaging the beam trunnions and securing the beam body against the lower surface of the housing; means coupled to the housing for depressing the beam tabs while the beam body is secured against the lower surface of the housing; means coupled to the trunnion engaging means and the beam tab depressing means for selectively actuating the trunnion engaging means and the beam tab depressing means from a position remote from the nuclear reactor; and means connectable to the housing for selectively changing the directional orientation of the beam

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Calculation method for the seasonal performance of heat pump compact units and validation. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Wemhoener, C.; Dott, R.; Afjei, Th. [University of Applied Sciences Northwestern Switzerland, Institute of Energy in Buildings, Muttenz (Switzerland); Huber, H.; Helfenfinger, D.; Keller, P.; Furter, R. [University of Applied Sciences Lucerne (HTA), Test center HLKS, Horw (Switzerland)

    2007-02-15

    This comprehensive final report for the Swiss Federal Office of Energy (SFOE) takes a look at compact heat pump units that have been developed for the heating of low energy consumption houses built to MINERGIE or MINERGIE-P standards. These units, which combine the functions of space heating, domestic hot water preparation and ventilation in one unit are described. A testing procedure developed at the University of Applied Science in Lucerne, Switzerland, using a test rig for the measurement of the seasonal performance factor (SPF) is described. A calculation method based on temperature classes for the calculation of the SPF of combined heat pump systems for space heating and domestic hot water preparation that was developed by the Institute of Energy in Buildings at the University of Applied Sciences Northwestern Switzerland is examined. Two pilot plants allowing detailed field monitoring of two compact units are described. One pilot plant installed in a single-family house built to MINERGIE standard in Gelterkinden, Switzerland, provided data on a compact unit. These results of measurements made on this and a further installation in a MINERGIE-P ultra-low energy consumption house in Zeiningen, Switzerland, are presented and discussed. Calculation methods, including exergy considerations are reviewed and their validation is discussed.

  12. Research on enhancement of natural circulation capability in lead–bismuth alloy cooled reactor by using gas-lift pump

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Juanli, E-mail: Jenyzuo@163.com; Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn; Chen, Ronghua, E-mail: ronghua.chen@stu.xjtu.edu.cn; Qiu, Suizheng; Su, Guanghui, E-mail: ghsu@mail.xjtu.edu.cn

    2013-10-15

    Highlights: • The gas-lift pump has been adopted to enhance the natural circulation capability. • LENAC code is developed in my study. • The calculation results by LENAC code show good agreement with experiment results. • Gas mass flow rate, bubble diameter, rising pipe length are important parameters. -- Abstract: The gas-lift pump has been adopted to enhance the natural circulation capability in the type of lead–bismuth alloy cooled reactors such as Accelerator Driven System (ADS) and Liquid–metal Fast Reactor (LMFR). The natural circulation ability and the system safety are obviously influenced by the two phase flow characteristics of liquid metal–inert gas. In this study, LENAC (LEad bismuth alloy NAtural Circulation capability) code has been developed to evaluate the natural circulation capability of lead–bismuth cooled ADS with gas-lift pump. The drift flow theory, void fraction prediction model and friction pressure drop prediction model have been incorporated into LENAC code. The calculation results by LENAC code show good agreement with experiment results of CIRCulation Experiment (CIRCE) facility. The effects of the gas mass flow rate, void fraction, gas quality, bubble diameter and the rising pipe height or the potential difference between heat exchanger and reactor core on natural circulation capability of gas-lift pump have been analyzed. The results showed that in bubbly flow pattern, for a fixed value of gas mass flow rate, the natural circulation capability increased with the decrease of the bubble diameter. In the bubbly flow, slug flow, churn flow and annular flow pattern, with the gas mass flow rate increasing, the natural circulation capability initially increased and then declined. And the flow parameters influenced the thermal hydraulic characteristics of the reactor core significantly. The present work is helpful for revealing the law of enhancing the natural circulation capability by gas-lift pump, and providing theoretical

  13. Operating experience feedback report: Reliability of safety-related steam turbine-driven standby pumps. Commercial power reactors, Volume 10

    International Nuclear Information System (INIS)

    Boardman, J.R.

    1994-10-01

    This report documents a detailed analysis of failure initiators, causes and design features for steam turbine assemblies (turbines with their related components, such as governors and valves) which are used as drivers for standby pumps in the auxiliary feedwater systems of US commercial pressurized water reactor plants, and in the high pressure coolant injection and reactor core isolation cooling systems of US commercial boiling water reactor plants. These standby pumps provide a redundant source of water to remove reactor core heat as specified in individual plant safety analysis reports. The period of review for this report was from January 1974 through December 1990 for licensee event reports (LERS) and January 1985 through December 1990 for Nuclear Plant Reliability Data System (NPRDS) failure data. This study confirmed the continuing validity of conclusions of earlier studies by the US Nuclear Regulatory Commission and by the US nuclear industry that the most significant factors in failures of turbine-driven standby pumps have been the failures of the turbine-drivers and their controls. Inadequate maintenance and the use of inappropriate vendor technical information were identified as significant factors which caused recurring failures

  14. Probabilistic study of primary pump trip in a P.W.R. reactor: use of response surface methodology

    International Nuclear Information System (INIS)

    Bars, C.; Duchemin, B.; Maigret, N.; Peltier, J.; Rostan, O.; Villeneuve, M.J. de; Lanore, J.M.

    1981-09-01

    This paper describes a probabilistic study about the consequences of the trip or blockage of one of the three PWR reactor primary pumps. The distribution of the input parameters is taken into account and the resulting distribution of the consequence (number of failed fuel rods) is assessed. The necessity to do this study with the response surface methodology and the precautions to take are outlined. The results show that the probability to have failed fuel rods is about 10 -4 for pump trip and 0.16 for blockage with, in this case, a mean of 196 failed rods, that is 0.5 % of total number of rods

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  16. Final Report: Reactor Sharing, September 30, 1996 - September 29, 1998

    International Nuclear Information System (INIS)

    Williams, John G.

    1999-01-01

    Under the support provided by the DOE Reactor Sharing Program the Reactor Laboratory has provided tours, lectures, and demonstrations in support of science teaching in Arizona high schools. The reactor has also been used in a very successful summer program to encourage high school students who are members of population groups underrepresented in engineering to consider careers in engineering fields. This program is in the form of one or two week on-campus workshops given several times each summer to students at different levels of junior or senior high school. The Reactor Laboratory was one of six or eight areas of engineering to which the participants were introduced. The degree of involvement ranged from tours and demonstrations of reactor operation in small groups for the younger students, to neutron activation analysis experiments, with student participation, at the higher grade levels. The reactor time funded by this DOE grant has provided significant service to students and faculty from other educational institutes using our facilities. In addition, we have had the opportunity to provide public education in nuclear reactor science and engineering to a wide variety of groups, especially school children

  17. RETRAN sensitivity studies of light water reactor transients. Final report

    International Nuclear Information System (INIS)

    Burrell, N.S.; Gose, G.C.; Harrison, J.F.; Sawtelle, G.R.

    1977-06-01

    This report presents the results of sensitivity studies performed using the RETRAN/RELAP4 transient analysis code to identify critical parameters and models which influence light water reactor transient predictions. Various plant transients for both boiling water reactors and pressurized water reactors are examined. These studies represent the first detailed evaluation of the RETRAN/RELAP4 transient code capability in predicting a variety of plant transient responses. The wide range of transients analyzed in conjunction with the parameter and modeling studies performed identify several sensitive areas as well as areas requiring future study and model development

  18. Study of an electromagnetic pump applied to a primary main pump of a large scale sodium cooled reactor

    International Nuclear Information System (INIS)

    Aizawa, Kosuke; Kotake, Shoji; Chikazawa, Yoshitaka; Ara, Kuniaki; Araseki, Hideo; Aizawa, Rie; Ota, Hiroyuki

    2009-01-01

    This paper describes a future innovative design options with a parallel electromagnetic pump (EMP) system as the main circulating pump of the JSFR design. A conceptual design of EMPs integrated with an intermediate heat exchanger (IHX) is carried out. The major design parameters are consistent with the current JSFR design, where the main flow rate is 630 m 3 /min and the flow halving time is the same of the mechanical pump with the similar reliability. As a result of several design studies, a five parallel EMPs with IHX system has been selected from the geometry suitability for JSFR design. The EMP advantages comparing with mechanical pumps are investigated from the views of in-service inspection, maintenance and reliability. Numerical analysis with two dimensional MHD codes is conducted on a former experiment of a 160 m 3 /min flow rate EMP. The overall trend of the experimental data and the numerical results agrees with that in small-scale EMPs. However, the difference between the experimental data and the numerical results seems larger compared with the small-scale EMPs, which comes from large magnetic Reynolds number and interaction parameter of 160 m 3 /min EMP. (author)

  19. 105-H Reactor Interim Safe Storage Project Final Report

    International Nuclear Information System (INIS)

    Ison, E.G.

    2008-01-01

    The following information documents the decontamination and decommissioning of the 105-H Reactor facility, and placement of the reactor core into interim safe storage. The D and D of the facility included characterization, engineering, removal of hazardous and radiologically contaminated materials, equipment removal, decontamination, demolition of the structure, and restoration of the site. The ISS work also included construction of the safe storage enclosure, which required the installation of a new roofing system, power and lighting, a remote monitoring system, and ventilation components.

  20. Growth rates of breeder reactor fuel. Final report

    International Nuclear Information System (INIS)

    Ott, K.O.

    1979-01-01

    During the contract period, a consistent formalism for the definition of the growth rates (and thus the doubling time) of breeder reactor fuel has been developed. This formalism was then extended to symbiotic operation of breeder and converter reactors. Further, an estimation prescription for the growth rate has been developed which is based upon the breeding worth factors. The characteristics of this definition have been investigated, which led to an additional integral concept, the breeding bonus

  1. Principle tests of reactor pumped laser of 3He-Ar-Xe system

    International Nuclear Information System (INIS)

    Chen Hande; Jin Xingxing; Yang Chengde

    1994-01-01

    A nuclear reactor-pumped laser was demonstrated firstly by using the 3 He(P,n) 3 H reaction to excite a 3 He-Ar-Xe laser. Lasing was observed on the 1.73 μm (5d (3/2) 1 -6p(5/2) 1 transition) in Xe I. The CAEP Pulsed Reactor (CFBR-II) was used as a source of fast neutrons moderated by a 50 mm thick polyethylene sleeve around the laser cell. A laser cell was constructed from K 9 glass tubing 600 mm long and Φ34 mm with each end cut at Brewster's angle (quartz). The laser cell was baked out to 10 -3 Pa prior to filling with research-grade mixture ( 3 He/Ar/Xe = 34.7: 34.7: 0.267 kPa). A dielectric-coated 2 m radius-curvature back mirror (99.7% reflectivity at 1.73 μm) and a flat output coupler (Φ30 mm) were used to form the optical cavity. The cavity optics were placed outside the cell as shown in Fig.3. A schematic of the experimental configuration is shown. Typical laser signals obtained with the Ge detector (-80 degree C) are also shown. Laser output lags the fast neutron pulse by the neutron thermalization time (∼278 μs). The observed light output was shown to be caused by stimulated emission, because the signal disappeared when the reflecting (flat) mirror was covered. This precludes the possibility that the phenomenon was the result of γ-radiation interaction with the detector or the air in the light path. The measured laser energy was 80 times the maximum possible fluorescence energy. The observed wavelength was measured to be 1.73 μm by using glass filters. Laser output duration was approximately 735 μs

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

    International Nuclear Information System (INIS)

    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

  3. Monitoring device for shaft oscillation of reactor incorporated-type recycling pump

    International Nuclear Information System (INIS)

    Miyashita, Kaoru; Shibasaki, Kimiyuki.

    1995-01-01

    The present invention concerns monitoring of recycling pump shaft oscillation in a BWR type reactor, which monitors by separating a rotation pulse signal and a shaft oscillation waveform signal obtained in a non-contact type displacement meter. Namely, a threshold value calculation means of a separation processing section takes in original waveform data and selects the maximum value and the minimum value among them. A threshold value is calculated based on the values. An average value of the original waveform data for portions which do not exceed the threshold value is calculated. A first calculation means compares each of the original data with the threshold value, and if the original data are greater than the threshold value, they are outputted as the original data corresponding to the rotation pulse signal. When the original data are smaller than the threshold value, they are outputted as they are as a shaft oscillation waveform signal. On the other hand, a second calculation means calculates an average value for the pulse of the original waveform data corresponding to the rotation pulse signal. An average value of the original waveform data which do not exceed the threshold value are subtracted from the average value, to form the shaft oscillation waveform signal and output the same. (I.S.)

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

    International Nuclear Information System (INIS)

    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

  5. Development of ABWR inertia-increased reactor internal pump and thicker sleeve nozzle

    International Nuclear Information System (INIS)

    Takahashi, Shirou; Shiina, Kouji; Matsumura, Seiichi

    2002-01-01

    The conventional reactor internal pumps (RIPs) in the ABWR have an inertia moment coming from the shafts and Motor-Generator sets, enabling the RIPs to continue running for a few seconds, when a trip of all RIPs event occurs. It is possible to simplify the RIPs' power supply system without affecting the core flow supply when the above event occurs by eliminating M-G sets, if the rotating inertia is increased. This inertia increase due to an additional flywheel, which leads to gains in weight and length, requires the larger diameter nozzle with the thicker sleeve. However, too large a nozzle diameter may change the hydraulic performance. In authors' previous study, the optimum nozzle diameter (492 mm) was selected through 1/5-scale test. In this study, the 492 mm nozzle and the inertia-increased RIP were verified through the full-scale tests. The rotating inertia time constant on coastdown characteristics (behavior of the RIP speed in the event of power loss) for the inertia-increased RIP doubled compared with the current RIP. The casing and the shaft vibration were also confirmed to satisfy the design criteria. Moreover, hydraulic performance and heat increase in the motor casing due to the flywheel were evaluated. The inertia increased RIP with the 492 mm nozzle maintained good performance. (author)

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

    International Nuclear Information System (INIS)

    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

  7. Mirror Advanced Reactor Study (MARS) final report summary

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Carlson, G.A.

    1983-01-01

    The Mirror Advanced Reactor Study (MARS) has resulted in an overview of a first-generation tandem mirror reactor. The central cell fusion plasma is self-sustained by alpha heating (ignition), while electron-cyclotron resonance heating and negative ion beams maintain the electrostatic confining potentials in the end plugs. Plug injection power is reduced by the use of high-field choke coils and thermal barriers, concepts to be tested in the Tandem Mirror Experiment-Upgrade (TMX-U) and Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory

  8. Stainless steel clad for light water reactor fuels. Final report

    International Nuclear Information System (INIS)

    Rivera, J.E.; Meyer, J.E.

    1980-07-01

    Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

  9. Decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East. Project final report

    International Nuclear Information System (INIS)

    Fellhauer, C.R.; Clark, F.R.

    1997-10-01

    The decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East (ANL-E) was completed in October 1997. Descriptions and evaluations of the activities performed and analyses of the results obtained during the JANUS D and D Project are provided in this Final Report. The following information is included: objective of the JANUS D and D Project; history of the JANUS Reactor facility; description of the ANL-E site and the JANUS Reactor facility; overview of the D and D activities performed; description of the project planning and engineering; description of the D and D operations; summary of the final status of the JANUS Reactor facility based upon the final survey results; description of the health and safety aspects of the project, including personnel exposure and OSHA reporting; summary of the waste minimization techniques utilized and total waste generated by the project; and summary of the final cost and schedule for the JANUS D and D Project

  10. Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Zhujie [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Bartels, David [Univ. of Notre Dame, IN (United States)

    2015-04-02

    This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that the effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.

  11. Actinide behavior in the Integral Fast Reactor. Final project report

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  12. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  13. Tokamak Fusion Test Reactor. Final conceptual design report

    International Nuclear Information System (INIS)

    1976-02-01

    The TFTR is the first U.S. magnetic confinement device planned to demonstrate the fusion of D-T at reactor power levels. This report addresses the physics objectives and the engineering goals of the TFTR project. Technical, cost, and schedule aspects of the project are included

  14. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

    1994-01-01

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro's Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel

  15. Actinide behavior in the Integral Fast Reactor. Final project report

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  17. Quarterly review of 241-SY-101 mixer pump data: January - March 1999; FINAL

    International Nuclear Information System (INIS)

    CONNER, J.M.

    1999-01-01

    This report presents data obtained on 241-SY-101 pump performance. The period covered is January 1 through March 31, 1999. During the quarter: There were changes in pumping parameters. Both the pump volute pressure and amperage decreased during the quarter. It is not clear whether this was due to changes in waste properties (due to less solids or more entrained gas) or due to degradation of the pump. There was an indication of a 7.5-inch increase in the waste level at riser 1 A, and an average growth rate of 0.082 inches per day. There was an indication of a 5.7-inch increase in the waste level at riser 1C. This riser was flushed with water several times, which would lower the level of the crust at this location. Gases continued to be released at less than the pre-pump installation baseline rate, indicating a decrease in the gas generation rate, or an increase in gas retention, or both. The release rate was about 78 percent of the rate in the previous few quarters, and only 34 percent of the generation rate calculated prior to mixer pump installation in 1993. Key controls exist for waste temperature, gas concentration, pump parameters, and long-term waste behavior associated with the safe operation of the mixer pump that mitigates the buoyant displacement gas release event behavior of 241-SY-101. Table 1-1 compares the key controls and the current state of the waste as of March 3 1. 1999. The pump was run 28 times between January 1 and March 31, 1999. All of the pump runs were intended to be normal 25-minute, 1000-rpm excavation runs performed to mix the waste and release gas. Because of the pump oil often reached the high temperature alarm setpoint of 190 F, many of the runs were shortened (by as many as 8 minutes). This phenomenon was identified in November 1998, but got progressively worse over the quarter. The pump schedule was nominally three runs per week. However, core sampling activities interrupted the usual pump schedule several times during the quarter

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

    International Nuclear Information System (INIS)

    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

  19. Pumps and pump facilities. 2. ed.

    International Nuclear Information System (INIS)

    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

  20. Sources of radioiodine at pressurized water reactors. Final report

    International Nuclear Information System (INIS)

    Pelletier, C.A.; Cline, J.E.; Barefoot, E.D.; Hemphill, R.T.; Voilleque, P.G.; Emel, W.A.

    1978-11-01

    The report determines specific components and operations at operating pressurized water reactors that have a potential for being significant emission sources of radioactive iodine. The relative magnitudes of these specific sources in terms of the chemical forms of the radioiodine and the resultant annual averages from major components are established. The data are generalized for broad industry use for predictive purposes. The conclusions of this study indicate that the majority of radioiodine emanating from the primary side of pressurized water reactors comes from a few major areas; in some cases these sources are locally treatable; the interaction of radioiodine with plant interior surfaces is an important phenomenon mediating the source and affecting its release to the atmosphere; the chemical form varies depending on the circumstances of the release

  1. Repair welding of fusion reactor components. Final technical report

    International Nuclear Information System (INIS)

    Chin, B.A.; Wang, C.A.

    1997-01-01

    The exposure of metallic materials, such as structural components of the first wall and blanket of a fusion reactor, to neutron irradiation will induce changes in both the material composition and microstructure. Along with these changes can come a corresponding deterioration in mechanical properties resulting in premature failure. It is, therefore, essential to expect that the repair and replacement of the degraded components will be necessary. Such repairs may require the joining of irradiated materials through the use of fusion welding processes. The present ITER (International Thermonuclear Experimental Reactor) conceptual design is anticipated to have about 5 km of longitudinal welds and ten thousand pipe butt welds in the blanket structure. A recent study by Buende et al. predict that a failure is most likely to occur in a weld. The study is based on data from other large structures, particularly nuclear reactors. The data used also appear to be consistent with the operating experience of the Fast Flux Test Facility (FFTF). This reactor has a fuel pin area comparable with the area of the ITER first wall and has experienced one unanticipated fuel pin failure after two years of operation. The repair of irradiated structures using fusion welding will be difficult due to the entrapped helium. Due to its extremely low solubility in metals, helium will diffuse and agglomerate to form helium bubbles after being trapped at point defects, dislocations, and grain boundaries. Welding of neutron-irradiated type 304 stainless steels has been reported with varying degree of heat-affected zone cracking (HAZ). The objectives of this study were to determine the threshold helium concentrations required to cause HAZ cracking and to investigate techniques that might be used to eliminate the HAZ cracking in welding of helium-containing materials

  2. Distributions and activities of ammonia oxidizing bacteria and polyphosphate accumulating organisms in a pumped-flow biofilm reactor.

    Science.gov (United States)

    Wu, Guangxue; Nielsen, Michael; Sorensen, Ketil; Zhan, Xinmin; Rodgers, Michael

    2009-10-01

    The spatial distributions and activities of ammonia oxidizing bacteria (AOB) and polyphosphate accumulating organisms (PAOs) were investigated for a novel laboratory-scale sequencing batch pumped-flow biofilm reactor (PFBR) system that was operated for carbon, nitrogen and phosphorus removal. The PFBR comprised of two 16.5l tanks (Reactors 1 and 2), each with a biofilm module of 2m(2) surface area. To facilitate the growth of AOB and PAOs in the reactor biofilms, the influent wastewater was held in Reactor 1 under stagnant un-aerated conditions for 6 h after feeding, and was then pumped over and back between Reactors 1 and 2 for 12 h, creating aerobic conditions in the two reactors during this period; as a consequence, the biofilm in Reactor 2 was in an aerobic environment for almost all the 18.2 h operating cycle. A combination of micro-sensor measurements, molecular techniques, batch experiments and reactor studies were carried out to analyse the performance of the PFBR system. After 100 days operation at a filtered chemical oxygen demand (COD(f)) loading rate of 3.46 g/m(2) per day, the removal efficiencies were 95% COD(f), 87% TN(f) and 74% TP(f). While the PFBR microbial community structure and function were found to be highly diversified with substantial AOB and PAO populations, about 70% of the phosphorus release potential and almost 100% of the nitrification potential were located in Reactors 1 and 2, respectively. Co-enrichment of AOB and PAOs was realized in the Reactor 2 biofilm, where molecular analyses revealed unexpected microbial distributions at micro-scale, with population peaks of AOB in a 100-250 microm deep sub-surface zone and of PAOs in the 0-150 microm surface zone. The micro-distribution of AOB coincided with the position of the nitrification peak identified during micro-sensor analyses. The study demonstrates that enrichment of PAOs can be realized in a constant or near constant aerobic biofilm environment. Furthermore, the findings suggest

  3. Process Inherent Ultimate Safety (PIUS) reactor evaluation study: Final report

    International Nuclear Information System (INIS)

    1987-02-01

    This report presents the results of an independent study by United Engineers and Constructors (UNITED) of the SECURE-P Process Inherent Ultimate Safety (PIUS) Reactor Concept which is presently under development by the Swedish light water reactor vendor ASEA-ATOM of Vasteras, Sweden. This study was performed to investigate whether there is any realistic basis for believing that the PIUS reactor could be a viable competitor in the US energy market in the future. Assessments were limited to the technical, economic and licensing aspects of PIUS. Socio-political issues, while certainly important in answering this question, are so broad and elusive that it was considered that addressing them with the limited perspective of one small group from one company would be of questionable value and likely be misleading. Socio-political issues aside, the key issue is economics. For this reason, the specific objectives of this study were to determine if the estimated PIUS plant cost will be competitive in the US market and to identify and evaluate the technical and licensing risks that might make PIUS uneconomical or otherwise unacceptable

  4. Data acquisition for the LVR-15 research reactor. Final report

    International Nuclear Information System (INIS)

    Dusek, J.; Holy, J.; Rysavy, J.

    1993-11-01

    The activities are reviewed carried out under contract No. 5686 between the IAEA and the Nuclear Research Institute at Rez. A list of components, their description and block diagrams of the LVR-15 reactor are presented. Totally, 40 failures during testing and 48 failures during operation were recorded for the period 1991 to 1993. The failure causes and development are briefly described. Information on the failures was classified and included into the system. The contribution to data classification and processing is presented. A number of additional variants are pointed out, of the exact use of non-parametric and parametric statistical methods when developing the comprehensive probabilistic model. A list is given of initiating events as starting points of accident sequences collected from the operating experience. The report consists of three supplements: (i) Data collection on the LVR-15 research reactor; (ii) Some statistical methods for the data processing; (iii) Initiating events data of research reactor for the use of probabilistic safety assessment. (J.B.) 54 tabs., 17 figs., 14 refs

  5. Sensitivity analysis of the reactor safety study. Final report

    International Nuclear Information System (INIS)

    Parkinson, W.J.; Rasmussen, N.C.; Hinkle, W.D.

    1979-01-01

    The Reactor Safety Study (RSS) or Wash 1400 developed a methodology estimating the public risk from light water nuclear reactors. In order to give further insights into this study, a sensitivity analysis has been performed to determine the significant contributors to risk for both the PWR and BWR. The sensitivity to variation of the point values of the failure probabilities reported in the RSS was determined for the safety systems identified therein, as well as for many of the generic classes from which individual failures contributed to system failures. Increasing as well as decreasing point values were considered. An analysis of the sensitivity to increasing uncertainty in system failure probabilities was also performed. The sensitivity parameters chosen were release category probabilities, core melt probability, and the risk parameters of early fatalities, latent cancers and total property damage. The latter three are adequate for describing all public risks identified in the RSS. The results indicate reductions of public risk by less than a factor of two for factor reductions in system or generic failure probabilities as high as one hundred. There also appears to be more benefit in monitoring the most sensitive systems to verify adherence to RSS failure rates than to backfitting present reactors. The sensitivity analysis results do indicate, however, possible benefits in reducing human error rates

  6. Gas-core reactor power transient analysis. Final report

    International Nuclear Information System (INIS)

    Kascak, A.F.

    1972-01-01

    The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)

  7. Summary report of the final technical meeting on 'International Reactor Dosimetry File: IRDF-2002'

    International Nuclear Information System (INIS)

    Griffin, Patrick J.; Paviotti-Corcuera, R.

    2003-10-01

    Presentations, recommendations and conclusions of the Final Technical Meeting on 'International Reactor Dosimetry File: IRDF-2002' are summarized in this report. The main aims of this meeting were to discuss scientific and technical matters related to reactor dosimetry and to assign responsibilities for the preparation of the final version of the IRDF- 2002 library and the associated TECDOC. Tasks were assigned and deadlines were agreed. Participants emphasized that accurate and complete nuclear data for reactor dosimetry are essential to improve the assessment accuracies for reactor pressure vessel service lifetimes in nuclear power plants, as well as for other neutron metrology applications such as boron neutron capture therapy, therapeutic use of medical isotopes, nuclear physics measurements, and reactor safety applications. (author)

  8. A method for emergency flooding of the gland in the main circulating pump of pressurized water reactors and the connection therefor

    International Nuclear Information System (INIS)

    Skalicky, A.

    1978-01-01

    A method is described for the emergency flooding of the main circulating pumps of a pressurized water reactor such that in pressure drop in the flooded gland owing to pump suction, the pump head is connected by the pressure difference action to the flooding gland pipe, this via the heat sink and the filter of the emergency flooding circuit connected to the pump head. The emergency flooding circuit consisting of a pressure reducing valve, a check valve and a stop valve is connected to the pump head, behind the heat sink and the filter. The pressure reducing valve separates two pressure spaces. The former is connected to the pump head via the check valve and to the flooding pipe via the stop valve and the check valve. The latter is connected to the suction pump. (B.S.)

  9. Cryosorption pumping of deuterium by MS-5A at temperatures above 4.2 K for fusion reactor applications

    International Nuclear Information System (INIS)

    Fisher, P.W.; Watson, J.S.

    1977-01-01

    An Excalibur CVR-1106 cryosorption pump was fitted with a special cooling system to permit measurement of deuterium pumping speeds at temperatures between 6 and 20 0 K. Pumping speeds were found to be a function of feed rate, loading prior to each run, loading during runs, and thermal treatment between runs. At feed rates -4 Torr-l s -1 cm -2 , speeds were near 1 l s -1 cm -2 initially and declined monotonically with loading. At high feed rates, speeds reached a higher maximum (approx. 3 l s -1 cm -2 ) but also generally declined with loading; however, after 50 to 100 Torr l had accumulated, the pump underwent a spontaneous transition which effected a return to the original (high) pumping speed. This transition was accompanied by pressure spikes in the test chamber and temperature spikes in the sieve panel. Initial speeds for each consecutive run equaled the final speed for the preceding run if the pump was maintained at operating temperature; however, if it was warmed to 77 0 K and recooled, a restoration to the maximum speed was observed at the beginning of the next run

  10. Study of O2(1Δ) production by nuclear pumping. Final report, 12 April 1982-30 June 1983

    International Nuclear Information System (INIS)

    Miley, G.H.

    1983-01-01

    Studies concerned with the possible use of nuclear pumping to produce O 2 ( 1 Δ) are described. The goal is to develop methods to produce adequate O 2 ( 1 Δ) to use in an O 2 -I 2 mixing-type laser. Although the O 2 ( 1 Δ) concentrations generated in the nuclear induced discharge are too small for the operation of an iodine laser, one of the most promising approaches identified involves the initial production of ozone by nuclear pumping of a high pressure (approx. 5 atm) O 2 -He mixture. The ozone mixture is then photodecomposed by use of a KrF nuclear pumped flashlamp. Based on a combination of experimental and theoretical data, it is estimated that O 2 ( 1 Δ)/O 2 ratios > 0.3 can be produced this way with approx. 7 Torr of O 2 ( 1 Δ) in a mixture initially containing 2 Torr of O 3 . This is well above the minimum requirements for operation of an iodine laser. Experimental results of nuclear pumping O 2 -noble gas mixtures in a pulsed TRIGA reactor are described and scaling relations developed

  11. Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM; Simulacion de la obstruccion de flujo de una bomba jet en un reactor BWR con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Filio L, C., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose M. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2016-09-15

    This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

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

    Directory of Open Access Journals (Sweden)

    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

  13. Dual shell pressure balanced reactor vessel. Final project report

    International Nuclear Information System (INIS)

    Robertus, R.J.; Fassbender, A.G.

    1994-10-01

    The Department of Energy's Office of Energy Research (OER) has previously provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. Pacific Northwest Laboratory (PNL) engineers have recently developed and patented (U.S. patent 5,167,930 December 1, 1992) a concept for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The technology could be immediately useful in continuing commercialization of an R ampersand D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER. Innotek Corporation is a small business that would be one logical end-user of the DSPBV reactor technology. Innotek is working with several major U.S. engineering firms to evaluate the potential of this technology in the disposal of wastes from sewage treatment plants. PNL entered into a CRADA with Innotek to build a bench-scale demonstration reactor and test the system to advance the economic feasibility of a variety of high pressure chemical processes. Hydrothermal processing of corrosive substances on a large scale can now be made significantly safer and more economical through use of the DSPBV. Hydrothermal chemical reactions such as wet-air oxidation and supercritical water oxidation occur in a highly corrosive environment inside a pressure vessel. Average corrosion rates from 23 to 80 miles per year have been reported by Rice (1994) and Latanision (1993)

  14. Loviisa Power Station - final disposal of reactor waste

    International Nuclear Information System (INIS)

    Vaajasaari, Marja

    1987-01-01

    This report is based on the earlier published results of research into the properties and function of the candidate backfill materials. The results of the backfill material research, and the sealing concepts presented in the literature have been evaluatedand applied to sealing the Loviisa Reactor Waste Repository taking into consideration the local rock and groundwater conditions. It is emphasised that the applicability of the presented backfill materials and plugs to repository sealing must still be carefully evaluated on the basis of detailed studies and the local environment. 24 refs

  15. Component failures at pressurized water reactors. Final report

    International Nuclear Information System (INIS)

    Reisinger, M.F.

    1980-10-01

    Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis

  16. Study on the safety and on international developments of small modular reactors (SMR). Final report; Studie zur Sicherheit und zu internationalen Entwicklungen von Small Modular Reactors (SMR). Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Sebastian; Kruessenberg, Anne; Schaffrath, Andreas; Zipper, Reinhard

    2015-05-15

    This report documents the work and results of the project RS1521 Study of Safety and International Development of Small Modular Reactors (SMR). The aims of this study can be summarized as - setting-up of a sound overview on SMR, - identification of essential issues of reactor safety research and future R and D projects, - identification of needs for adaption of system codes of GRS used in reactor safety research. The sound overview consists of the descriptions of in total 69 SMR (Small and Medium Sized Rector) concepts (32 light water reactors (LWR), 22 liquid metal cooled reactors (LMR), 2 heavy water reactors, 9 gas cooled reactors (GCR) and 4 molten salt reactors (MSR)). It provides information about the core, the cooling circuits and the safety systems. The quality of the given specifications depends on their availability and public accessibility. Using the safety requirements for nuclear power plants and the fundamental safety functions, the safety relevant issues of the described SMR concepts were identified. The systems and measures used in the safety requirements were summarized in table form. Finally it was evaluated whether these systems and measures can be already simulated with the nuclear simulation chain of GRS and where further code development and validation is necessary. The results of this study can be summarized as follows: Many of the current SMR concepts are based on integral design. Here the main components like steam generators, intermediate heat exchangers or - in case of forced convection core cooling - main cooling pumps are located within the reactor pressure vessel. Most of the SMR fulfil highest safety standards and their safety concepts are mainly based on passive safety systems. The safety of theses reactors is achieved indefinitely without energy supply or additional measures of the operators. Since SMR's aim is not only to produce electricity but also couple them with chemical or physical process plants, the safety aspects of

  17. On-line vibration monitoring for submerged vertical shaft pumps: Final report

    International Nuclear Information System (INIS)

    Walter, T.J.; Marchione, M.M.

    1988-03-01

    The overall goal of this project was to extend to vertical pumps the capability that presently exists to monitor and diagnose vibration problems in horizontal pumps. Specific objectives included the development of analytical techniques to interpret vibration measurements, the verification of these techniqeus by in-plant tests, and the development of recommendations for procuring submergible vibration sensors. A concurrent analytical and experimental approach was used to accomplish these objectives. Rotordynamic analyses of selected pumps were accomplished, and each pump was instrumented and monitored for extended periods of time. The models were used to determine important frequencies and optimum sensor locations and to predict the effect that wear, imbalance, misalighment, and other mechanical changes would have on measured vibration. The predictive ability of the models was confirmed by making changes to instrumented pumps and observing actual changes in pump vibration. Simplified guidelines have been developed to assist the interested user to develop a computer model that realistically predicts the rotordynamic performance of the installed pump. Based on the work accomplished, typical sensor locations have been established. Experience gained in application of commercially available submergible sensors is also related. 11 refs., 11 figs

  18. Heat pump cycle by hydrogen-absorbing alloys to assist high-temperature gas-cooled reactor in producing hydrogen

    International Nuclear Information System (INIS)

    Satoshi, Fukada; Nobutaka, Hayashi

    2010-01-01

    A chemical heat pump system using two hydrogen-absorbing alloys is proposed to utilise heat exhausted from a high-temperature source such as a high-temperature gas-cooled reactor (HTGR), more efficiently. The heat pump system is designed to produce H 2 based on the S-I cycle more efficiently. The overall system proposed here consists of HTGR, He gas turbines, chemical heat pumps and reaction vessels corresponding to the three-step decomposition reactions comprised in the S-I process. A fundamental research is experimentally performed on heat generation in a single bed packed with a hydrogen-absorbing alloy that may work at the H 2 production temperature. The hydrogen-absorbing alloy of Zr(V 1-x Fe x ) 2 is selected as a material that has a proper plateau pressure for the heat pump system operated between the input and output temperatures of HTGR and reaction vessels of the S-I cycle. Temperature jump due to heat generated when the alloy absorbs H 2 proves that the alloy-H 2 system can heat up the exhaust gas even at 600 deg. C without any external mechanical force. (authors)

  19. Federal Technology Alert: Ground-Source Heat Pumps Applied to Federal Facilities-Second Edition; FINAL

    International Nuclear Information System (INIS)

    Hadley, Donald L

    2001-01-01

    This Federal Technology Alert, which was sponsored by the U.S. Department of Energy's Office of Federal Energy Management Programs, provides the detailed information and procedures that a Federal energy manager needs to evaluate most ground-source heat pump applications. This report updates an earlier report on ground-source heat pumps that was published in September 1995. In the current report, general benefits of this technology to the Federal sector are described, as are ground-source heat pump operation, system types, design variations, energy savings, and other benefits. In addition, information on current manufacturers, technology users, and references for further reading are provided

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

    Directory of Open Access Journals (Sweden)

    Lefèvre Grégory

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Austel, W.; Korbe, H.

    1986-01-01

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

  2. Coupling of high temperature nuclear reactor with chemical plant by means of steam loop with heat pump

    Directory of Open Access Journals (Sweden)

    Kopeć Mariusz

    2017-01-01

    Full Text Available High temperature nuclear reactors (HTR can be used as an excellent, emission-free source of technological heat for various industrial applications. Their outlet helium temperature (700°-900°C allows not only for heat supply to all processes below 600°C (referred to as “steam class”, but also enables development of clean nuclear-assisted hydrogen production or coal liquefaction technologies with required temperatures up to 900°C (referred to as “chemical class”. This paper presents the results of analyses done for various configurations of the steam transport loop coupled with the high-temperature heat pump designed for “chemical class” applications. The advantages and disadvantages as well as the key issues are discussed in comparison with alternative solutions, trying to answer the question whether the system with the steam loop and the hightemperature heat pump is viable and economically justified.

  3. Feedback control of primary pump using midplane temperature of lower density lock for a PIUS-type reactor

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Haga, Katsuhiro; Tamaki, Masayoshi

    1993-01-01

    A new automatic pump speed control system, using a measurement of the temperature distribution in the lower density lock, is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the lower density lock in order to prevent the poison water from penetrating into the core during normal operation. In a startup test, the effectiveness of this control system to bring the system quickly to the stable state from a very small initial temperature difference between top and bottom of the lower density lock has been confirmed. The effectiveness of the primary pump trip at the limit speed in the control system to shutdown the core power safely in an accident such as a loss-of-feedwater accident with and without the primary loop isolation has also been proved

  4. Drive reinforcement neural networks for reactor control. Final report

    International Nuclear Information System (INIS)

    Williams, J.G.; Jouse, W.C.

    1995-01-01

    In view of the loss of the third year funding, the scope of the project goals has been revised. The revision in project scope no longer allows for the detailed modeling of the EBR-11 start-up task that was originally envisaged. The authors are continuing, however, to model the control of the rapid power ascent of the University of Arizona TRIGA reactor using a model-based controller and using a drive reinforcement neural network. These will be combined during the concluding period of the project into a hierarchical control architecture. In addition, the modeling of a PWR feedwater heater has continued, and an autonomous fault-tolerant software architecture for its control has been proposed

  5. Next Generation Electromagnetic Pump Analysis Tools (PLM DOC-0005-2188). Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Stregy, Seth [GE Hitachi Nuclear Energy Americas LLC, Wilmington, NC (United States); Dasilva, Ana [GE Hitachi Nuclear Energy Americas LLC, Wilmington, NC (United States); Yilmaz, Serkan [GE Hitachi Nuclear Energy Americas LLC, Wilmington, NC (United States); Saha, Pradip [GE Hitachi Nuclear Energy Americas LLC, Wilmington, NC (United States); Loewen, Eric [GE Hitachi Nuclear Energy Americas LLC, Wilmington, NC (United States)

    2015-10-29

    This report provides the broad historical review of EM Pump development and details of MATRIX development under this project. This report summarizes the efforts made to modernize the legacy performance models used in previous EM Pump designs and the improvements made to the analysis tools. This report provides information on Tasks 1, 3, and 4 of the entire project. The research for Task 4 builds upon Task 1: Update EM Pump Databank and Task 3: Modernize the Existing EM Pump Analysis Model, which are summarized within this report. Where research for Task 2: Insulation Materials Development and Evaluation identified parameters applicable to the analysis model with Task 4, the analysis code was updated, and analyses were made for additional materials. The important design variables for the manufacture and operation of an EM Pump that the model improvement can evaluate are: space constraints; voltage capability of insulation system; maximum flux density through iron; flow rate and outlet pressure; efficiency and manufacturability. The development of the next-generation EM Pump analysis tools during this two-year program provides information in three broad areas: Status of analysis model development; Improvements made to older simulations; and Comparison to experimental data.

  6. Known volume air sampling pump. Final summary report Jun 1975--Nov 1976

    International Nuclear Information System (INIS)

    McCullough, J.E.; Peterson, A.

    1976-11-01

    The purpose of this development program was to design and develop a known volume air sampling pump for use in measuring the amount of radioactive material in the atmosphere of an underground uranium mine. The principal nuclear radiation hazard to underground uranium mines comes from the mine atmosphere. Daughter products of radon-222 are inhaled by the miner resulting in a relatively high lung cancer rate among these workers. Current exposure control practice employs spot sampling in working areas to measure working level values. Currently available personal air sampling pumps fail to deliver known volumes of air under widely changing differential pressures. A unique type of gas pump known as the scroll compressor, developed by Arthur D. Little, Inc., that has no values and few moving parts is expected to provide a practical, efficient, and dependable air pump for use in dosimeters. The three deliverable known volume air sampling pumps resulting from this work incorporate a scroll pump, drive motor, speed control electronics, and battery pack in a container suitable for attachment to a miner's belt

  7. Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

    International Nuclear Information System (INIS)

    Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

    1982-03-01

    A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

  8. Multi-Application Small Light Water Reactor Final Report

    International Nuclear Information System (INIS)

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-01-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO 2 , 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept. Applications such as cogeneration

  9. Experimental study of defect power reactor fuel. Final report

    International Nuclear Information System (INIS)

    Forsyth, R.S.; Jonsson, T.

    1982-01-01

    Two BWR fuel rods, one intact and one defect, with the same manufacturing and irradiation data have been examined in a comparative study. The defect rod has been irradiated in a defect condition during approximately one reactor cycle and has consequently some secondary defects. The defect rod has two penetrating defects at a distance of about 1.5 meters from each other. Comparison with the intact rod shows a large Cs loss from the defect rod, especially between the cladding defects, where the loss is measured to about 30 %. The leachibility in deionized water is higher for Cs, U and Cm for fuel from the defect rod. The leaching results are more complex for Sr-90, Pu and Am. The fuel in the defect rod has undergone a change of structure with gain growth and formation of oriented fuel structure. The cladding of the defect rod is hydrided locally in some parts of the lower part of the rod and furthermore over a more extended region near the end of the rod. (Authors)

  10. High-efficiency pump for space helium transfer. Final Technical Report

    International Nuclear Information System (INIS)

    Hasenbein, R.; Izenson, M.G.; Swift, W.L.; Sixsmith, H.

    1991-12-01

    A centrifugal pump was developed for the efficient and reliable transfer of liquid helium in space. The pump can be used to refill cryostats on orbiting satellites which use liquid helium for refrigeration at extremely low temperatures. The pump meets the head and flow requirements of on-orbit helium transfer: a flow rate of 800 L/hr at a head of 128 J/kg. The overall pump efficiency at the design point is 0.45. The design head and flow requirements are met with zero net positive suction head, which is the condition in an orbiting helium supply Dewar. The mass transfer efficiency calculated for a space transfer operation is 0.99. Steel ball bearings are used with gas fiber-reinforced teflon retainers to provide solid lubrication. These bearings have demonstrated the longest life in liquid helium endurance tests under simulated pumping conditions. Technology developed in the project also has application for liquid helium circulation in terrestrial facilities and for transfer of cryogenic rocket propellants in space

  11. Norwegian participation in the IEA Heat Pump Programme Annex 34 - final report

    Energy Technology Data Exchange (ETDEWEB)

    Nordtvedt, S.R.

    2012-07-01

    This report is the Norwegian team contribution to Task A within the IEA Heat Pumps Programme Annex 34 on #Left Double Quotation Mark#Thermally driven heat pumps#Right Double Quotation Mark#. It aims to give an overview of the thermally driven heat pump (TDHP) and chiller (TDC) market, recent developments in the area of TDHP and TDC, as well as to give an outlook of the progress of this technology. There are no Norwegian manufacturers on the TDHP market. There are only seven existing thermally driven absorption system installations in Norway. Three are district heat driven water/LiBr chillers, one is a steam driven water/LiBr chiller, one water/LiBr heat pump for flue gas condensation in a wood chip heating unit, and two gas driven ammonia-water chillers. The future market for thermally driven heat pumps and chillers in Norway is expected to be in combination of biomass, district heat or waste heat. (Author)

  12. Loviisa power station - final disposal of reactor waste

    International Nuclear Information System (INIS)

    Kankainen, Tuovi

    1986-10-01

    This study forms a part of the research done to assess the suitability of the rapakivi granitic bedrock of the island of Haestholmen, southern Finland, for the management of reactor waste. The aim is to assess the residence time and the origin of the groundwater. In addition, microfossil analyses and conservative ion data were used in deciphering the origin of the groundwater. Fracture mineral studies were limeted to 13 C determinations on two fracture calcites. Groundwater was sampled at several levels of four drill holes, reaching to a depth of some 200 m. The isotopic results were compared with those of water from a percussion drill hole, shallow dug wells, and the Gulf of Finland. The main conclusions are based on 3 H bundances in groundwater, mean residence time of groundwater deduced from 14 C analyses, and stabile isotope content of groundwater, combined with conservative ion data. Additional information was gained from activity ratios of uranium, and sulphur isotope ratios of sulphate. The groundwater of Haestholmen consists of a surface layer of fresh water, and deeper down, of saline water. The fresh water flows and changes rapidly; most of it has precipitated and infiltrated less than 30 years ago. It intermixes with saline water only at the fresh-saline groundwater interface. The saline water underneath the intermediate zone is relatively stagnant. It mainly consists of sea water from the Litorina Sea stage, intermixed with less than 20% glacial melt water. The evolution of the Haestholmen groundwater towards its present stage began during the melting phase of the Weichselian glaciation. Then the groundwater conditions chanced, and infiltration of melt water along open fractures in the bedrock occured. During the Litorian Sea stage heavy saline Litorina sea water slowly infiltrated in the bedrock and displaced the fresh water almost totally. The Haestholmen island rose above the sea level more than 4000 years ago. Then formation of the surficial layer

  13. Advanced steam cycles for light water reactors. Final report

    International Nuclear Information System (INIS)

    Mitchell, R.C.

    1975-07-01

    An appraisal of the potential of adding superheat to improve the overall LWR plant cycle performance is presented. The study assesses the economic and technical problems associated with the addition of approximately 500 0 F of superheat to raise the steam temperature to 1000 0 F. The practicality of adding either nuclear or fossil superheat to LWR's is reviewed. The General Electric Company Boiling Water Reactor (BWR) model 238-732 (BWR/6) is chosen as the LWR starting point for this evaluation. The steam conditions of BWR/6 are representative of LWR's. The results of the fossil superheat portion of the evaluation are considered directly applicable to all LWR's. In spite of the potential of a nuclear superheater to provide a substantial boost to the LWR cycle efficiency, nuclear superheat offers little promise of development at this time. There are difficult technical problems to resolve in the areas of superheat fuel design and emergency core cooling. The absence of a developed high integrity, high temperature fuel for operation in the steam/water environment is fundamental to this conclusion. Fossil superheat offers the potential opportunity to utilize fossil fuel supplies more efficiently than in any other mode of central station power generation presently available. Fossil superheat topping cycles evaluated included atmospheric fluidized beds (AFB), pressurized fluidized beds, pressurized furnaces, conventional furnaces, and combined gas/steam turbine cycles. The use of an AFB is proposed as the preferred superheat furnace. Fossil superheat provides a cycle efficiency improvement for the LWR of two percentage points, reduces heat rejection by 15 percent per kWe generated, increases plant electrical output by 54 percent, and burns coal with an incremental net efficiency of approximately 40 percent. This compares with a net efficiency of 36--37 percent which might be achieved with an all-fluidized bed fossil superheat plant design

  14. A pump/intermediate heat exchanger assembly for a liquid metal reactor

    International Nuclear Information System (INIS)

    Nathenson, R.D.; Alexion, C.C.; Sumpman, W.C.

    1987-01-01

    A heat exchanger and electromagnetic pump assembly is disclosed comprising a heat exchanger housing defining an annularly shaped cavity and supporting therein a plurality of heat transfer tubes. An electromagnetic pump disposed beneath the heat exchanger comprises a circular array of flow couplers. Each flow coupler comprises a pump duct receiving primary liquid metal and a generator duct receiving a pumped intermediate liquid metal. A first plenum chamber is in communication with the generator ducts of all the flow couplers and receives intermediate liquid metal from inlet duct. The generator ducts exit their flows of intermediate liquid metal to a second plenum chamber in communication with the heat exchanger annularly shaped cavity to permit the flow of the intermediate liquid metal therethrough. A third plenum chamber receives collectively the flows of the primary liquid metal from the tubes and directs the primary liquid metal to the pump ducts of the flow couplers. The annular magnetic field of the electromagnetic pump is produced by a circular array of electromagnets having hollow windings cooled by a flow of intermediate liquid metal via tubes and manifolds. The leads to the electromagnets pass through an annular space around the inlet duct. (author)

  15. A study on the jet pump characteristic curve in boiling water reactor

    International Nuclear Information System (INIS)

    Liao, L.Y.

    1990-01-01

    The jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From the investigation of the momentum equations, it is found that the normal quadrant jet pump models of these codes are essentially the same. In this paper, it is found that the relationship between the flow ratio, M, and the heat ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions the M - N characteristic curve will not change with power level, recirculation pump speed and loop flow rate. The effect of the gravitational head on the M - N curve has been found to be significant for low flow conditions. As a result, a guideline has been given to the definition of the specific energy (or the head ratio). Sensitivity studies on the key parameters have been performed. It is found that the generic M - N curve should not be used for a jet pump which does not have the same nozzle to throat area ratio as that of the generic jet pump

  16. Remote servicing considerations for near term tokamak power reactors (TNS). Final summary

    International Nuclear Information System (INIS)

    Spampinato, P.T.

    1977-01-01

    Next generation Tokamaks require special consideration for remote servicing. Three major problems are highlighted: (1) movement of heavy components, (2) remote connection/disconnection of joints, and (3) remote cutting, welding, and leak detection. The first problem is assumed to be handled with existing expertise and is not considered. The remaining problems are thought to be minimized by considering two engineering departures from conventional tokamak design; locating the field shaping coils outside of the toroidal coils and enclosing the total device within an evacuated reactor cell. Five topics under this vacuum building concept are discussed: incremental cost, vacuum pumping, tritium containment, activation topology, and first year operations

  17. Feed-pump hydraulic performance and design improvement, Phase I: research program design. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Brown, W.H.; Gopalakrishnan, S.; Fehlau, R.; Thompson, W.E.; Wilson, D.G.

    1982-03-01

    As a result of prior EPRI-sponsored studies, it was concluded that a research program should be designed and implemented to provide an improved basis for the design, procurement, testing, and operation of large feed pumps with increased reliability and stability over the full range of operating conditions. This two-volume report contains a research plan which is based on a review of the present state of the art and which defines the necessary R and D program and estimates the benefits and costs of the program. The recommended research program consists of 30 interrelated tasks. It is designed to perform the needed research; to verify the results; to develop improved components; and to publish computer-aided design methods, pump specification guidelines, and a troubleshooting manual. Most of the technology proposed in the research plan is applicable to nuclear power plants as well as to fossil-fired plants. This volume discusses the design, performance and failures of feed pumps, and recommendations for research on pump dynamics, design, and specifications.

  18. PWR [pressurized water reactor] pressurizer transient response: Final report

    International Nuclear Information System (INIS)

    Murphy, S.I.

    1987-08-01

    To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model

  19. Development and Validation of a Gas-Fired Residential Heat Pump Water Heater - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Michael Garrabrant; Roger Stout; Paul Glanville; Janice Fitzgerald; Chris Keinath

    2013-01-21

    For gas-fired residential water heating, the U.S. and Canada is predominantly supplied by minimum efficiency storage water heaters with Energy Factors (EF) in the range of 0.59 to 0.62. Higher efficiency and higher cost ($700 - $2,000) options serve about 15% of the market, but still have EFs below 1.0, ranging from 0.65 to 0.95. To develop a new class of water heating products that exceeds the traditional limit of thermal efficiency, the project team designed and demonstrated a packaged water heater driven by a gas-fired ammonia-water absorption heat pump. This gas-fired heat pump water heater can achieve EFs of 1.3 or higher, at a consumer cost of $2,000 or less. Led by Stone Mountain Technologies Inc. (SMTI), with support from A.O. Smith, the Gas Technology Institute (GTI), and Georgia Tech, the cross-functional team completed research and development tasks including cycle modeling, breadboard evaluation of two cycles and two heat exchanger classes, heat pump/storage tank integration, compact solution pump development, combustion system specification, and evaluation of packaged prototype GHPWHs. The heat pump system extracts low grade heat from the ambient air and produces high grade heat suitable for heating water in a storage tank for domestic use. Product features that include conventional installation practices, standard footprint and reasonable economic payback, position the technology to gain significant market penetration, resulting in a large reduction of energy use and greenhouse gas emissions from domestic hot water production.

  20. Flow in water-intake pump bays: A guide for utility engineers. Final report

    International Nuclear Information System (INIS)

    Ettema, R.

    1998-09-01

    This report is intended to serve as a guide for power-plant engineers facing problems with flow conditions in pump bays in water-intake structures, especially those located alongside rivers. The guide briefly introduces the typical prevailing flow field outside of a riverside water intake. That flow field often sets the inflow conditions for pump bays located within the water intake. The monograph then presents and discusses the main flow problems associated with pump bays. The problems usually revolve around the formation of troublesome vortices. A novel feature of this monograph is the use of numerical modeling to reveal diagnostically how the vortices form and their sensitivities to flow conditions, such as uniformity of approach flow entering the bay and water-surface elevation relative to pump-bell submergence. The modeling was carried out using a computer code developed specially for the present project. Pump-bay layouts are discussed next. The discussion begins with a summary of the main variables influencing bay flows. The numerical model is used to determine the sensitivities of the vortices to variations in the geometric parameters. The fixes include the use of flow-control vanes and suction scoops for ensuring satisfactory flow performance in severe flow conditions; notably flows with strong cross flow and shallow flows. The monograph ends with descriptions of modeling techniques. An extensive discussion is provided on the use of numerical model for illuminating bay flows. The model is used to show how fluid viscosity affects bay flow. The effect of fluid viscosity is an important consideration in hydraulic modeling of water intakes

  1. Norwegian participation in IEA Heat Pump Programme Annex 29 - final report

    International Nuclear Information System (INIS)

    Stene, Joern

    2006-06-01

    Ground source heat is a collective name for tapping, recharging and storing thermal energy at moderate temperature in the bedrock, ground water and soil (renewable energy). By means of heat pumps, ground source heating systems can be used as highly energy efficient and environmentally-friendly heating and cooling in all kinds of buildings. IEA's heat pump programme (IEA HPP) has in the period March 2004 to June 2006 carried out an Annex on heating pump systems based on ground source heat: 'Annex 29 - Ground-source heat pumps overcoming market and technical barriers'. The project has i.a. covered the development of more cost- and energy efficient solutions, identification and analysis of market obstacles, technology transfer and the task of making visible the energy economising potential and environmental gains that this type of systems represent. The six participating countries have been Canada, Japan, Norway, Sweden, USA and Austria (operating agent). Norway's participation in IEA Annex 29 has been financed by ENOVA SF, while SINTEF Energy research AS has been in charge of planning and executing the Norwegian project activities. This report provides a general presentation of ground source heat pump systems, and gives a summary of the most important Norwegian activities in IEA Annex 29, which includes: The development of a Norwegian internet home page (http://www.energy.sintef.no/prosjekt/annex29), preparation of Norwegian status report, initiation, coordination and carrying out of Norwegian research projects and reports, communication of results (technology transfer) in Norway, networking activities towards important market actors in Norway, participation and delivering discourse at international seminars (workshops), and participation at international working sessions (ml)

  2. Sealing, especially of dropped ceiling of cubide where main circulating pump of nuclear reactor is installed

    International Nuclear Information System (INIS)

    Paucek, V.

    1980-01-01

    The sealing of the dropped ceiling of the cubicle for main circulating pump installation consists of a reinforcing ring in the space of the pump installation and the top annulus of biological shielding whose inner diameter is smaller than the outer diameter of the bottom biological shielding annulus mounted onto the pump body. The sealing comprises a vertically slidable support ring provided on its underside with an outer ring insert and an outer splice bar for functional gap formation. Slidably mounted in the gap is a guide annular plate tightly attached by means of an inner splice bar and bolts to the bottom biological shielding annulus. Spacers are provided between the reinforcing ring underside and the support ring top surface. (B.S.)

  3. Systems for apartment buildings heat pumps. Final report; System foer fastighetsvaermepumpar. Slutrapport

    Energy Technology Data Exchange (ETDEWEB)

    Aakervall, Daniel (WSP Environmental, Stockholm (Sweden)); Rogstam, Joergen; Grotherus, Maarten (Sveriges Energi- och Kylcentrum, Katrineholm (Sweden))

    2009-05-15

    A fast growing segment of the heat pump business is the apartment building applications. Historically the experience base related to such installations is limited. However, this application is more complex than the much more widely spread domestic applications. The core idea of the project was to generate and collect information to avoid mistakes and to encourage the heat pump technology. By interviewing the 'market' and compiling the information good recommendations has been achieved. It is of great importance to convey the best available recommendations to the installers on the market to avoid pit holes and highlight the opportunities for installers and end consumers. A website has been design and built, www.sfvp.se, which contains useful tools for installers and potential system customers. A number of calculation tools to estimate heat requirement, cost of different heating systems, primary energy need, LCC, etc. are all available. The core of the site is the database containing heat pump installations with related data of importance. The data available is such key data as performance figures, cost of installation, etc. Today 104 system installations are collected in the database and these are marked in a colour coding to indicate the quality of the input data. Installations verified with measurement are given higher significance. It has been an unexpected challenge to find documented systems, so one of the conclusions of the project is that there is a great need for further measurement on the field. More information should be directed to the customers to request such equipment when systems are installed. The database enables statistical analysis of the key figures and it can be seen that the average seasonal COP is 3.2 and there are small differences between exhaust air and ground source heat pumps. It should be emphasised that the number of installations do not give statistical confidence for all kinds of analysis yet. Field measurements in apartment

  4. Diode-Pumped Mode-Locked LiSAF Laser; FINAL

    International Nuclear Information System (INIS)

    None

    1996-01-01

    Under this contract we have developed Cr(sup 3+):LiSrAlF(sub 6) (Cr:LiSAF, LiSAF) mode-locked lasers suitable for generation of polarized electrons for CEBAF. As 670 nm is an excellent wavelength for optical pumping of Cr:LiSAF, we have used a LIGHTWAVE developed 670 nm diode pump module that combines the output of ten diode lasers and yields approximately 2 Watts of optical power. By the use of a diffraction limited pump beam however, it is possible to maintain a small mode size through the length of the crystal and hence extract more power from Cr:LiSAF laser. For this purpose we have developed a 1 Watt, red 660nm laser (LIGHTWAVE model 240R) which serves as an ideal pump for Cr:LiSAF and is a potential replacement of costly and less robust krypton laser. This new system is to compliment LIGHTWAVE Series 240, and is currently being considered for commercialization. Partially developed under this contract is LIGHTWAVEs product model 240 which has already been in our production lines for a few months and is commercially available. This laser produces 2 Watts of output at 532 nm using some of the same technology developed for production of the 660nm red system. It is a potential replacement for argon ion lasers and has better current and cooling requirements and is an excellent pump source for Ti:Al(sub 2)O(sub 3). Also, as a direct result of this contract we now have the capability of commercially developing a mode-locked 100MHz Cr:LiSAF system. Such a laser could be added to our 100 MHz LIGHTWAVE Series 131. The Series 131 lasers provide pico second pulses and were originally developed under another DOE SBIR. Both models of LIGHTWAVE Series 240 lasers, the fiber coupled pump module and the 100MHz LiSAF laser of Series 131 have been partially developed under this contract, and are commercially competitive products

  5. Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report; WTZ mit Russland. Transientenanalysen fuer wassergekuehlte Kernreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Kozmenkov, Yaroslav [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Institute of Physics and Power Engineering, Obninsk (Russian Federation); Pivovarov, Valeri; Matveev, Yurij [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    2010-12-15

    The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

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

    International Nuclear Information System (INIS)

    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

  7. Finalize field testing of cold climate heat pump (CCHP) based on tandem vapor injection compressors

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Bo [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Baxter, Van D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Abdelaziz, Omar [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rice, C. Keith [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    This report describes the system diagram and control algorithm of a prototype air-source cold climate heat pump (CCHP) using tandem vapor injection (VI) compressors. The prototype was installed in Fairbanks, Alaska and underwent field testing starting in 09/2016. The field testing results of the past six months, including compressor run time fractions, measured COPs and heating capacities, etc., are presented as a function of the ambient temperature. Two lessons learned are also reported.

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

    International Nuclear Information System (INIS)

    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

  9. Tritium system for a tokamak reactor with a self-pumped limiter

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Sze, D.K.

    1986-01-01

    Benefits of the self-pumping system are the elimination of vacuum ducts, pumps, and penetration shielding (except for a very small startup system), and the reduction of tritium recycle and refueling. In addition, a self-pumped system may perform better and last longer than alternative systems such as a pumped limiter. The reference case here is a self-cooled lithium/vanadium blanket with a first wall/limiter. This concept combines the functions of first wall and limiter into a single first-wall structure. The wall is shaped in accordance with the outermost plasma flux surface. Trapping material is added to the plasma scrape-off or edge region where it is transported to the wall. The entire wall area is used for helium trapping. The tritium inventory, tritium permeation rate, and plasma protium concentration for the vanadium wall as a function of the number of years of operation are calculated. The tritium inventory is acceptable, the protium concentration in the plasma is acceptably small, and the tritium permeation rate is moderate. At the start of operation, it is equal to about five times the tritium burnup rate. This tritium will enter the coolant and the cost of the blanket tritium recovery system will be higher

  10. Magnetohydrodynamic pumps for molten salts in cooling loops of high-temperature nuclear reactors

    Czech Academy of Sciences Publication Activity Database

    Doležel, Ivo; Kotlan, V.; Ulrych, B.

    2011-01-01

    Roč. 87, č. 5 (2011), s. 28-33 ISSN 0033-2097 Grant - others:GA MŠk(CZ) MEB051041 Institutional research plan: CEZ:AV0Z20570509 Keywords : magnetohydrodynamic pump * molten salt * electric field Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 0.244, year: 2011 http://pe.org.pl/

  11. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    International Nuclear Information System (INIS)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy's (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher's workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead

  12. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  13. Final qualification of an industrial wide range neutron instrumentation in the Osiris MTR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Barbot, L.; Normand, S. [CEA, LIST, Laboratoire Capteur et Architectures Electroniques, F-91191 Gif Sur Yvette (France); Pasdeloup, P. [AREVA TA, Controle Commande and Mesures, F-13762 Les Milles (France); Lescop, B. [CEA, INSTN, UEIN, F-91191 Gif Sur Yvette (France)

    2009-07-01

    This work deals with the final qualification of the IRINA in-core neutron flux measurement system in the MTR Osiris reactor. A specific irradiation device has been set up to validate the last changes in the complete system (electronic, transmitting cable and monitor). Experimental results show the IRINA measurement system meet entirely the in-core reactor conditions requirements: a thermal neutron flux from 10{sup 7} n.cm{sup -2}.s{sup -1} up to 10{sup 14} n.cm{sup -2}.s{sup -1} and a temperature of 300 C degrees during a minimum operating time of 1000 hours. (authors)

  14. The slightly-enriched spectral shift control reactor. Final report, September 30, 1988--September 30, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Martin, W.R.; Lee, J.C.; Larsen, E.W. [Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering; Edlund, M.C. [Virginia Polytechnic Inst. and State Univ., Blacksburg, VA (United States). Dept. of Mechanical and Nuclear Engineering

    1991-11-01

    An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technology retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.

  15. Research and development of an air-cycle heat-pump water heater. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Dieckmann, J.T.; Erickson, A.J.; Harvey, A.C.; Toscano, W.M.

    1979-10-01

    A prototype reverse Brayton air cycle heat pump water heater has been designed and built for residential applications. The system consists of a compressor/expander, an air-water heat exchanger, an electric motor, a water circulation pump, a thermostat, and fluid management controls. The prototype development program consisted of a market analysis, design study, and development testing. A potential residential market for the new high-efficiency water heater of approximately 480,000 units/y was identified. The retail and installation cost of this water heater is estimated to be between $500 and $600 which is approximately $300 more than a conventional electric water heater. The average payback per unit is less than 3-1/2 y and the average recurring energy cost savings after the payback period is approximately $105/y at the average seasonal coefficient of performance (COP) of 1.7. As part of the design effort, a thermodynamic parametric analysis was performed on the water heater system. It was determined that to obtain a coefficient of performance of 1.7, the isentropic efficiency of both the compressor and the expander must be at least 85%. The selected mechanical configuration is described. The water heater has a diameter of 25 in. and a height of 73 in. The results of the development testing of the prototype water heater system showed: the electrical motor maximum efficiency of 78%; the compressor isentropic efficiency is 95 to 119% and the volumetric efficiency is approximately 85%; the expander isentropic efficiency is approximately 58% and the volumetric efficiency is 92%; a significant heat transfer loss of approximately 16% occurred in the expander; and the prototype heat pump system COP is 1.26 which is less than the design goal of at least 1.7. Future development work is recommended.

  16. Feasibility study of a Green Power Plant. Final report. [Offshore pumped hydro storage

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-03-15

    This project is a technical evaluation and a feasibility study of a concept called the Green Power Plant (GPP), developed by Seahorn Energy Aps. The Green Power Plant is an offshore pumped hydro storage facility constructed from prefabricated elements and with integrated renewable energy production. Pumped hydro storage is a known technology with a proven roundtrip energy storage efficiency of 80%. The focus of the GPP project is on simplifying and industrializing the construction of the reservoir wall, thereby achieving a cost efficient solution. The reservoir structure is dependent on the site on which the reservoir is established, thus Kriegers Flak in the Baltic Sea has been chosen as basis for the technical evaluation. As soil conditions vary, the technical evaluations have been based on a general soil profile. A water depth of 25m has been chosen as basis for the evaluation. A reservoir with a diameter of 2 km has been evaluated as baseline scenario. Feasibility of the GPP was evaluated based on the cost and income estimates. For the baseline scenario an internal rate of return of 6.6% was found for a period of 35 years. A sensitivity analysis reveals internal rates of return over 35 years varying from 4.9% to 10.9%. Especially larger reservoir diameters increase profitability of the GPP. The results from this project will be utilized in raising funds for further development of the GPP concept. Seahorn Energy Aps aims at optimizing the wind turbine integration, the steel pile wall structure and the pump-turbine integration in a future project towards construction of a demonstration facility. (LN)

  17. Large scale breeder reactor plant prototype mechanical pump conceptual design study

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

    This report includes engineering memorandums, drawings, key feature descriptions, and other data. Some of the reports, such as manufacturability and some stress analysis, were done by consultants for Byron Jackson. Review of this report indicates that the design is feasible. The pump can be manufactured to system and specification requirements. The overall length and weight of some pieces will require special consideration, but is within the scope of equipment and technology available today. The fabricated parts are large and heavy, but can be manufactured and machined. Only the high temperature is unique to this size, since previous sodium pumps were smaller. Nondestructive tests as required by the Code are described and are feasible. The performance test of the prototype has been studied thoroughly. It is feasible for a cold water test. There are some problem areas. However, all of them can be solved. Development needs include building and testing a small scale model.

  18. Development of diode-pumped solid-state laser HALNA for fusion reactor driver

    International Nuclear Information System (INIS)

    Kawashima, Toshiyuki; Kanabe, Tadashi; Matsumoto, Osamu

    2005-01-01

    The diode-pumped slab laser for inertial fusion energy driver has been demonstrated, which produces the 1053-nm output energy of 10 J at 10 Hz. The glass slab laser amplifier has been pumped by quasi-CW 290 kW AlGaAs laser-diode arrays at 803 nm. The optical system can compensate for thermal effects by use of zig-zag optical propagation, image-relayed telescope, and 45deg Faraday rotator. The output energy of 10.6 J at 1 Hz with the optical to optical conversion efficiency of 19.9% has been successfully obtained. Also the 10 Hz operation has been performed with a 5.1 J output energy. (author)

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

    International Nuclear Information System (INIS)

    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

  20. Feedback control of a primary pump for safe and stable operation of a PIUS-type reactor

    International Nuclear Information System (INIS)

    Tasaka, K.; Imai, S.; Masaoka, H.; Tamaki, M.; Kukita, Y.

    1993-01-01

    A new automatic pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the density lock in order to prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments such as start-up and power ramping tests for the stable normal operation and a loss-of-feedwater test for the safe shutdown in an accident condition, using a small scale atmospheric pressure test loop which simulated the PIUS principle. (orig.)

  1. Final Technical Report for EE0006091: H2Pump Hydrogen Recycling System Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Staudt, Rhonda [H2Pump LLC, Latham, NY (United States)

    2017-02-21

    The objective of this project is to demonstrate the product readiness and to quantify the benefits and customer value proposition of H2Pump’s Hydrogen Recycling System (HRS-100™) by installing and analyzing the operation of multiple prototype 100-kg per day systems in real world customer locations. The data gathered will be used to measure reliability, demonstrate the value proposition to customers, and validate our business model. H2Pump will install, track and report multiple field demonstration systems in industrial heat treating and semi-conductor applications. The customer demonstrations will be used to develop case studies and showcase the benefits of the technology to drive market adoption.

  2. Air-Source Integrated Heat Pump System Development – Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, Van D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rice, C. Keith [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Munk, Jeffrey D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ally, Moonis R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Shen, Bo [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Uselton, R. B. [Lennox Industries, Inc., Knoxville, TN (United States)

    2017-07-01

    Between October 2007 and September 2017, Oak Ridge National Laboratory (ORNL) and Lennox Industries, Inc. (Lennox) engaged in a Cooperative Research and Development Agreement (CRADA) to develop an air-source integrated heat pump (AS-IHP) system for the US residential market. The Lennox AS-IHP concept consisted of a high-efficiency air-source heat pump (ASHP) for space heating and cooling services and a separate heat pump water heater/dehumidifier (WH/DH) module for domestic water heating and dehumidification (DH) services. A key feature of this system approach with the separate WH/DH is capability to pretreat (i.e., dehumidify) ventilation air and dedicated whole-house DH independent of the ASHP. Two generations of laboratory prototype WH/DH units were designed, fabricated, and lab tested. Performance maps for the system were developed using the latest research version of the US Department of Energy/ORNL heat pump design model (Rice 1992; Rice and Jackson 2005; Shen et al. 2012) as calibrated against the lab test data. These maps served as the input to TRNSYS (Solar Energy Laboratory et al. 2010) to predict annual performance relative to a baseline suite of equipment meeting minimum efficiency standards in effect in 2006 (i.e., a combination of an ASHP with a seasonal energy efficiency ratio (SEER) of 13 and resistance water heater with an energy factor (EF) of 0.9). Predicted total annual energy savings (based on use of a two-speed ASHP and the second-generation WH/DH prototype for the AS-IHP), while providing space conditioning, water heating, and dehumidification for a tight, well-insulated 2600 ft2 (242 m2) house at three US locations, ranged from 33 to 36%, averaging 35%, relative to the baseline system. The lowest savings were seen at the cold-climate Chicago location. Predicted energy use for water heating was reduced by about 50 to 60% relative to a resistance WH.

  3. Internal pump monitoring device

    International Nuclear Information System (INIS)

    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.)

  4. Fast pyrolysis of biomass in the rotating cone reactor. Reactor development and operation. Final report

    International Nuclear Information System (INIS)

    Gansekoele, E.; Wagenaar, B.M.

    2001-07-01

    This report describes the design and characteristics of BTGs pyrolysis plant with a biomass throughput capacity of 50 kg per hour. The pilot plant has been developed for 2 reasons: to produce modest quantities of bio-oil for application purposes, and to generate know-how for the development of a larger 200 kg/hr pilot plant. The design of the 50 kg/hr plant continues the development line which started in 1995 when a similar unit was delivered to China. Major design improvements of the current pyrolysis unit are that it can be operated in a continuous mode and utilizes the combustion heat of the produced char to heat the pyrolysis process. A measurement program has meanwhile been executed as a means to characterize the pyrolysis plant. Results of the characterization study were the following: the pilot plant produces approx. 35 liters of bio-oil per hour and thus achieves a maximum oil yield of 70 weight percent. The bio-oil yield of the plant was inversely proportional with the reactor temperature and inversely proportional with the gas phase residence time. As a result of the pilot plant operation, a few tons of bio-oil have been produced; alongside with a bulk of know-how. All know-how has successfully been utilized in the development of the 200 kg per hour facility

  5. A conceptual composite blanket design for the Tokamak type of thermonuclear reactor incorporating thermoelectric pumping of liquid lithium

    International Nuclear Information System (INIS)

    Dutta Gupta, P.B.

    1981-01-01

    The conceptual liquid lithium blanket design for the tokamak type of thermonuclear reactor put forward is a modification of the initial simple but novel design concept enunciated earlier that exploits the availability of suitably oriented magnetic fields and temperature gradients within the blanket to pump the liquid as has been shown feasible by laboratory model experiments. The modular construction of the blanket cells is retained but the earlier simple back to back double spiralling channel module is replaced by a composite unit of three radially nested layer-structures to optimise heat and tritium extraction from the blanket. The layer-structure at the first wall generates liquid lithium circulation by thermoelectric magnetohydrodynamic forces and the segregated double spiralling channels serve as inlet-outlet driving devices. The outermost layer-structure is cooled by helium. Liquid lithium in the intermediate layer-structure is pumped at a very slow rate. The choice of the relative dimensional proportions of the three layer-structure and the channel cross-section, material property and the spiralling contour is of critical importance for the design. This paper presents the design data for a conceptual design of such a blanket with a 5000 MW (th) rating. (author)

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

    International Nuclear Information System (INIS)

    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

  7. Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Fromm, Bradley [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hauch, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sridharan, Kumar [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulation tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.

  8. Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

    International Nuclear Information System (INIS)

    Fromm, Bradley; Hauch, Benjamin; Sridharan, Kumar

    2016-01-01

    A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulation tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.

  9. Advanced variable speed air source integrated heat pump (AS-IHP) development - CRADA final report

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, Van D. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Rice, C. Keith [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Munk, Jeffrey D. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Ally, Moonis Raza [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Shen, Bo [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

    2015-09-30

    Between August 2011 and September 2015, Oak Ridge National Laboratory (ORNL) and Nordyne, LLC (now Nortek Global HVAC LLC, NGHVAC) engaged in a Cooperative Research and Development Agreement (CRADA) to develop an air-source integrated heat pump (AS-IHP) system for the US residential market. Two generations of laboratory prototype systems were designed, fabricated, and lab-tested during 2011-2013. Performance maps for the system were developed using the latest research version of the DOE/ORNL Heat Pump Design Model, or HPDM, (Rice 1991; Rice and Jackson 2005; Shen et al 2012) as calibrated against the lab test data. These maps were the input to the TRNSYS (SOLAR Energy Laboratory, et al, 2010) system to predict annual performance relative to a baseline suite of equipment meeting minimum efficiency standards in effect in 2006 (combination of 13 SEER air-source heat pump (ASHP) and resistance water heater with Energy Factor (EF) of 0.9). Predicted total annual energy savings, while providing space conditioning and water heating for a tight, well insulated 2600 ft2 (242 m2) house at 5 U.S. locations, ranged from 46 to 61%, averaging 52%, relative to the baseline system (lowest savings at the cold-climate Chicago location). Predicted energy use for water heating was reduced 62 to 76% relative to resistance WH. Based on these lab prototype test and analyses results a field test prototype was designed and fabricated by NGHVAC. The unit was installed in a 2400 ft2 (223 m2) research house in Knoxville, TN and field tested from May 2014 to April 2015. Based on the demonstrated field performance of the AS-IHP prototype and estimated performance of a baseline system operating under the same loads and weather conditions, it was estimated that the prototype would achieve ~40% energy savings relative to the minimum efficiency suite. The estimated WH savings were >60% and SC mode savings were >50%. But estimated SH savings were only about 20%. It is believed that had the test

  10. Estimation on the Pressure Loss of the Conceptual Primary Cooling System and Design of the Primary Cooling Pump for a Research Reactor

    International Nuclear Information System (INIS)

    Seo, Kyoung Woo; Oh, Jae Min; Park, Jong Hark; Chae, Hee Taek; Seo, Jae Kwang; Park, Cheon Tae; Yoon, Ju Hyeon; Lee, Doo Jeong

    2009-01-01

    A new conceptual primary cooling system (PCS) for a research reactor has been designed for an adequate cooling to the reactor core which has various powers ranging from 30MW through 80MW. The developed primary cooling system consisted of decay tanks, pumps, heat exchangers, vacuum breakers, some isolation and check valves, connection piping, and instruments. Because the system flow rate should be determined by the thermal hydraulic design analysis for the core, the heads to design the primary cooling pumps (PCPs) in a PCS will be estimated by the variable system flow rates. The heads of the part of a research reactor vessel was evaluated by the previous study. The various pressure losses of the PCS can be calculated by the dimensional analysis of the pipe flow and the head loss coefficient of the components. The purpose of this research is to estimate the various pressure losses and to design the PCPs

  11. Metal hydride/chemical heat-pump development project. Phase I. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Argabright, T.A.

    1982-02-01

    The metal hydride/chemical heat pump (MHHP) is a chemical heat pump containing two hydrides for the storage and/or recovery of thermal energy. It utilizes the heat of reaction of hydrogen with specific metal alloys. The MHHP design can be tailored to provide heating and/or cooling or temperature upgrading over a wide range of input and ambient temperatures. The system can thus be used with a variety of heat sources including waste heat, solar energy or a fossil fuel. The conceptual design of the MHHP was developed. A national market survey including a study of applications and market sectors was conducted. The technical tasks including conceptual development, thermal and mechanical design, laboratory verification of design and material performance, cost analysis and the detailed design of the Engineering Development Test Unit (EDTU) were performed. As a result of the market study, the temperature upgrade cycle of the MHHP was chosen for development. Operating temperature ranges for the upgrader were selected to be from 70 to 110/sup 0/C (160 to 230/sup 0/F) for the source heat and 140 to 190/sup 0/C (280 to 375/sup 0/F) for the product heat. These ranges are applicable to many processes in industries such as food, textile, paper and pulp, and chemical. The hydride pair well suited for these temperatures is LaNi/sub 5//LaNi/sub 4/ /sub 5/Al/sub 0/ /sub 5/. The EDTU was designed for the upgrade cycle. It is a compact finned tube arrangement enclosed in a pressure vessel. This design incorporates high heat transfer and low thermal mass in a system which maximizes the coefficient of performance (COP). It will be constructed in Phase II. Continuation of this effort is recommended.

  12. The evaluation of a 4000-home geothermal heat pump retrofit at Fort Polk, Louisiana: Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, P.J.; Shonder, J.A.

    1998-03-01

    This report documents an independent evaluation of an energy retrofit of 4,003 family housing units at Fort Polk, Louisiana, under an energy savings performance contract (ESPC). Replacement of the heating, cooling, and water heating systems in these housing units with geothermal heat pumps (GHPs) anchored the retrofit; low-flow shower heads and compact fluorescent lighting were also installed, as well as attic insulation where needed. Statistically valid findings indicate that the project will save 25.8 million kWh, or 32.5% of the pre-retrofit whole-community electrical consumption, and 100% of the whole-community natural gas previously used for space conditioning and water heating (260,000 therms) in a typical meteorological year. At the end-use level, the GHPs were found to save about 42% of the pre-retrofit electrical consumption for heating, cooling, and water heating in housing units that were all-electric in the pre-retrofit period. This report also demonstrates an improved method of predicting energy savings. Using an engineering model calibrated to pre-retrofit energy use data collected in the field, the method predicted actual energy savings on one of the electric feeders at Fort Polk with a very high degree of accuracy. The accuracy of this model was in turn dependent on data-calibrated models of the geothermal heat pump and ground heat exchanger that are described in this report. In addition this report documents the status of vertical borehole ground heat exchanger (BHEx) design methods at the time this project was designed, and demonstrates methods of using data collected from operating GHP systems to benchmark BHEx design methods against a detailed engineering model calibrated to date. The authors also discuss the ESPC`s structure and implementation and how the experience gained here can contribute to the success of future ESPCs.

  13. Heat pump centered integrated community energy systems: system development. Georgia Institute of Technology final report

    Energy Technology Data Exchange (ETDEWEB)

    Wade, D.W.; Trammell, B.C.; Dixit, B.S.; McCurry, D.C.; Rindt, B.A.

    1979-12-01

    Heat Pump Centered-Integrated Community Energy Systems (HP-ICES) show the promise of utilizing low-grade thermal energy for low-quality energy requirements such as space heating and cooling. The Heat Pump - Wastewater Heat Recovery (HP-WHR) scheme is one approach to an HP-ICES that proposes to reclaim low-grade thermal energy from a community's wastewater effluent. This report develops the concept of an HP-WHR system, evaluates the potential performance and economics of such a system, and examines the potential for application. A thermodynamic performance analysis of a hypothetical system projects an overall system Coefficient of Performance (C.O.P.) of from 2.181 to 2.264 for waste-water temperatures varying from 50/sup 0/F to 80/sup 0/F. Primary energy source savings from the nationwide implementation of this system is projected to be 6.0 QUADS-fuel oil, or 8.5 QUADS - natural gas, or 29.7 QUADS - coal for the period 1980 to 2000, depending upon the type and mix of conventional space conditioning systems which could be displaced with the HP-WHR system. Site-specific HP-WHR system designs are presented for two application communities in Georgia. Performance analyses for these systems project annual cycle system C.O.P.'s of 2.049 and 2.519. Economic analysis on the basis of a life cycle cost comparison shows one site-specific system design to be cost competitive in the immediate market with conventional residential and light commercial HVAC systems. The second site-specific system design is shown through a similar economic analysis to be more costly than conventional systems due mainly to the current low energy costs for natural gas. It is anticipated that, as energy costs escalate, this HP-WHR system will also approach the threshold of economic viability.

  14. Final safety and hazards analysis for the Battelle LOCA simulation tests in the NRU reactor

    International Nuclear Information System (INIS)

    Axford, D.J.; Martin, I.C.; McAuley, S.J.

    1981-04-01

    This is the final safety and hazards report for the proposed Battelle LOCA simulation tests in NRU. A brief description of equipment test design and operating procedure precedes a safety analysis and hazards review of the project. The hazards review addresses potential equipment failures as well as potential for a metal/water reaction and evaluates the consequences. The operation of the tests as proposed does not present an unacceptable risk to the NRU Reactor, CRNL personnel or members of the public. (author)

  15. Final hazard classification and auditable safety analysis for the 105-C Reactor Interim Safe Storage Project

    International Nuclear Information System (INIS)

    Rodovsky, T.J.; Larson, A.R.; Dexheimer, D.

    1996-12-01

    This document summarizes the inventories of radioactive and hazardous materials present in the 105-C Reactor Facility and the operations associated with the Interim Safe Storage Project which includes decontamination and demolition and interim safe storage of the remaining facility. This document also establishes a final hazard classification and verifies that appropriate and adequate safety functions and controls are in place to reduce or mitigate the risk associated with those operations

  16. Tritium inventory in fusion reactors. Summary report of the final research coordination meeting

    International Nuclear Information System (INIS)

    Clark, R.E.H.

    2007-11-01

    Detailed discussions were held during the final Research Coordination Meeting (RCM) at IAEA Headquarters on 25-27 September 2006, with the aim of reviewing the work accomplished by the Coordinated Research Project (CRP) on 'Tritium Inventory in Fusion Reactors'. Participants summarized the specific results obtained during the final phase of the CRP, and considered the impact of the data generated on the design of fusion devices. Conclusions were formulated and several specific recommendations for future fusion machines were agreed. The discussions, conclusions and recommendations of the RCM are briefly described in this report. (author)

  17. Thermodynamic analysis of the use a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water-splitting cycle for hydrogen production

    International Nuclear Information System (INIS)

    Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

    2008-01-01

    Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved 'steam' parameters (outlet temperatures up to 625degC and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600degC. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the 'nuclear' heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of

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

    International Nuclear Information System (INIS)

    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)

  19. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    2000-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  20. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1998-01-01

    Full text: In 1998 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  1. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1996-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  2. LMFBR with booster pump in pumping loop

    International Nuclear Information System (INIS)

    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

  3. BWR series pump recirculation system

    International Nuclear Information System (INIS)

    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

  4. Research of heat releasing element of an active zone of gaseous nuclear reactor with pumped through nuclear fuel - uranium hexafluoride (UF6)

    International Nuclear Information System (INIS)

    Batyrbekov, G.; Batyrbekov, E.; Belyakova, E.; Kunakov, S.; Koltyshev, S.

    1996-01-01

    The purpose of the offered project is learning physics and substantiation of possibility of creation gaseous nuclear reactor with pumped through nuclear fuel-hexafluoride of uranium (Uf6).Main problems of this work are'. Determination of physic-chemical, spectral and optical properties of non-equilibrium nuclear - excited plasma of hexafluoride of uranium and its mixtures with other gases. Research of gas dynamics of laminar, non-mixing two-layer current of gases of hexafluoride of uranium and helium at availability and absence of internal energy release in hexafluoride of uranium with the purpose to determinate a possibility of isolation of hexafluoride of uranium from walls by inert helium. Creation and research of gaseous heat releasing element with pumped through fuel Uf6 in an active zone of research nuclear WWR-K reactor. Objects of a research: Non-equilibrium nuclear - excited plasma of hexafluoride of uranium and its mixtures with other gases. With use of specially created ampoules will come true in-reactor probe and spectral diagnostics of plasma. Calculations of kinetics with the account of main elementary processes proceeding in it, will be carried out. Two-layer non-mixed streams of hexafluoride of uranium and helium at availability and absence of internal energy release. Conditions of obtaining and characteristics of such streams will be investigated. Gaseous heat releasing element with pumped through fuel - Uf6 in an active zone of nuclear WWR-K reactor

  5. 76 FR 17160 - Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs...

    Science.gov (United States)

    2011-03-28

    ... design certification (DC) application for new nuclear power reactors under Title 10 of the Code of... NUCLEAR REGULATORY COMMISSION [NRC-2010-0033; DC/COL-ISG-021] Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby...

  6. Exergy analysis of a system using a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water splitting cycle

    International Nuclear Information System (INIS)

    Granovskii, M.; Dincer, I.; Rosen, M. A.; Pioro, I

    2007-01-01

    The power generation efficiency of nuclear plants is mainly determined by the permissible temperatures and pressures of the nuclear reactor fuel and coolants. These parameters are limited by materials properties and corrosion rates and their effect on nuclear reactor safety. The advanced materials for the next generation of CANDU reactors, which employ steam as a coolant and heat carrier, permit the increased steam parameters (outlet temperature up to 625 degree C and pressure of about 25 MPa). Supercritical water-cooled (SCW) nuclear power plants are expected to increase the power generation efficiency from 35 to 45%. Supercritical water-cooled nuclear reactors can be linked to thermochemical water splitting cycles for hydrogen production. An increased steam temperature from the nuclear reactor makes it also possible to utilize its energy in thermochemical water splitting cycles. These cycles are considered by many as one of the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require a heat supply at the temperatures over 550-600 degree C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump which increases the temperature the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. A high temperature chemical heat pump which employs the reversible catalytic methane conversion reaction is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with a SCW nuclear plant on one side and thermochemical water splitting cycle on the other, increases the temperature level of the 'nuclear' heat and, thus, the intensity of

  7. Final environmental statement for La Crosse Boiling Water Reactor: (Docket No. 50-409)

    International Nuclear Information System (INIS)

    1980-04-01

    A Final Environmental Statement for the Dairyland Power Cooperative for the conversion from a provisional to a full-term operating license for the La Crosse Boiling Water Reactor, located in Vernon County, Wisconsin, has been prepared by the Office of Nuclear Reactor Regulation. This statement provides a summary of environmental impacts and adverse effects of operation of the facility, and a consideration of principal alternatives (including removal of LACBWR from service, alternative cooling methodology, and alternative waste treatment systems). Also included are the comments of federal, state, and local governmental agencies and certain non-governmental organizations on the La Crosse Draft Environmental Statement and staff responses to these comments. After weighing environmental, economic, and technical benefits and liabilities, the staff recommends conversion from a provisional operating license to a full-term operating license, subject to specific environmental protection limitations. An operational monitoring program shall be established as part of the Environmental Technical Specifications. 64 refs., 20 figs., 48 tabs

  8. 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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  12. Groundwater heat pump performance improvement with pre-coolers and pump modification: Final report for the 1985-86 SOMED (School of Mines and Energy Development) project year

    Energy Technology Data Exchange (ETDEWEB)

    Kavanaugh, S.

    1986-09-30

    Improved performance of groundwater heat pumps can be realized with a more effective and efficient utilization of the thermal properties of shallow groundwater. These systems circulate water from aquifers through water source heat pumps to achieved high efficiencies and capacities. This project concludes that a 10 to 15 percent cooling performance improvement can be realized by pre-cooling the room air with the 55/sup 0/ to 67/sup 0/F groundwater available in large portions of the Southeast. Proper design of these pre-coolers eliminates unnecessary auxiliary energy requirements. The efficiency of the overall system can be further improved with modifications to current methods of water circulation system design. Pressure requirements are minimized by maintaining a low unit inlet pressure (8 psig maximum), removing unnecessary loop restrictions and injection below the water table. Standard submersible water pumps exceed the resulting required size for residential groundwater heat pumps. Simple modifications can be made by the manufacturer to correct this problem. The result is an overall 15 to 40 percent performance improvement over high efficiency air source heat pumps with a simple payback of between 0 to 10 years in most cases.

  13. Three-dimensional finite-element analysis of the cellular convection phenomena in the Clinch River Breeder Reactor Plant prototype pump

    International Nuclear Information System (INIS)

    Silver, A.H.; Lee, J.Y.

    1983-01-01

    Cellular convection was studied rigorously during the development of the Clinch River Breeder Reactor Plant (CRBRP) Program Pumps. This paper presents the development of a three-dimensional finite-element heat transfer model which accounts for the cellular convection phenomena. A buoyancy driven cellular convection flow pattern is introduced in the annulus region between the upper inner structure and the pump tank. Steady-state thermal data were obtained for several test conditions for argon gas pressures up to 93 psig (741 kPa) and sodium operating temperatures to 1000 0 F (811 0 K). Test temperature distributions on the pump tank and inner structure were correlated with numerical results and excellent agreement was obtained

  14. Influence of gamma irradiation on the deterioration of reactor pressure vessel materials and on reactor dosimetry measurements. Final report

    International Nuclear Information System (INIS)

    Boehmer, B.; Konheiser, J.; Kumpf, H.; Noack, K.; Vladimirov, P.

    2002-10-01

    Radiation embrittlement of pressure vessel steel in mixed neutron-gamma fields is mostly determined by neutrons, but in some cases also by gamma-radiation. Depending on the reactor type, gamma radiation can influence evaluations of lead factors of surveillance specimens, effect the interpretation of results of irradiation experiments and finally, it can result in changed pressure vessel lifetime evaluations. The project aimed at the evaluation of the importance of gamma radiation for RPV steel damage for several types of light-water reactors. Absolute neutron and gamma fluence rate spectra had been calculated for the Russian PWR types VVER-440 and two core loading variants of VVER-1000, for a German 1300 MW PWR and a German 900 MW BWR. Based on the calculated spectra several flux integrals and radiation damage parameters were derived for the region of the azimuthal flux maxima in the mid-planes for different radial positions between core and biological shield, especially, at the inner and outer surfaces of the PV walls, at the (1/4)-PV-thickness and at the surveillance positions. Fissionable materials are often used as activation detectors for neutron fluence measurements. To get the real value the analysis demands to take into account the gamma induced fissions. Therefore, the part of these fissions in the total number of fissions was estimated for the detector reactions 237 Np(n,f) and 238 U(n,f) in the calculated neutron/gamma fields. It has been found that considerable corrections of the neutron fluence measurements can be necessary, especially in case of 238 U(n,f). Most of the calculations were performed using a three-dimensional synthesis of 2D/1D-flux distributions obtained by the S N -code DORT with the BUGLE-96T group cross-section library. (orig.) [de

  15. Development of key technologies in DPSSL system for fast-ignition, laser fusion reactor - FIREX, HALNA, and protection of final optics

    International Nuclear Information System (INIS)

    Norimatsu, T.; Azechi, H.; Fujimoto, Y.; Jitsuno, T.; Kanabe, T.; Kodama, R.; Kondo, K.; Miyanaga, N.; Nagatomo, H.; Nakatsuka, M.; Shiraga, H.; Tanaka, K.A.; Tsubakimoto, K.; Yamanaka, M.; Yasuhara, R.; Izawa, Y.; Kawashima, T.; Kurita, T.; Matsumoto, O.; Tsuchiya, Y.; Sekine, T.; Kan, H.

    2005-01-01

    A critical path to a laser fusion power plant is construction of a reliable, efficient, high repetitive energy driver including the relation with the reactor environment. At ILE, Osaka University, FIREX project has been proposed and the phase I to show heating of compressed fuel to 5 keV has started with construction of the FIREX laser. This project will demonstrate physics of fast ignition and elemental studies are carried out to obtain persuasive data to find the path to the goal. A diode-laser-pumped, solid-state-laser (DPSSL) HALNA-10 succeeded in operation of 7.5J output power at 10 Hz rep-rate. Contamination of final optics by metal vapor was studied using a 1/10 model of the beam duct. The result indicated that contamination can be controlled with high speed shutters and a low pressure buffer gas. (author)

  16. Electromagnetic pump

    International Nuclear Information System (INIS)

    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.)

  17. Large electro-magnetic pump design for application in the ASTRID sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Laffont, Guy; Rey, Frédéric; Aizawa, Rie; Suziki, Tetsu

    2013-01-01

    Conclusion: • Use of a LEMP motivated by several advantages in terms of the reactor design, operation and maintenance. • Collaboration agreement between the CEA and TOSHIBA Corporation came into force in April 2012 to carry out a joint work program on the ASTRID EMP design and development. • Preliminary LEMP calculations carried out by the CEA and TOSHIBA are in good agreement and provide a good confidence in the feasibility of the annular LEMP for the ASTRID intermediate sodium loop. • Theoretical and experimental investigations are currently underway at the CEA with the aim to improve the numerical tools. • In parallel, the ASTRID EMP conceptual design studies are ongoing at TOSHIBA (thermal and thermo-mechanical analyses to demonstrate the LEMP self-cooling, structural analysis of the casing, the supporting legs and the mechanical interfaces, definition of the power supply unit, instrumentation and remote control procedure). • This program is aiming at consolidating the ASTRID EMP conceptual design report and to support the design option choice for the ASTRID basic design

  18. Damage of reactor recirculation pump of No.3 plant in Fukushima No.2 Nuclear Power Station, Tokyo Electric Power Co., Inc

    International Nuclear Information System (INIS)

    1990-01-01

    No.3 plant in Fukushima No.2 Nuclear Power Station is a BWR plant with the rated output of 1100 MW, and as for the damage of its reactor recirculation pump, the investigation of the cause and the examination of countermeasures are advanced by the special committee. It is presumed that the submerged bearing ring of this pump caused the fatigue fracture due to the insufficient penetration in the fillet-welded part. The ring broke into pieces, and the main disk of the impeller was broken by the wear due to the pieces. Further, the damage of washers and the falling-off of bolts occurred. The metallic particles generated by wear were about 30 kg, and it is presumed that several kg of them adhered to fuel. Hereafter, the investigation of the cause will be continued, and the countermeasures for preventing the recurrence are examined. On January 1, 1989, the vibration of the pump increased, and the alarm was issued. However, the operation was continued carefully. On January 6, the vibration increased again, and on January 7, the reactor was stopped. The third regular inspection was begun on the same day, and the damage of the pump was found. (K.I.)

  19. Experimental fusion power reactor conceptual design study. Final report. Volume II

    International Nuclear Information System (INIS)

    Baker, C.C.

    1976-12-01

    This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following sections: (1) reactor components, (2) auxiliary systems, (3) operations, (4) facility design, (5) program considerations, and (6) conclusions and recommendations

  20. Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory

    International Nuclear Information System (INIS)

    Fellhauer, C.R.; Boing, L.E.; Aldana, J.

    1997-03-01

    The Final Report for the Decontamination and Decommissioning (D ampersand D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D ampersand D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D ampersand D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D ampersand D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a open-quotes Radiologically Controlled Area,close quotes noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion)

  1. Jet pumps hydrdynamics for application on BWRS

    International Nuclear Information System (INIS)

    Girardi, G.; Pitimada, D.

    1979-01-01

    An analysis of single-phase jet-pump hydrodynamics is carried out by this paper with special regard to the applications on cooling water recirculation in the boiling water reactors (BWR). Firstly, in order to asses on efficiency of jet pumps, several theories regarding the hydrodynamic of these machines are also investigated. The results of the above theories are critically analysed and compared regarding to water-jet-pump design, to operational performance curves and to section limits. Some general criteria in jet-pump design are introduced and values of geometric and kinematic parameters are suggested together with losses coefficients which are all concerned with the ''high ratio'' type jet pump of this typical application. Finally, the experimental test program following the sim of this research is briefly described

  2. 10-75-kWe-reactor-powered organic Rankine-cycle electric power systems (ORCEPS) study. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    1977-03-30

    This 10-75 kW(e) Reactor-ORCEPS study was concerned with the evaluation of several organic Rankine cycle energy conversion systems which utilized a /sup 235/U-ZrH reactor as a heat source. A liquid metal (NaK) loop employing a thermoelectric converter-powered EM pump was used to transfer the reactor energy to the organic working fluid. At moderate peak cycle temperatures (750/sup 0/F), power conversion unit cycle efficiencies of up to 25% and overall efficiencies of 20% can be obtained. The required operating life of seven years should be readily achievable. The CP-25 (toluene) working fluid cycle was found to provide the highest performance levels at the lowest system weights. Specific weights varies from 100 to 50 lb/kW(e) over the power level range 10 to 75 kW(e). (DLC)

  3. Pumping behavior of sputter ion pumps

    International Nuclear Information System (INIS)

    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

  4. Pumps in nuclear power plants

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.

  6. Development of CFD software for the simulation of thermal hydraulics in advanced nuclear reactors. Final report

    International Nuclear Information System (INIS)

    Bachar, Abdelaziz; Haslinger, Wolfgang; Scheuerer, Georg; Theodoridis, Georgios

    2015-01-01

    The objectives of the project were: Improvement of the simulation accuracy for nuclear reactor thermo-hydraulics by coupling system codes with three-dimensional CFD software; Extension of CFD software to predict thermo-hydraulics in advanced reactor concepts; Validation of the CFD software by simulation different UPTF TRAM-C test cases and development of best practice guidelines. The CFD module was based on the ANSYS CFD software and the system code ATHLET of GRS. All three objectives were met: The coupled ATHLET-ANSYS CFD software is in use at GRS and TU Muenchen. Besides the test cases described in the report, it has been used for other applications, for instance the TALL-3D experiment of KTH Stockholm. The CFD software was extended with material properties for liquid metals, and validated using existing data. Several new concepts were tested when applying the CFD software to the UPTF test cases: Simulations with Conjugate Heat Transfer (CHT) were performed for the first time. This led to better agreement between predictions and data and reduced uncertainties when applying temperature boundary conditions. The meshes for the CHT simulation were also used for a coupled fluid-structure-thermal analysis which was another novelty. The results of the multi-physics analysis showed plausible results for the mechanical and thermal stresses. The workflow developed as part of the current project can be directly used for industrial nuclear reactor simulations. Finally, simulations for two-phase flows with and without interfacial mass transfer were performed. These showed good agreement with data. However, a persisting problem for the simulation of multi-phase flows are the long simulation times which make use for industrial applications difficult.

  7. Feasibility study on commercialized fast reactor cycle systems. Phase II final report

    International Nuclear Information System (INIS)

    Ieda, Yoshiaki; Uchikawa, Sadao; Okubo, Tsutomu; Ono, Kiyoshi; Kato, Atsushi; Kurisaka, Kenichi; Sakamoto, Yoshihiko; Sato, Kazujiro; Sato, Koji; Chikazawa, Yoshitaka; Nakai, Ryodai; Nakabayashi, Hiroki; Nakamura, Hirofumi; Namekawa, Takashi; Niwa, Hajime; Nomura, Kazunori; Hayashi, Hideyuki; Hayafune, Hiroki; Hirao, Kazunori; Mizuno, Tomoyasu; Muramatsu, Toshiharu; Ando, Masato; Ono, Katsumi; Ogata, Takanari; Kubo, Shigenobu; Kotake, Shoji; Sagayama, Yutaka; Takakuma, Katsuyuki; Tanaka, Toshihiko; Namba, Takashi; Fujii, Sumio; Muramatsu, Kazuyoshi

    2006-06-01

    A joint project team of Japan Atomic Energy Agency and the Japan Atomic Power Company (as the representative of the electric utilities) started the feasibility study on commercialized fast reactor cycle systems (F/S) in July 1999 in cooperation with Central Research Institute of Electric Power Industry and vendors. On the major premise of safety assurance, F/S aims to present an appropriate picture of commercialization of fast reactor (FR) cycle system which has economic competitiveness with light water reactor cycle systems and other electricity base load systems, and to establish FR cycle technologies for the future major energy supply. In the period from Japanese fiscal year (JFY) 1999 to 2000, the phase-I of F/S was carried out to screen our representative FR, reprocessing and fuel fabrication technologies. In the phase-II (JFY 2001-2005), the design study of FR cycle concepts, the development of significant technologies necessary for the feasibility evaluation, and the confirmation of key technical issues were performed to clarify the promising candidate concepts toward the commercialization. In this final phase-II report clarified the most promising concept, the R and D plan until around 2015, and the key issues for the commercialization. Based on the comprehensive evaluation in F/S, the combination of the sodium-cooled FR with MOX fuel core, the advanced-aqueous reprocessing process and the simplified-pelletizing fuel fabrication process was recommended as the mainline choice for the most promising concept. The concept exceeds in technical advancement, and the conformity to the development targets was higher compared with that of the others. Alternative technologies are prepared to be decrease the development risk of innovative technologies in the mainline choice. (author)

  8. Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

    International Nuclear Information System (INIS)

    Okrent, D.

    1997-01-01

    This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, with the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident

  9. Research program in reactor core diagnostics with neutron noise methods: Stage 3. Final report

    International Nuclear Information System (INIS)

    Pazsit, I.; Garis, N.S.; Karlsson, J.; Racz, A.

    1997-09-01

    Stage 3 of the program has been executed 96-04-12. The long term goal is to develop noise methods for identification and localization of perturbations in reactor cores. The main parts of the program consist of modelling the noise source, calculation of the space- and frequency dependent transfer function, calculation of the neutron noise via a convolution of the transfer function of the system and the noise source, i.e. the perturbation, and finally finding an inversion or unfolding procedure to determine noise source parameters from the neutron noise. Most previous work is based on very simple (analytical) reactor models for the calculation of the transfer function as well as analytical unfolding methods. The purpose of this project is to calculate the transfer function in a more realistic model as well as elaborating powerful inversion methods that do not require analytical transfer functions. The work in stage 3 is described under the following headlines: Further investigation of simplified models for the calculation of the neutron noise; Further investigation of methods based on neural networks; Further investigation of methods for detecting the vibrations and impacting of detectors; Application of static codes for determination of the neutron noise using the adiabatic approximation

  10. Experimental fusion power reactor conceptual design study. Final report. Volume III

    International Nuclear Information System (INIS)

    Baker, C.C.

    1976-12-01

    This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following appendices: (1) tradeoff code analysis, (2) residual mode transport, (3) blanket/first wall design evaluations, (4) shielding design evaluation, (5) toroidal coil design evaluation, (6) E-coil design evaluation, (7) F-coil design evaluation, (8) plasma recycle system design evaluation, (9) primary coolant purification design evaluation, (10) power supply system design evaluation, (11) number of coolant loops, (12) power conversion system design evaluation, and (13) maintenance methods evaluation

  11. The final report of ''on-the-job training'' on the CANDU reactor

    International Nuclear Information System (INIS)

    Kim, D.H.; Koh, B.J.

    1983-01-01

    This is the final Report for the technical ''on-the-job traning'' for the Wolsung CANDU nuclear power plant which is the first Pressurized Heavy Water Reactor setting up in Korea. The technical ''on-the-job traning'' was established to increase the capability for the nuclear safety evaluation in order to contribute the future safe operation of the CANDU nuclear power plant. The training has been excuted through three level courses as elementary, intermediate and ''on-the-job training'' at Wolsung power plant. The elementary course was introduction to the CANDU basics and fundamentals. The intermediate course was the more advanced course, and the detailed concepts and engineering explanations of the CANDU system had been instructed. The third course was the ''on-the-job training'' at the Wolsung plant site, which was the most emphasized course during the project. (Author)

  12. The undersea location of the Swedish Final Repository for reactor waste, SFR - human intrusion aspects

    International Nuclear Information System (INIS)

    Eng, T.

    1989-01-01

    The Swedish Final Repository for reactor waste, SFR, is built under the Baltic sea close to the Forsmark nuclear power plant. Sixty metres of rock cover the repository caverns under the seabed. The depth of the Baltic sea is about 5-6 m at this location. A human intrusion scenario that in normal inland locations has shown to be of great importance, is a well that is drilled through or in the close vicinity of the repository. Since the land uplift in the SFR area is about 6 mm/year the undersea location of SFR ensures that no well will be drilled at this location for a considerable time while the area is covered by the Baltic sea

  13. Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

    International Nuclear Information System (INIS)

    1976-08-01

    This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

  14. Research Project 'RB research nuclear reactor' (operation and maintenance), Final report; Naucnoistrazivacki projekt 'Istrazivacki nuclearni reaktor RB, (pogon i odrzavanje), Zavrsni elaborat projekta

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    This final report covers operation and maintenance activities at the RB reactor during period from 1981-1985. First part covers the RB reactor operation, detailed description of reactor components, fuel, heavy water, reactor vessel, cooling system, equipment and instrumentation, auxiliary systems. It contains data concerned with dosimetry and radiation protection, reactor staff, and financial data. Second part deals maintenance, regular control and testing of reactor equipment and instrumentation. Third part is devoted to basic experimental options and utilization of the RB reactor including training.

  15. Development and validation of three-dimensional CFD techniques for reactor safety applications. Final report; Entwicklung und Validierung dreidimensionaler CFD Verfahren fuer Anwendungen in der Reaktorsicherheit. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Sebastian; Palazzo, Simone; Papukchiev, Angel; Scheurer Martina

    2016-12-15

    The overall goal of the project RS 1506 ''Development and Validation of Three Dimensional CFD Methods for Reactor Safety Applications'' is the validation of Computational Fluid Dynamics (CFD) software for the simulation of three -dimensional thermo-hydraulic heat and fluid flow phenomena in nuclear reactors. For this purpose a wide spectrum of validation and test cases was selected covering fluid flow and heat transfer phenomena in the downcomer and in the core of pressurized water reactors. In addition, the coupling of the system code ATHLET with the CFD code ANSYS CFX was further developed and validated. The first choice were UPTF experiments where turbulent single- and two-phase flows were investigated in a 1:1 scaled model of a German KONVOI reactor. The scope of the CFD calculations covers thermal mixing and stratification including condensation in single- and two-phase flows. In the complex core region, the flow in a fuel assembly with spacer grid was simulated as defined in the OECD/NEA Benchmark MATIS-H. Good agreement are achieved when the geometrical and physical boundary conditions were reproduced as realistic as possible. This includes, in particular, the consideration of heat transfer to walls. The influence of wall modelling on CFD results was investigated on the TALL-3D T01 experiment. In this case, the dynamic three dimensional fluid flow and heat transfer phenomena were simulated in a Generation IV liquid metal cooled reactor. Concurrently to the validation work, the coupling of the system code ATHLET with the ANSYS CFX software was optimized and expanded for two-phase flows. Different coupling approaches were investigated, in order to overcome the large difference between CPU-time requirements of system and CFD codes. Finally, the coupled simulation system was validated by applying it to the simulation of the PSI double T-junction experiment, the LBE-flow in the MYRRA Spallation experiment and a demonstration test case

  16. Strategies for reactor safety: Preventing loss of coolant accidents. Final report

    International Nuclear Information System (INIS)

    Lydell, B.O.Y.

    1997-12-01

    This final report on the NKS/RAK-1.2 summarizes the main features of the PIFRAP PC-program and its intended implementation. Regardless of the preferred technical approach to LOCA frequency estimation, the analysis approach must include recognition of the following technical issues: a) Degradation and failure mechanisms potentially affecting piping systems within the reactor coolant pressure boundary (RCPB) and the potential consequences; b) In-service inspection practices and how they influence piping reliability; and c) The service experience with piping systems. The report consists of six sections and one appendix. A Nordic perspective on LOCA and nuclear safety is given. It includes summaries of results from research in material sciences and current regulatory philosophies regarding piping reliability. A summary of the LOCA concept is applied in Nordic PSA studies. It includes a discussion on deterministic and probabilistic views on LOCA. The R and D on piping reliability by SKI and the PIFRAP model is summarized. Next, Section 6 presents conclusion and recommendations. Finally, Appendix A contains a list of abbreviations and acronyms, together with a glossary of technical terms. (EG)

  17. RA Reactor operation and maintenance (I-IX), part V, Task 3.08/04-06, Refurbishment of the heavy water pumps; Pogon i odrzavanje reaktora RA (I-IX), V Deo, Zadatak 3.08/04-06 Remont teskovodnih pumpi

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V; Nikolic, M; Milic, J [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    In addition to detailed instructions for maintenance and repair of the heavy water pumps at the RA reactor this document includes nine annexes. They are as follows: cleaning the heavy water pump Avala with distilled water; instructions for repair of the pump CEN-132 (two annexes); list of operating characteristics of the pumps before repair; conclusions of the experts concerning the worn out bearings of the heavy water pump Avala, with the analysis of the stellite layer; report on the completed repair actions on the pumps Avala and CEN-132; report on the measurements done on the pump Avala; and the certificate concerning inspection of the pump.

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

    International Nuclear Information System (INIS)

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

    1979-10-01

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

  19. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  20. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 2

    International Nuclear Information System (INIS)

    Singh, B.N.; Johansen, B.S.; Taehtinen, S.; Moilanen, P.; Saarela, S.; Jacquet, P.; Dekeyser, J.; Stubbins, J.F.

    2008-01-01

    The main objective of the present work was to determine experimentally the mechanical response and resulting microstructural changes in CuCrZr (HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Using specially designed test facilities for this purpose, in-reactor creep-fatigue tests have been performed at strain amplitudes of 0.25 and 0.35 % with a holdtime of 10s in the BR-2 reactor at Mol (Belgium). These tests were performed at the ambient temperatures of 326K and 323K. For comparison purposes corresponding out-of-reactor creep-fatigue tests were also carried out. In the following we first describe the details of the creep-fatigue experiments. We then present the main results on the mechanical response of the material in the form of hysteresis loops and the maximum stress amplitude as a function of the number of creep-fatigue cycles during the out-of-reactor and the in-reactor tests carried out at different strain amplitudes. Finally, the dependence of the number of cycles to failure (i.e. creep-fatigue lifetime) on the strain amplitudes is shown. The details of microstructure of the specimens tested out-of-reactor as well as in the reactor were investigated using transmission electron microscopy. The main results on the mechanical response as well as changes in the microstructure are briefly discussed. The main conclusion emerging from the present work is that the lifetime of the in-reactor tested specimens is by a factor of about two longer than in the case of corresponding out-of-reactor tests. (au)

  1. CLEAN-AIR heat pump. Reduced energy consumption for ventilation in buildings by integrating air cleaning and heat pump. Final Report; CLEAN-AIR heat pump - Reduceret energiforbrug til ventilation af bygninger ved luftrensning integreret med luft varmepumpe. Slut rapport

    Energy Technology Data Exchange (ETDEWEB)

    Fang, L.; Olesen, Bjarne W.; Molinaro, G.; Simmonsen, P.; Skocajic, S. [Danmarks Tekniske Univ. Institut for Byggeri og Anlaeg, Lyngby (Denmark); Hummelshoej, R.M.; Carlassara, L. [COWI A/S, Lyngby, (Denmark); Groenbaek, H.; Hansen, Ole R. [Exhausto A/S, Langeskov (Denmark)

    2011-07-01

    This report summarizes task 1 of the Clean Air Heat Pump project - modelling and simulation on energy savings when using the clean air heat pump for ventilation, air cleaning and energy recovery. The total energy consumption of the proposed ventilation systems using clean air heat pump technology was calculated by a theoretical model and compared with the reference ventilation systems (conventional ventilation systems). The energy compared between the two systems includes energy used for heating, cooling and fan. The simulation and energy saving calculation was made for the application of the clean air heat pump in three typical climate conditions, i.e. mild-cold, mild-hot and hot and wet climates. Real climate data recorded from three cities in 2002 was used for the calculation. The three cities were Copenhagen (Denmark), Milan (Italy) and Colombo (Sir Lanka) which represent the above three typical climate zones. For the Danish climate (the mild cold climate), the calculations show that the ventilation system using clean air heat pump technology can save up to 42% of energy cost in winter compared to the conventional ventilation system. The energy saving in summer can be as high as 66% for the ventilation system with humidity control and 9% for the ventilation system without the requirement of humidity control. Since the Danish summer climate is very mild, over 80% of the yearly energy consumption for ventilation is used during winter season. It is, therefore, estimated that more than 35% annual energy saving for ventilation is expected in Denmark using the clean air heat pump ventilation technology. For the mild hot climate, e.g. the Italian climate, the calculations show that up to 63% of the energy saving can be achieved in summer season. For the winter mode, 17% reduction of the energy cost can be expected for the domestic use. For industrial use, the energy cost of the clean air heat pump may not be favourable due to the industrial price of gas in Italy is

  2. Extension of Surveillance Test Interval of Safety Injection Pump for APR-1400 Reactors to Improve Reliability and Availability of the Pump

    Energy Technology Data Exchange (ETDEWEB)

    Osama, A. Rezk; Jung, J. C.; Lee, Yong-Kwan [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    The safety features function to localize, control, mitigate, and terminate such incidents and to hold exposure levels below applicable limits. The safety injection system is comprised of four independent mechanical trains without any tie line among the injection paths and two electrical divisions. Each train has one active Safety Injection Pump (SIP) and one passive Safety Injection Tank (SIT) equipped with a Fluidic Device (FD), each train provides 50% of the minimum injection flow rate for breaks larger than the size of a direct vessel injection line. For breaks equal to or smaller than the size of a direct vessel injection line, each train has 100% of the required capacity. The low pressure injection pumps with common header installed in the conventional design are eliminated, and the functions for safety injection and shutdown cooling are separated. The arrangement of safety injection system for APR-1400 as shown in figure (1). The results obtained in this work show that STI extensions for the SIS feasible without any unacceptable increase in the plant total risk, STI extensions are acceptable for safety injection system to provide plant operational flexibility in the performance of both corrective and preventive maintenance for the safety injection system.

  3. University Reactor Instrumentation Grant. Final report 08/06/1998 - 08/13/1999

    International Nuclear Information System (INIS)

    Bajorek, S. M.

    2000-01-01

    A noble gas air monitoring system was purchased through the University Reactor Instrumentation Grant Program. This monitor was installed in the Kansas State TRIGA reactor bay at a location near the top surface of the reactor pool according to recommendation by the supplier. This system is now functional and has been incorporated into the facility license

  4. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

    2004-08-01

    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  5. 75 FR 69709 - Office of New Reactors; Notice of Availability of the Final Staff Guidance; Standard Review Plan...

    Science.gov (United States)

    2010-11-15

    ... the Final Staff Guidance; Standard Review Plan, Section 13.6.6, Revision 0 on Cyber Security Plan... Reports for Nuclear Power Plants,'' Section 13.6.6, Revision 0 on ``Cyber Security Plan'' (Agencywide... reviews to amendments to licenses for operating reactors or for activities associated with review of...

  6. University Reactor Sharing Program. Final report, September 30, 1992--September 29, 1994

    International Nuclear Information System (INIS)

    Wehring, B.W.

    1995-01-01

    Over the past 20 years, the number of nuclear reactors on university campuses in the US declined from more than 70 to less than 40. Contrary to this trend, The University of Texas at Austin constructed a new reactor facility at a cost of $5.8 million. The new reactor facility houses a new TRIGA Mark II reactor which replaces an in-ground TRIGA Mark I reactor located in a 50-year old building. The new reactor facility was constructed to strengthen the instruction and research opportunities in nuclear science and engineering for both undergraduate and graduate students at The University of Texas. On January 17, 1992, The University of Texas at Austin received a license for operation of the new reactor. Initial criticality was achieved on March 12, 1992, and full power operation, on March 25, 1992. The UT-TRIGA research reactor provides hands-on education, multidisciplinary research and unique service activities for academic, medical, industrial, and government groups. Support by the University Reactor Sharing Programs increases the availability of The University of Texas reactor facility for use by other educational institutions which do not have nuclear reactors

  7. A feasibility study of a linear laser heated solenoid fusion reactor. Final report

    International Nuclear Information System (INIS)

    Steinhauer, L.C.

    1976-02-01

    This report examines the feasibility of a laser heated solenoid as a fusion or fusion-fission reactor system. The objective of this study, was an assessment of the laser heated solenoid reactor concept in terms of its plasma physics, engineering design, and commercial feasibility. Within the study many pertinent reactor aspects were treated including: physics of the laser-plasma interaction; thermonuclear behavior of a slender plasma column; end-losses under reactor conditions; design of a modular first wall, a hybrid (both superconducting and normal) magnet, a large CO 2 laser system; reactor blanket; electrical storage elements; neutronics; radiation damage, and tritium processing. Self-consistent reactor configurations were developed for both pure fusion and fusion-fission designs, with the latter designed both to produce power and/or fissile fuels for conventional fission reactors. Appendix A is a bibliography with commentary of theoretical and experimental studies that have been directed at the laser heated solenoid

  8. Characterization and isolation of a light driven sodium pump from membranes of Halobacterium halobium. Final technical progress report

    International Nuclear Information System (INIS)

    MacDonald, R.E.

    1982-01-01

    We investigated three aspects of the light driven sodium pump (halorhodopsin, which appear to be crucial to our understanding of the mechanisms employed by Halobacterium halobium and to further investigate this unique system of energy conservation. We characterized the molecular mechanisms of transmembrane sodium transport in vesicles from H. halobium with particular reference to the mechanism of couplins of light energy to net sodium translocation. We develop procedures and techniques for extracting the components of the light driven sodium pump from membranes and incorporating them into artificial membrane systems. We examine the mechanism of conversion of bacteriorhodopsin from an active to an inactive form in membrane vesicles and to relate this alternative state of this pigment to the presence of the light driven sodium pump

  9. Municipal water-based heat pump heating and/or cooling systems: Findings and recommendations. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bloomquist, R.G. [Washington, State Univ., Pullman, WA (United States); Wegman, S. [South Dakota Utilities Commission (United States)

    1998-04-01

    The purpose of the present work was to determine if existing heat pump systems based on municipal water systems meet existing water quality standards, to analyze water that has passed through a heat pump or heat exchanger to determine if corrosion products can be detected, to determine residual chlorine levels in municipal waters on the inlet as well as the outlet side of such installations, to analyses for bacterial contaminants and/or regrowth due to the presence of a heat pump or heat exchanger, to develop and suggest criteria for system design and construction, to provide recommendations and specifications for material and fluid selection, and to develop model rules and regulations for the installation, operation, and monitoring of new and existing systems. In addition, the Washington State University (WSU) has evaluated availability of computer models that would allow for water system mapping, water quality modeling and system operation.

  10. Accident analysis for new reactor concepts and VVER type reactor design with advanced fuel. STC with Russia. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.

    2000-10-01

    In the frame of a project on scientific-technical cooperation funded by BMBF/BMWi, the 3D reactor dynamics code DYN3D developed at Forschungszentrum Rossendorf (FZR), has been transferred to the Institute of Physics and Power Engineering (IPPE) Obninsk in Russia and integrated into the software package of IPPE. DYN3D has been coupled to a thermohydraulic system code used in IPPE making available 3D neutron kinetics within this software package. A new macroscopic cross section library has been created using a modified version of the WIMS/D4 code. This library includes data for modernized fuel design containing burnable absorbers in different concentrations, which is tested in VVER-1000 type reactors. The cross section library has been connected to DYN3D. Calculations were performed to check the library in comparison with other data libraries and codes. The code DYN3D and the coupled 3D neutron kinetics/thermal hydraulics code system were used to perform analyses of Anticipated Transients Without Scram (ATWS) for the reactor design ABV-67, an integral reactor concept with small power developed under participation of IPPE. The fluid dynamics code DINCOR developed at IPPE was transferred to FZR. It was used in validation calculations on test problems for the short-term core melt behaviour (CORVIS experiments). (orig.) [de

  11. Pumps for cryogenic liquids with superconducting magnetic bearings. Final report; Pumpen fuer kryogene Fluessigkeiten mit supraleitenden Magnetlagern. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Gerlach, G.; Fuchs, G.; Sorber, J.; Brosche, H.; Richter, M.; Frenzel, C.

    2000-07-01

    A liquid nitrogen pump with contactless superconducting magnetic bearings was to be developed on the basis of an available motor with superconducting bearings. Contactless superconducting magnetic bearings require practically no servicing. A high demand for pumps for cryogenic liquids is expected with the impending use of hydrogen as an energy source. The pumping of liquid nitrogen was demonstrated successfully with the new test aggregate. The maximum pumped volume was 17 l/min at a lift of 0.5 m and 6 l/min at a lift of 1 m. In all, 15 hours of operation were registered in the superconducting state of the bearing, which included 2 hours of uninterrupted pump operation. The higher speed range for which magnetic bearings are optimally suited was not reached. Operation at higher frequencies was impossible either because of stronger resonance amplituees or because the power system was too weak. [German] Ziel des Vorhabens war die Entwicklung einer Pumpe fuer fluessigen Stickstoff mit beruehrungslosen supraleitenden Magnetlagern auf der Basis eines vorhandenen supraleitend gelagerten Motors. Die beruehrungslose supraleitende Magnetlager sind praktisch wartungsfrei. Ein Bedarf an Pumpen fuer kryogene Fluessigkeiten entsteht insbesondere durch den in naher Zukunft zu erwartenden Einsatz von Wasserstoff als Energietraeger. Mit dem entworfenen Aggregat wurde das Pumpen von Fluessigstickstoff erfolgreich demonstriert. Der Foerderstrom betrug bei 0,5m Foerderhoehe maximal 17 l/min; beim 1m Foerderhoehe wurden maximal 6 l/min gemessen. Es wurden insgesamt ca. 15 Betriebsstunden in supraleitenden Zustand des Lagers, darunter 2 Stunden ununterbrochener Pumpbetrieb registriert. Der hoehere Drehzahlbereich, fuer den das Magnetlager eigentlich paedestiniert ist, konnte nicht erreicht werden. Ein Betrieb bei hoeheren (Ist-)Frequenzen war nicht moeglich, entweder durch staerkere Resonanzausschlaege oder durch einen zu schwachen Antrieb. (orig.)

  12. Simulation of heat-pump systems in Polysun 4 - Final report; Simulation von Waermepumpen-Systemen in Polysun 4 - Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Marti, J.; Witzig, A. [Vela Solaris AG, Winterthur (Switzerland); Huber, A.; Ochs, M. [Huber Energietechnik AG, Zuerich (Switzerland)

    2009-01-15

    Polysun 4 is a software program for the simulation of heating systems. The simulation kernel applies a time stepping algorithm and dynamically calculates all relevant system parameters over a one year period, based on statistical weather data. On the one hand, Polysun draws out by physics-based simulation scheme and its modularity, which allows any arrangement of the system components. On the other hand, Polysun offers a unique set of component catalogues which cover a large number of commercially available system components. In this project, three kinds of heat pumps have been integrated in Polysun, namely the air/water, water/water and brine/water heat pumps. Furthermore, the relevant heat sources have been implemented, namely ambient air, soil and groundwater. In consequence, Polysun now covers a large, and almost complete, range of renewable energy systems. Simulation parameters are the measured heat pump COP values (in accordance with EN 255 and EN 14511). A linear interpolation scheme has been developed in this project in order to simulate systems for arbitrary source and heat pump temperatures and to interpolate the power consumption. For the dynamic simulation of the ground source heat pump, the numerical algorithm from the Program EWS (calculation module developed in 1997) has been integrated into Polysun. Groundwater probes are calculated with respect to the soil temperatures. Heat pumps and probes were implemented as independent components in Polysun. In the graphical user interface, they can be arbitrarily placed and connected with other hydraulic components. The timestepping simulation calculates inlet temperature, electric power consumption and heat transfer in the entire system. The Polysun catalogs have been extended accordingly with total over 300 component entries and a number of relevant system templates. (authors)

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

    International Nuclear Information System (INIS)

    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)

  14. Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

    International Nuclear Information System (INIS)

    Temple, S.M.; Robbins, T.R.

    1986-09-01

    This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR

  15. Resorption heat pump

    International Nuclear Information System (INIS)

    Vasiliev, L.L.; Mishkinis, D.A.; Antukh, A.A.; Kulakov, A.G.; Vasiliev, L.L.

    2004-01-01

    Resorption processes are based on at least two solid-sorption reactors application. The most favorable situation for the resorption heat pumps is the case, when the presence of a liquid phase is impossible. From simple case--two reactors with two salts to complicated system with two salts + active carbon fiber (fabric) and two branch of the heat pump acting out of phase to produce heat and cold simultaneously, this is the topic of this research program

  16. DOE University Reactor Sharing Program. Final technical report for 1996--1997

    International Nuclear Information System (INIS)

    Chappas, W.J.; Adams, V.G.

    1998-01-01

    The Department of Energy University Reactor Sharing Program at University of Maryland, College Park (UMCP) has, once again, stimulated a broad use of the reactor and radiation facilities by undergraduate and graduate students, visitors, and professionals. Participants are exposed to topics such as nuclear engineering, radiation safety, and nuclear reactor operations. This information is presented through various means including tours, slide presentations, experiments, and discussions. Student research using the MUTR is also encouraged. In addition, the Reactor Sharing Program here at the University of Maryland does not limit itself to the confines of the TRIGA reactor facility. Incorporated in the program are the Maryland University Radiation Effects Laboratory, and the UMCP 2 x 4 Thermal Hydraulic Loop. These facilities enhance and give an added dimension to the tours and experiments. The Maryland University Training Reactor (MUTR) and the associated laboratories are made available to any interested institution six days a week on a scheduled basis. Most institutions are scheduled at the time of their first request--a reflection of their commitment to the Reactor Sharing Program. The success of the past years by no means guarantees future success. Therefore, the reactor staff is more aggressively pursuing its outreach program, especially with junior colleges and universities without reactor or radiation facilities; more aggressively developing demonstration and training programs for students interested in careers in nuclear power and radiation technology; and more aggressively up-grading the reactor facilities--not only to provide a better training facility but to prepare for relicensing in the year 2000

  17. Experimental results of acetone hydrogenation on a heat exchanger type reactor for solar chemical heat pump; Solar chemical heat pump ni okeru acetone suisoka hanno netsu kaishu jikken

    Energy Technology Data Exchange (ETDEWEB)

    Takashima, T; Doi, T; Tanaka, T; Ando, Y [Electrotechnical Laboratory, Tsukuba (Japan); Miyahara, R; Kamoshida, J [Shibaura Institute of Technology, Tokyo (Japan)

    1996-10-27

    With the purpose of converting solar heat energy to industrial heat energy, an experiment of acetone hydrogenation was carried out using a heat exchanger type reactor that recovers heat generated by acetone hydrogenation, an exothermic reaction, and supplies it to an outside load. In the experiment, a pellet-like activated carbon-supported ruthenium catalyst was used for the acetone hydrogenation with hydrogen and acetone supplied to the catalyst layer at a space velocity of 400-1,200 or so. In the external pipe of the double-pipe type reactor, a heating medium oil was circulated in parallel with the flow of the reactant, with the heat of reaction recovered that was generated from the acetone hydrogenation. In this experiment, an 1wt%Ru/C catalyst and a 5wt%Ru/C catalyst were used so as to examine the effects of variation in the space velocity. As a result, from the viewpoint of recovering the heat of reaction, it was found desirable to increase the reaction speed by raising catalytic density and also to supply the reactant downstream inside the reaction pipe by increasing the space velocity. 1 ref., 6 figs., 1 tab.

  18. Experimental results of 2-propanol dehydrogenation with a falling-liquid film reactor for solar chemical heat pump; Solar chemical heat pump ni okeru ryuka ekimakushiki 2-propanol bunkai hanno jikken

    Energy Technology Data Exchange (ETDEWEB)

    Doi, T; Tanaka, T; Ando, Y; Takashima, T [Electrotechnical Laboratory, Tsukuba (Japan); Koike, M; Kamoshida, J [Shibaura Institute of Technology, Tokyo (Japan)

    1997-11-25

    A solar chemical heat pump is intended to attempt multi-purposed effective utilization of solar energy by raising low temperature solar heat of about 100 degC to 150 to 200 degC by utilizing chemical reactions. The chemical heat pump under the present study uses a 2-propanol (IPA)/acetone/hydrogen system which can utilize low-temperature solar heat and has large temperature rising degree. It was found from the result of experiments and analyses that IPA dehydrogenation reaction can improve more largely the heat utilization rate in using a falling-liquid film reactor than using a liquid phase suspended system. As an attempt to improve further the heat utilization rate, this paper reports the result of experimental discussions on inclination angles of a reaction vessel and feed liquid flow rate which would affect the fluid condition of the liquid film. As a result of the experiments, the initial deterioration in the catalyst has settled in about 15 hours, and its activity has decreased to about 60% of the initial activity. It was made clear that the influence of the inclination angle of the reaction vessel on the reaction is small. 5 refs., 7 figs.

  19. Job task and functional analysis of the Division of Reactor Projects, office of Nuclear Reactor Regulation. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Morzinski, J.A.; Gilmore, W.; Hahn, H.A.

    1998-07-10

    A job task and functional analysis was recently completed for the positions that make up the regional Divisions of Reactor Projects. Among the conclusions of that analysis was a recommendation to clarify roles and responsibilities among site, regional, and headquarters personnel. As that analysis did not cover headquarters personnel, a similar analysis was undertaken of three headquarters positions within the Division of Reactor Projects: Licensing Assistants, Project Managers, and Project Directors. The goals of this analysis were to systematically evaluate the tasks performed by these headquarters personnel to determine job training requirements, to account for variations due to division/regional assignment or differences in several experience categories, and to determine how, and by which positions, certain functions are best performed. The results of this analysis include recommendations for training and for job design. Data to support this analysis was collected by a survey instrument and through several sets of focus group meetings with representatives from each position.

  20. Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

    International Nuclear Information System (INIS)

    Vierow, Karen; Aldemir, Tunc

    2009-01-01

    The project entitled, 'Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors', was conducted as a DOE NERI project collaboration between Texas A and M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

  1. Final report. U.S. Department of Energy University Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, John A

    2003-01-21

    Activities supported at the MIT Nuclear Reactor Laboratory under the U.S. DOE University Reactor Sharing Program are reported for Grant DE FG02-95NE38121 (September 16, 1995 through May 31, 2002). These activities fell under four subcategories: support for research at thesis and post-doctoral levels, support for college-level laboratory exercises, support for reactor tours/lectures on nuclear energy, and support for science fair participants.

  2. Studies in fusion reactor technology. Final report, September 1, 1974--August 31, 1977

    International Nuclear Information System (INIS)

    Axtmann, R.C.; Perkins, H.K.

    1977-08-01

    Two independent measurements of hydrogen permeation through stainless steel at driving pressures in the range from 10 -6 to 1 Pa indicate that most extant predictions of tritium permeation through fusion reactors are probably overestimated grossly. A comprehensive analysis demonstrates that, given available structural materials, the prospects are negligible for the economic production of synthetic fuels via radiolytic reactions in fusion reactor systems

  3. Studies on the properties of hard-spectrum, actinide fissioning reactors. Final report

    International Nuclear Information System (INIS)

    Nelson, J.B.; Prichard, A.W.; Schofield, P.E.; Robinson, A.H.; Spinrad, B.I.

    1980-01-01

    It is technically feasible to construct an operable (e.g., safe and stable) reactor to burn waste actinides rapidly. The heart of the concept is a driver core of EBR-II type, with a central radial target zone in which fuel elements, made entirely of waste actinides are exposed. This target fuel undergoes fission, as a result of which actinides are rapidly destroyed. Although the same result could be achieved in more conventionally designed LWR or LMFBR systems, the fast spectrum reactor does a much more efficient job, by virtue of the fact that in both LWR and LMFBR reactors, actinide fission is preceded by several captures before a fissile nuclide is formed. In the fast spectrum reactor that is called ABR (actinide burning reactor), these neutron captures are short-circuited

  4. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  5. Decontamination and decommissioning of the SPERT-I Reactor Building at the Idaho National Engineering Laboratory. Final report

    International Nuclear Information System (INIS)

    Dolenc, M.R.

    1986-02-01

    This final report documents the decontamination and decommissioning of the SPERT-I Reactor Building. This 20- by 40-ft galvanized steel building was dismantled; and the resultant contaminated sludge, liquid, and carbon steel were disposed of at the Radioactive Waste Management Complex of the Idaho National Engineering Laboratory. This report presents the results of the characterization, decision analysis, planning, and decommissioning of the facility. The total cost of these activities was $139,500. Of this total, $103,500 was required for decommissioning operations. (This latter figure represents a 20% savings over the estimated costs generated during the planning effort.) The objectives of decommissioning this facility were to stabilize the seepage pit area and remove the reactor building. The D and D work was divided into two parts; the seepage pit was decommissioned in 1984, and the reactor building in 1985. The entire area was backfilled with radiologically clean soil, graded, and seeded. Two markers were installed to identify the locations of the pit and reactor building. The only isotopes found in either decommissioning operation were cesium-137 and uranium-235 in very low concentrations. Decommissioning operations of the reactor building were carried out during August 1985. The project generate 297 ft 3 of radioactive waste. No personnel radiation exposure above background was received by D and D workers

  6. Auditable Safety Analysis and Final Hazard Classification for the 105-N Reactor Zone and 109-N Steam Generator Zone Facility

    International Nuclear Information System (INIS)

    Kloster, G.L.

    1998-07-01

    This document is a graded auditable safety analysis (ASA) and final hazard classification (FHC) for the Reactor/Steam Generator Zone Segment. The Reactor/Steam Generator Zone Segment, part of the N Reactor Complex, that is also known as the Reactor Building and Steam Generator Cells. The installation of the modifications described within to support surveillance and maintenance activities are to be completed by July 1, 1999. The surveillance and maintenance activities addressed within are assumed to continue for the next 15- 20 years, until the initiation of facility D ampersand D (i.e., Interim Safe Storage). The graded ASA in this document is in accordance with EDPI-4.30-01, Rev. 1, Safety Analysis Documentation, (BHI-DE-1) and is consistent with guidance provided by the U.S. Department of Energy. This ASA describes the hazards within the facility and evaluates the adequacy of the measures taken to reduce, control, or mitigate the identified hazards. This document also serves as the FHC for the Reactor/Steam Generator Zone Segment. This FHC is developed through the use of bounding accident analyses that envelope the potential exposures to personnel

  7. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  8. Outline of a method for final storage of low- and medium-active waste from possible Danish power reactors

    International Nuclear Information System (INIS)

    Brodersen, K.; Jensen, J.; Oestergaard, K.

    1977-02-01

    A method is outlined for the final storage of Danish low-and medium-active power reactor waste. The waste drums are contained in large concretre blocks placed just below the ground surface. A plant for storing waste by means of this method is sketched. It consists of a system of reinforced concrete pits with the top level with the ground surface. Each pit measures c. 5 x 5 m and is c. 6 m deep. The pits are envisaged cast with a permanent inside, step-like shuttering of thin steel plates. The volume between the drums will be cast with concrete when a pit is filled. Calculations are given of the construction and running costs. It is estimated that the final storage of reactor wastes is only a small problem regarding economy and space, and also that there is hardly doubt that full safety can be achieved. (B.P.)

  9. Electromagnetic pump technology

    International Nuclear Information System (INIS)

    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

  10. Fusion reactor technology studies. Final report for period August 1, 1972 - October 31, 1978

    International Nuclear Information System (INIS)

    Kulcinski, G.L.; Maynard, C.W.

    1984-04-01

    Major accomplishments for the period August 1, 1972 - October 31, 1978 include the publishing of four comprehensive fusion reactor conceptual design studies; experimental studies in the areas of radiation damage, plasma-wall interactions, superconducting magnets and 14-MeV neutron cross sections; development of the concepts of carbon curtains and ISSEC's for use in fusion reactors; development of a neutron and gamma heating computer code, a radioactivity and afterheat computer code and a neutral transport computer code; and studies in the areas of RF heating for tokamaks and resource assessment for fusion reactors

  11. Some safety studies for conceptual fusion--fission hybrid reactors. Final report

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Okrent, D.

    1978-07-01

    The objective of this study was to make a preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors. The study and subsequent analysis was largely based upon reference to one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The blanket is a fast-spectrum, uranium carbide, helium cooled, subcritical reactor, optimized for the production of fissile fuel. An attempt was made to generalize the results wherever possible

  12. Design study of a fusion-driven tokamak hybrid reactor for fissile fuel production. Final report

    International Nuclear Information System (INIS)

    Rose, R.P.

    1979-05-01

    This study evaluated conceptual approaches for a tokamak fusion-driven fuel producing reactor. The conceptual design of this hybrid reactor was based on using projected state-of-the-art technology for the late 1980s. This reactor would be a demonstration plant and, therefore, first-of-a-kind considerations have been included. The conceptual definitions of two alternatives for the fusion driver were evaluated. A Two-Component Tokamak (TCT) concept, based on the TFTR plasma physics parameters, was compared to a Beam-Driven Thermonuclear (BDTN) concept, based on the USSR T-20 plasma physics parameters

  13. Some safety considerations in laser-controlled thermonuclear reactors. Final report

    International Nuclear Information System (INIS)

    Botts, T.E.; Breton, D.; Chan, C.K.; Levy, S.I.; Sehnert, M.; Ullman, A.Z.

    1978-07-01

    A major objective of this study was to identify potential safety questions for laser controlled thermonuclear reactors. From the safety viewpoint, it does not appear that the actual laser controlled thermonuclear reactor conceptual designs present hazards very different than those of magnetically confined fusion reactors. Some aspects seem beneficial, such as small lithium inventories, and the absence of cryogenic devices, while other aspects are new, for example the explosion of pressure vessels and laser hazards themselves. Major aspects considered in this report include: (a) general safety considerations, (b) tritium inventories, (c) system behavior during loss of flow accidents, and (d) safety considerations of laser related penetrations

  14. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    International Nuclear Information System (INIS)

    McDeavitt, Sean M.

    2011-01-01

    Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500 C to 600 C) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: (1) Hot working fabrication using mechanical alloying and extrusion - Design, fabricate, and assemble extrusion equipment - Extrusion database on DU metal - Extrusion database on U-10Zr alloys - Extrusion database on U-20xx-10Zr alloys - Evaluation and testing of tube sheath metals (2) Low-temperature sintering of U alloys - Design, fabricate, and assemble equipment - Sintering database on DU metal - Sintering database on U-10Zr alloys - Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research and Development (FCR and D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the

  15. Pump safety device

    International Nuclear Information System (INIS)

    Timmermans, Francis; Vandervorst, Jean.

    1981-01-01

    Safety device for longitudinally leak proofing the shaft of a pump in the event of the fracture of the dynamic seal separating the pump fluid high pressure chamber from the low pressure chamber. It is designed for fitting to the primary pumps of nuclear reactors. It includes a hollow cyclindrical piston located coaxially around the pump shaft and normally housed in a chamber provided for this purpose in the fixed housing of the dynamic seal, and means for moving this piston coaxially so as to compress a safety O ring between the shaft and the piston in the event of the dynamic seal failing [fr

  16. Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

    International Nuclear Information System (INIS)

    Wegst, Ulrike G.K.; Sridharan, Kumar

    2014-01-01

    Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

  17. Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Wegst, Ulrike G.K. [Dartmouth College, Hanover, NH (United States). Thayer School of Engineering; Allen, Todd [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States)

    2014-04-07

    Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

  18. Operability design review of prototype large breeder reactor (PLBR) designs. Final report, September 1981

    International Nuclear Information System (INIS)

    Beakes, J.H.; Ehman, J.R.; Jones, H.M.; Kinne, B.V.T.; Price, C.M.; Shores, S.P.; Welch, J.K.

    1981-09-01

    Prototype Large Breeder Reactor (PLBR) designs were reviewed by personnel with extensive power plant operations experience. Fourteen normal and off-normal events, such as startup, shutdown, refueling, reactor scram and loss of feedwater, were evaluated using an operational evaluation methodology which is designed to facilitate talk-through sessions on operational events. Human factors engineers participated in the review and assisted in developing and refining the review methodologies. Operating experience at breeder reactor facilities such as Experimental Breeder Reactor-II (EBR-II), Enrico Fermi Atomic Power Plant - Unit 1, and the Fast Flux Test Facility (FFTF) was gathered, analyzed, and used to determine whether lessons learned from operational experience had been incorporated into the PLBR designs. This eighteen month effort resulted in approximately one hundred specific recommendations for improving the operability of PLBR designs

  19. NCSU PULSTAR reactor instrumentation upgrade. Final technical report, September 6, 1990--March 19, 1993

    International Nuclear Information System (INIS)

    Bilyj, S.J.; Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University initiated an upgrade program at the NCSU PULSTAR Reactor in 1990. Twenty-year-old instrumentation is currently undergoing replacement with solid-state and current technology equipment. The financial assistance from the United States Department of Energy has been the primary source of support. This report provides the status of the first two phases of the upgrade program

  20. Sharing of the RPI Reactor Critical Facility (RCF). Final summary report, January 1988--September 1995

    International Nuclear Information System (INIS)

    Harris, D.R.

    1995-01-01

    Rensselaer Polytechnic Institute (RPI) has participated for a number of years in Sharing of the Reactor Critical Facility (RCF) under the U.S. Department of Energy University Reactor Sharing Program. In September of each year a Sharing invitation is sent to 92 public and private high schools and to 74 colleges and universities within about a 3 hour drive to the RCF (Appendix B). Each year about 10 such educational institutions send groups to share the RCF

  1. Study of the modifications on the synchronous generators, heavy water pumps and condenser batteries of the RA reactor - Annex 17; Prilog 17 - Elaborat o izmenama u semama sinhronih generatora, teskovodnih pumpi i kondenzatorskih baterija reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    Modifications done on the synchronous generators are related to the emergency power supply system, meaning one of the most important devices responsible for reactor safety. Without reducing the efficiency of the heavy water pumps the improved stability of generators operation was achieved by reducing the possibility of errors and simplifying manipulation. Condensator batteries were improved in order to decrease the leakage currents.

  2. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    International Nuclear Information System (INIS)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

  3. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    International Nuclear Information System (INIS)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

  4. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

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

    International Nuclear Information System (INIS)

    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

  6. Generic component reliability data for research reactors PSA. Final report of the CRP on data acquisition for research reactor PSA. Working material

    International Nuclear Information System (INIS)

    1993-01-01

    The scope of this document is to provide the final reference generic component reliability database information for a variety of research reactor types. As noted in Section 2.1 and Table 3a, many years of component data are represented in the database so that it is expected that the report should provide representative data valid for a number of years. The database provides component failure rates on a time and/or a demand related basis according to the operational modes of the components. At the current time an update of the database is not planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available from the contributing research reactor facilities themselves. As noted in Section 1.2, the report does not include detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussion regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, tabs

  7. Generic component reliability data for research reactors PSA. Final report of the CRP on data acquisition for research reactor PSA. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    The scope of this document is to provide the final reference generic component reliability database information for a variety of research reactor types. As noted in Section 2.1 and Table 3a, many years of component data are represented in the database so that it is expected that the report should provide representative data valid for a number of years. The database provides component failure rates on a time and/or a demand related basis according to the operational modes of the components. At the current time an update of the database is not planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available from the contributing research reactor facilities themselves. As noted in Section 1.2, the report does not include detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussion regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, tabs.

  8. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean M

    2011-04-29

    Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºC to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich

  9. Environmentally-assisted cracking in austenitic light water reactor structural materials. Final report of the KORA-I project

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.-P.; Ritter, S

    2009-03-15

    The following document is the final report of the KORA-I project, which was performed at the Paul Scherrer Institute (PSI) between 2006 and 2008 and was funded by the Swiss Nuclear Safety Inspectorate (ENSI). The three sub-projects of KORA-I covered the experimental characterisation of the effect of the reactor coolant environment on fatigue initiation and crack growth in austenitic stainless steels under boiling and pressurised water reactor conditions, the experimental evaluation of the potential and limits of the electrochemical noise measurement technique for the early detection of stress corrosion cracking initiation in austenitic stainless steels under boiling water reactor/normal water chemistry conditions, as well as the characterisation of the stress corrosion crack growth behaviour in the fusion line region of an Alloy 182-low-alloy reactor pressure vessel steel dissimilar metal weld. The main scientific results and major conclusions of the three sub-projects are discussed in three independent parts of this report. (author)

  10. Environmentally-assisted cracking in austenitic light water reactor structural materials. Final report of the KORA-I project

    International Nuclear Information System (INIS)

    Seifert, H.-P.; Ritter, S.

    2009-03-01

    The following document is the final report of the KORA-I project, which was performed at the Paul Scherrer Institute (PSI) between 2006 and 2008 and was funded by the Swiss Nuclear Safety Inspectorate (ENSI). The three sub-projects of KORA-I covered the experimental characterisation of the effect of the reactor coolant environment on fatigue initiation and crack growth in austenitic stainless steels under boiling and pressurised water reactor conditions, the experimental evaluation of the potential and limits of the electrochemical noise measurement technique for the early detection of stress corrosion cracking initiation in austenitic stainless steels under boiling water reactor/normal water chemistry conditions, as well as the characterisation of the stress corrosion crack growth behaviour in the fusion line region of an Alloy 182-low-alloy reactor pressure vessel steel dissimilar metal weld. The main scientific results and major conclusions of the three sub-projects are discussed in three independent parts of this report. (author)

  11. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Y. J.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2008-05-15

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes.

  12. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Yoon, C.; Rhee, B. W.; Chung, B. D.; Cho, Y. J.; Kim, M. W.

    2008-01-01

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes

  13. The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lancaster, D.

    1992-10-26

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

  14. University Reactor Instrumentation grant program. Final report, September 7, 1990--August 31, 1995

    International Nuclear Information System (INIS)

    Talnagi, J.W.

    1998-01-01

    The Ohio State University Nuclear Reactor Laboratory (OSU NRL) participated in the Department of Energy (DOE) grant program commonly denoted as the University Reactor Instrumentation (URI) program from the period September 1990 through August 1995, after which funding was terminated on a programmatic basis by DOE. This program provided funding support for acquisition of capital equipment targeted for facility upgrades and improvements, including modernizing reactor systems and instrumentation, improvements in research and instructional capabilities, and infrastructure enhancements. The staff of the OSU NRL submitted five grant applications during this period, all of which were funded either partially or in their entirety. This report will provide an overview of the activities carried out under these grants and assess their impact on the OSU NRL facilities

  15. Limiting factor analysis of high availability nuclear plants (boiling water reactors). Final report

    International Nuclear Information System (INIS)

    Frederick, L.G.; Brady, R.M.; Shor, S.W.W.; McCusker, J.T.; Alden, W.M.; Kovacs, S.

    1979-08-01

    The pertinent results are presented of a 16-month study conducted for Electric Power Research Institute by General Electric Company, Bechtel Power Corporation, and Philadelphia Electric Company. The study centered around the Peach Bottom 2 Atomic Power Station, but also included limited study of operations at 20 additional operating boiling water reactors. The purpose of the study was to identify and evaluate key factors limiting plant availability, and to identify potential improvements for eliminating or alleviating those limitations. The key limiting factors were found to be refueling activities; activities related to the reactor fuel; reactor scrams; activities related to 20 operating systems or major components; delays due to radiation, turbid water during refueling operations, facilities/working conditions, and dirt/foreign material; and general maintenance/repair of valves and piping. Existing programs to reduce the effect on plant unavailability are identified, and suggestions for further action are made

  16. Indigenously developed large pumping speed cryoadsorption cryopump

    International Nuclear Information System (INIS)

    Gangradey, Ranjana; Mukherjee, Samiran Shanti; Agarwal, Jyoti

    2015-01-01

    Indigenous cryoadsorption cryopump with large pumping speeds for fusion reactor application has been developed at the Institute for Plasma Research (IPR). Towards its successful realization, technological bottlenecks were identified, studied and resolved. Hydroformed cryopanels were developed from concept leading to the design and product realization with successful technology transfer to the industry. This has led to the expertise for developing hydroformed panels for any desired shape, geometry and welding pattern. Activated sorbents were developed, characterized using an experimental set up which measures adsorption isotherms down to 4K for hydrogen and helium. Special techniques were evolved for coating sorbents on hydroformed cryopanels with suitable cryo-adhesives. Various arrangements of cryopanels at 4 K surrounded by 80 K shields and baffles (which are also hydroformed) were studied and optimized by transmission probability analysis using Monte Carlo techniques. CFD analysis was used to study the temperature distribution and flow analysis during the cryogen flow through the panels. Integration of the developed technologies to arrive at the final product was a challenging task and this was meticulously planned and executed. This resulted in a cryoadsorption cryopump offering pumping speeds as high as 50,000 to 70,000 1/s for helium and 1,50,000 1/s for hydrogen with a 3.2 m 2 of sorbent panel area. The first laboratory scale pump integrating the developed technologies was a Small Scale CryoPump (SSCP-01) with a pumping speed of 2,000 1/s for helium. Subsequently, Single Panel CryoPump (SPCP-01) with pumping speed 10,000 1/s for helium and a Multiple Panel CryoPump (MPCP-08) with a pumping speed of 70,000 1/s for helium and 1,50,000 1/s for hydrogen respectively were developed. This paper describes the efforts in realizing these products from laboratory to industrial scales. (author)

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

  18. Innovations and Enhancements for a Consortium of Big-10 University Research and Training Reactors. Final Report

    International Nuclear Information System (INIS)

    Brenizer, Jack

    2011-01-01

    The Consortium of Big-10 University Research and Training Reactors was by design a strategic partnership of seven leading institutions. We received the support of both our industry and DOE laboratory partners. Investments in reactor, laboratory and program infrastructure, allowed us to lead the national effort to expand and improve the education of engineers in nuclear science and engineering, to provide outreach and education to pre-college educators and students and to become a key resource of ideas and trained personnel for our U.S. industrial and DOE laboratory collaborators.

  19. Development of a thermionic-reactor space-power system. Final summary report

    International Nuclear Information System (INIS)

    1973-01-01

    Initial experimental work led to the award of the first AEC thermionic contract on May 1, 1962, for the development of fission heated thermionic cells with an operating life of 10,000 hours or more. Two types of converters were fabricated: (1) electrically heated, and (2) fission heated where the fuel was either uranium carbide or uranium oxide. Competition between GGA and GE was climaxed on July 1, 1970 by the award to GGA of a contract to develop an in-core thermionic reactor. This report is divided into the following: thermionic research, materials technology, thermionic fuel element development, reactor technology, and systems technology

  20. Critical review of the reactor-safety study radiological health effects model. Final report

    International Nuclear Information System (INIS)

    Cooper, D.W.; Evans, J.S.; Jacob, N.; Kase, K.R.; Maletskos, C.J.; Robertson, J.B.; Smith, D.G.

    1983-03-01

    This review of the radiological health effects models originally presented in the Reactor Safety Study (RSS) and currently used by the US Nuclear Regulatory Commission (NRC) was undertaken to assist the NRC in determining whether or not to revise the models and to aid in the revision, if undertaken. The models as presented in the RSS and as implemented in the CRAC (Calculations of Reactor Accident Consequences) Code are described and critiqued. The major elements analyzed are those concerning dosimetry, early effects, and late effects. The published comments on the models are summarized, as are the important findings since the publication of the RSS

  1. Critical review of the reactor-safety study radiological health effects model. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, D.W.; Evans, J.S.; Jacob, N.; Kase, K.R.; Maletskos, C.J.; Robertson, J.B.; Smith, D.G.

    1983-03-01

    This review of the radiological health effects models originally presented in the Reactor Safety Study (RSS) and currently used by the US Nuclear Regulatory Commission (NRC) was undertaken to assist the NRC in determining whether or not to revise the models and to aid in the revision, if undertaken. The models as presented in the RSS and as implemented in the CRAC (Calculations of Reactor Accident Consequences) Code are described and critiqued. The major elements analyzed are those concerning dosimetry, early effects, and late effects. The published comments on the models are summarized, as are the important findings since the publication of the RSS.

  2. Intercomparison of liquid metal fast reactor seismic analysis codes. V. 3: Comparison of observed effects with computer simulated effects on reactor cores from seismic disturbances. Proceedings of a final research co-ordination meeting

    International Nuclear Information System (INIS)

    1996-05-01

    This publication contains the final papers summarizing the validation of the codes on the basis of comparison of observed effects with computer simulated effects on reactor cores from seismic disturbances. Refs, figs tabs

  3. The development and application of advanced analytical methods to commercial ICF reactor chambers. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Cousseau, P.; Engelstad, R.; Henderson, D.L. [and others

    1997-10-01

    Progress is summarized in this report for each of the following tasks: (1) multi-dimensional radiation hydrodynamics computer code development; (2) 2D radiation-hydrodynamic code development; (3) ALARA: analytic and Laplacian adaptive radioactivity analysis -- a complete package for analysis of induced activation; (4) structural dynamics modeling of ICF reactor chambers; and (5) analysis of self-consistent target chamber clearing.

  4. An innovative fuel design concept for improved light water reactor performance and safety. Final technical report

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Connell, R.G.

    1995-07-01

    Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. The purpose of this research was to explore a technique for extending fuel performance by thermally bonding LWR fuel with a non-alkaline liquid metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide (UO 2 ) fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Due to the thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high conductivity liquid metal thermally bonds the fuel to the cladding, and eliminates the large temperature change across the gap, while preserving the expansion and pellet loading capabilities. The resultant lower fuel temperature directly impacts fuel performance limit margins and also core transient performance. The application of liquid bonding techniques to LWR fuel was explored for the purposes of increasing LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) has been developed under the program to analyze the in-reactor performance of the liquid metal bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of Liquid Metal Bonded LWR fuel

  5. Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 2, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

    1992-03-01

    This report contains a review of design studies for Inertial Confinement reactor. This second of three volumes discussions is some detail the following: Objectives, requirements, and assumptions; rationale for design option selection; key technical issues and R&D requirements; and conceptual design selection and description.

  6. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    International Nuclear Information System (INIS)

    1997-01-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits

  7. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits.

  8. The development and application of advanced analytical methods to commercial ICF reactor chambers. Final report

    International Nuclear Information System (INIS)

    Cousseau, P.; Engelstad, R.; Henderson, D.L.

    1997-10-01

    Progress is summarized in this report for each of the following tasks: (1) multi-dimensional radiation hydrodynamics computer code development; (2) 2D radiation-hydrodynamic code development; (3) ALARA: analytic and Laplacian adaptive radioactivity analysis -- a complete package for analysis of induced activation; (4) structural dynamics modeling of ICF reactor chambers; and (5) analysis of self-consistent target chamber clearing

  9. ENHANCEMENT OF EQUILIBRIUMSHIFT IN DEHYDROGENATION REACTIONS USING A NOVEL MEMBRANE REACTOR; FINAL

    International Nuclear Information System (INIS)

    Shamsuddin Ilias, Ph.d., P.E.; Franklin G. King, D.Sc.

    2001-01-01

    With the advances in new inorganic materials and processing techniques, there has been renewed interest in exploiting the benefits of membranes in many industrial applications. Inorganic and composite membranes are being considered as potential candidates for use in membrane-reactor configuration for effectively increasing reaction rate, selectivity and yield of equilibrium limited reactions. To investigate the usefulness of a palladium-ceramic composite membrane in a membrane reactor-separator configuration, we investigated the dehydrogenation of cyclohexane by equilibrium shift. A two-dimensional pseudo-homogeneous reactor model was developed to study the dehydrogenation of cyclohexane by equilibrium shift in a tubular membrane reactor. Radial diffusion was considered to account for the concentration gradient in the radial direction due to permeation through the membrane. For a dehydrogenation reaction, the feed stream to the reaction side contained cyclohexane and argon, while the separation side used argon as the sweep gas. Equilibrium conversion for dehydrogenation of cyclohexane is 18.7%. The present study showed that 100% conversion could be achieved by equilibrium shift using Pd-ceramic membrane reactor. For a feed containing cyclohexane and argon of 1.64 x 10(sup -6) and 1.0 x 10(sup -3) mol/s, over 98% conversion could be readily achieved. The dehydrogenation of cyclohexane was also experimentally investigated in a palladium-ceramic membrane reactor. The Pd-ceramic membrane was fabricated by electroless deposition of palladium on ceramic substrate. The performance of Pd-ceramic membrane was compared with a commercially available hydrogen-selective ceramic membrane. From limited experimental data it was observed that by appropriate choice of feed flow rate and sweep gas rate, the conversion of cyclohexane to benzene and hydrogen can increased to 56% at atmospheric pressure and 200 C in a Pd-ceramic membrane reactor. In the commercial ceramic membrane

  10. Conceptuation of a continuously working vacuum pump train for fusion power plants

    International Nuclear Information System (INIS)

    Giegerich, Thomas; Day, Christian

    2013-01-01

    The Karlsruhe Institute of Technology (KIT) is developing a continuously working and non-cryogenic pumping solution for a demonstration power plant (DEMO). This pumping train shall cover the full operational pressure regime of a fusion reactor and is based on two pump types, namely diffusion pumps and liquid ring pumps. The whole pumping train must fulfill high safety and reliability requirements and it has to be made fully tritium compatible. In this paper, the design of a prototype pumping train and the special requirements for a DEMO machine are presented and discussed. A central feature of this pumping train is the use of a liquid metal as tritium compatible working fluid in both pump types, that leads to a pumping train which is able to cover a pressure range of 12 decades, namely from 10 −9 to 10 3 mbar. Finally, a test facility for pump testing over a wide pressure regime is described. In this facility (THESEUS), experiments with a diffusion pump have been performed and first results are presented

  11. Conceptuation of a continuously working vacuum pump train for fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Giegerich, Thomas, E-mail: thomas.giegerich@kit.edu; Day, Christian

    2013-10-15

    The Karlsruhe Institute of Technology (KIT) is developing a continuously working and non-cryogenic pumping solution for a demonstration power plant (DEMO). This pumping train shall cover the full operational pressure regime of a fusion reactor and is based on two pump types, namely diffusion pumps and liquid ring pumps. The whole pumping train must fulfill high safety and reliability requirements and it has to be made fully tritium compatible. In this paper, the design of a prototype pumping train and the special requirements for a DEMO machine are presented and discussed. A central feature of this pumping train is the use of a liquid metal as tritium compatible working fluid in both pump types, that leads to a pumping train which is able to cover a pressure range of 12 decades, namely from 10{sup −9} to 10{sup 3} mbar. Finally, a test facility for pump testing over a wide pressure regime is described. In this facility (THESEUS), experiments with a diffusion pump have been performed and first results are presented.

  12. International benchmark for coupled codes and uncertainty analysis in modelling: switching-Off of one of the four operating main circulation pumps at nominal reactor power at NPP Kalinin unit 3

    International Nuclear Information System (INIS)

    Tereshonok, V. A.; Nikonov, S. P.; Lizorkin, M. P.; Velkov, K; Pautz, A.; Ivanov, V.

    2008-01-01

    The paper briefly describes the Specification of an international NEA/OECD benchmark based on measured plant data. During the commissioning tests for nominal power at NPP Kalinin Unit 3 a lot of measurements of neutron and thermo-hydraulic parameters have been carried out in the reactor pressure vessel, primary and the secondary circuits. One of the measured data sets for the transient 'Switching-off of one Main Circulation Pump (MCP) at nominal power' has been chosen to be applied for validation of coupled thermal-hydraulic and neutron-kinetic system codes and additionally for performing of uncertainty analyses as a part of the NEA/OECD Uncertainty Analysis in Modeling Benchmark. The benchmark is opened for all countries and institutions. The experimental data and the final specification with the cross section libraries will be provided to the participants from NEA/OECD only after official declaration of real participation in the benchmark and delivery of the simulated results of the transient for comparison. (Author)

  13. Atomic data for heavy element impurities in fusion reactors. Summary report of the final research coordination meeting

    International Nuclear Information System (INIS)

    Clark, R.E.H.

    2009-04-01

    Eleven experts on the properties of heavy elements of relevance to fusion energy research attended the final Research Coordination Meeting (RCM) on Data for Heavy Element Impurities in Fusion Reactors, held at IAEA Headquarters on 4-6 March 2009. Participants summarized their accomplishments with respect to the revised work plan formulated at the second RCM. The overall work plan for the CRP was assessed and reviewed in detail, and achievements were noted. Discussions, conclusions and recommendations of the RCM are briefly described in this report. (author)

  14. Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen; Aldemir, Tunc

    2009-09-10

    The project entitled, “Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors”, was conducted as a DOE NERI project collaboration between Texas A&M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

  15. Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

    International Nuclear Information System (INIS)

    Hawari, Ayman

    2014-01-01

    This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

  16. Evaluation of performance of select fusion experiments and projected reactors. Final report

    International Nuclear Information System (INIS)

    Miley, G.H.

    1978-10-01

    The performance of NASA Lewis fusion experiments (SUMMA and Bumpy Torus) is compared with other experiments and that necessary for a power reactor. Key parameters cited are gain (fusion power/input power) and the time average fusion power, both of which may be more significant for real fusion reactors than the commonly used Lawson parameter. The NASA devices are over 10 orders of magnitude below the required powerplant values in both gain and time average power. The best experiments elsewhere are also as much as 4 to 5 orders of magnitude low. However, the NASA experiments compare favorably with other alternate approaches that have received less funding than the mainline experiments. The steady-state character and efficiency of plasma heating are strong advantages of the NASA approach. The problem, though, is to move ahead to experiments of sufficient size to advance in gain and average power parameters

  17. Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

    International Nuclear Information System (INIS)

    Lee, D.D.; Collins, J.L.

    2000-01-01

    One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required

  18. Tritium diffusion in nonmetallic solids of interest for fusion reactors. Final report

    International Nuclear Information System (INIS)

    Elleman, T.

    1979-01-01

    Tritium diffusion measurements have been conducted in Al 2 O 3 , BeO, Y 2 O 3 , SiC, B 4 C, Si 3 N 4 and pyrolytic carbon as a basis for evaluating these materials as potential tritium barriers in fusion reactors. Deuterium solubility measurements were conducted with Al 2 O 3 , SiC and pyrolytic carbon to establish the pressure and temperature dependence of solubility and to identify solubility ranges. Hydrogen permeability measurements on commercially available Al 2 O 3 and SiC materials were used as a check on calculated permeabilities and to provide data on hydrogen permeation rates in polycrystalline materials. The diffusion, solubility and permeation results are presented and discussed in terms of fusion reactor applications

  19. Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.D.; Collins, J.L.

    2000-02-01

    One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

  20. Store and process for intermediate or final storage of used fuel elements from a nuclear reactor

    International Nuclear Information System (INIS)

    Kumpf, H.

    1986-01-01

    The fuel elements are enclosed in boxes at the nuclear reactor and transported in these to the incoming station. Transport is a by truck, which makes it possible for the transport container to move in a vertical position, where the upper side is on the top side of the truck. The fuel elements in their boxes are handed over to a magazine there, which can be reached by a loading machine serving the storage room. (orig./HP) [de

  1. RETU The Finnish research programme on reactor safety 1995-1998. Final Symposium

    International Nuclear Information System (INIS)

    Vanttola, T.

    1998-01-01

    The Reactor Safety (RETU, 1995-1998) research programme concentrated on search of safe limits for nuclear fuel and the reactor core, accident management methods and risk management of nuclear power plants. The total volume of the programme was 100 person years and funding FIM 58 million. This symposium report summarises the research fields, the objectives and the main results obtained. In the field of operational margins of a nuclear reactor, the behaviour of high burnup nuclear fuel was studied both in normal operation and during power transients. The static and dynamic reactor analysis codes were developed and validated to cope with new fuel designs and complicated three-dimensional reactivity transients. Advanced flow models and numerical solution methods for the dynamics codes were developed and tested. Research on accident management developed and validated calculation methods needed to plan preventive measures and to train the personnel to severe accident mitigation. Efforts were made to reduce uncertainties in phenomena important in severe accidents and to study actions planned for accident management. The programme included experimental work, but also participation in large international tests. The Finnish thermal-hydraulic test facility PACTEL was used extensively for the evaluation of the VVER-440 plant accident behaviour, for the validation of the accident analysis computer codes and for the testing of passive safety system concepts for future plant designs. In risk management probabilistic methods were developed for safety related decision making and for complex event sequences. Effects of maintenance on safety were studied and effective methods for assessment of human reliability and safety critical organisations were searched. To enhance human competencies in control of complex environments, practical tools for training and continuous learning were worked out, and methods to evaluate appropriateness of control room design were developed. (orig)

  2. Mirror Advanced Reactor Study (MARS). Final report. Volume 2. Commercial fusion synfuels plant

    International Nuclear Information System (INIS)

    Donohue, M.L.; Price, M.E.

    1984-07-01

    Volume 2 contains the following chapters: (1) synfuels; (2) physics base and parameters for TMR; (3) high-temperature two-temperature-zone blanket system for synfuel application; (4) thermochemical hydrogen processes; (5) interfacing the sulfur-iodine cycle; (6) interfacing the reactor with the thermochemical process; (7) tritium control in the blanket system; (8) the sulfur trioxide fluidized-bed composer; (9) preliminary cost estimates; and (10) fuels beyond hydrogen

  3. University Reactor Instrumentation Program. Final report, September 30, 1993--March 31, 1996

    International Nuclear Information System (INIS)

    1997-01-01

    The University of Massachusetts Lowell Research Reactor has received a total of $115,723.00 from the Department of Energy (DOE) Instrumentation Program (DOE Grant No. DE-FG02-91ID13083) and $40,000 in matching funds from the University of Massachusetts Lowell administration. The University of Massachusetts Lowell Research Reactor has been serving the University and surrounding communities since it first achieved criticality in May 1974. The principle purpose of the facility is to provide a multidisciplinary research and training center for the University of Massachusetts Lowell and other New England academic institutions. The facility promotes student and industrial research, in addition to providing education and training for nuclear scientists, technicians, and engineers. The 1 MW thermal reactor contains a variety of experimental facilities which, along with a 0.4 megacurie cobalt source, effectively supports the research and educational programs of many university departments including Biology, Chemistry, Nuclear and Plastics Engineering, Radiological Sciences, Physics, and other campuses of the University of Massachusetts system. Although the main focus of the facility is on intra-university research, use by those outside the university is fully welcomed and highly encouraged

  4. PWR reactor pressure vessel internals license renewal industry report; revision 1. Final report

    International Nuclear Information System (INIS)

    Schwirian, R.; Robison, G.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). Pressurized water reactor (PWR) reactor pressure vessel (RPV) internals designed by all three U.S. PWR nuclear steam supply system vendors have been evaluated relative to the effects of age-related degradation mechanisms; the capability of current design limits; inservice examination, testing, repair, refurbishment, and other programs to manage these effects; and the assurance that these internals can continue to perform their intended safety functions in the license renewal term. This industry report (IR), one of a series of ten, provides a generic technical basis for evaluation of PWR reactor pressure vessel internals for license renewal

  5. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    1997-01-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits

  6. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits.

  7. Final report of the HFIR [High Flux Isotope Reactor] irradiation facilities improvement project

    International Nuclear Information System (INIS)

    Montgomery, B.H.; Thoms, K.R.; West, C.D.

    1987-09-01

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Final environmental statement, Liquid Metal Fast Breeder Reactor Program. Volume 1

    International Nuclear Information System (INIS)

    1975-12-01

    Information is presented under the following section headings: LMFBR program options and their compatibility with the major issues affecting commercial development, Proposed Final Environmental Statement for the LMFBR program, December 1974, WASH-1535, supplemental material, and material relating to Proposed Final Environmental Statement review

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

    International Nuclear Information System (INIS)

    Cahalana, J. E.; Hahn, D.

    2007-01-01

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  11. Decommissioning techniques for research reactors. Final report of a co-ordinated research project 1997-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-02-01

    In its international role, the IAEA is faced with a wide variety of national situations and different availability of technical, human and financial resources. While it is recognised that nuclear decommissioning is a mature industry in some developed countries, and may soon become a routine activity, the situation is by no means so clear in other countries. In addition, transfer of technologies and know-how from developed to developing countries is not a spontaneous, straightforward process, and will take time and considerable effort. As mandated by its own statute and Member States' requests, the IAEA continues to respond to its Member States by monitoring technological progress, ensuring development of safer and more efficient strategies and fostering international information exchange. Previous co-ordinated research projects (CRP) conducted respectively from 1984 to 1987, and from 1989 to 1993, investigated the overall domain of decommissioning. In those CRPs no distinction was made between decommissioning activities carried out at nuclear power plants, research reactors or nuclear fuel cycle facilities. With technological progress and experience gained, it became clear that decommissioning of research reactors had certain specific characteristics which needed a dedicated approach. In addition, a large number of research reactors reached a state of permanent shutdown in the 1990s and were candidates for prompt decommissioning. With the progressive ageing of research reactors, many more of these units will soon become redundant worldwide and require decommissioning. Within this context, a CRP on Decommissioning Techniques for Research Reactors was launched and conducted by the IAEA from 1997 to 2001 in order to prepare for eventual decommissioning. Concluding reports that summarized the work undertaken under the aegis of the CRP were presented at the third and final Research Co-ordination Meeting held in Kendal, United Kingdom, 14-18 May 2001, and are collected

  12. Decommissioning techniques for research reactors. Final report of a co-ordinated research project 1997-2001

    International Nuclear Information System (INIS)

    2002-02-01

    In its international role, the IAEA is faced with a wide variety of national situations and different availability of technical, human and financial resources. While it is recognised that nuclear decommissioning is a mature industry in some developed countries, and may soon become a routine activity, the situation is by no means so clear in other countries. In addition, transfer of technologies and know-how from developed to developing countries is not a spontaneous, straightforward process, and will take time and considerable effort. As mandated by its own statute and Member States' requests, the IAEA continues to respond to its Member States by monitoring technological progress, ensuring development of safer and more efficient strategies and fostering international information exchange. Previous co-ordinated research projects (CRP) conducted respectively from 1984 to 1987, and from 1989 to 1993, investigated the overall domain of decommissioning. In those CRPs no distinction was made between decommissioning activities carried out at nuclear power plants, research reactors or nuclear fuel cycle facilities. With technological progress and experience gained, it became clear that decommissioning of research reactors had certain specific characteristics which needed a dedicated approach. In addition, a large number of research reactors reached a state of permanent shutdown in the 1990s and were candidates for prompt decommissioning. With the progressive ageing of research reactors, many more of these units will soon become redundant worldwide and require decommissioning. Within this context, a CRP on Decommissioning Techniques for Research Reactors was launched and conducted by the IAEA from 1997 to 2001 in order to prepare for eventual decommissioning. Concluding reports that summarized the work undertaken under the aegis of the CRP were presented at the third and final Research Co-ordination Meeting held in Kendal, United Kingdom, 14-18 May 2001, and are collected

  13. Circulation pump mounting

    International Nuclear Information System (INIS)

    Skalicky, A.

    1976-01-01

    The suspension is described of nuclear reactor circulating pumps enabling their dilatation with a minimum reverse force consisting of spacing rods supported with one end in the anchor joints and provided with springs and screw joints engaging the circulating pump shoes. The spacing rods are equipped with side vibration dampers anchored in the shaft side wall and on the body of the circulating pump drive body. The negative reverse force F of the spacing rods is given by the relation F=Q/l.y, where Q is the weight of the circulating pump, l is the spatial distance between the shoe joints and anchor joints, and y is the deflection of the circulating pump vertical axis from the mean equilibrium position. The described suspension is advantageous in that that the reverse force for the deflection from the mean equilibrium position is minimal, dynamic behaviour is better, and construction costs are lower compared to suspension design used so far. (J.B.)

  14. High thermal efficiency, radiation-based advanced fusion reactors. Final report

    International Nuclear Information System (INIS)

    Taussig, R.T.

    1977-04-01

    A new energy conversion scheme is explored in this study which has the potential of achieving thermal cycle efficiencies high enough (e.g., 60 to 70 percent) to make advanced fuel fusion reactors attractive net power producers. In this scheme, a radiation boiler admits a large fraction of the x-ray energy from the fusion plasma through a low-Z first wall into a high-Z working fluid where the energy is absorbed at temperatures of 2000 0 K to 3000 0 K. The hot working fluid expands in an energy exchanger against a cooler, light gas, transferring most of the work of expansion from one gas to the other. By operating the radiation/boiler/energy exchanger as a combined cycle, full advantage of the high temperatures can be taken to achieve high thermal efficiency. The existence of a mature combined cycle technology from the development of space power plants gives the advanced fuel fusion reactor application a firm engineering base from which it can grow rapidly, if need be. What is more important, the energy exchanger essentially removes the peak temperature limitations previously set by heat engine inlet conditions, so that much higher combined cycle efficiencies can be reached. This scheme is applied to the case of an advanced fuel proton-boron 11 fusion reactor using a single reheat topping and bottoming cycle. A wide variety of possible working fluid combinations are considered and particular cycle calculations for the thermal efficiency are presented. The operation of the radiation boiler and energy exchanger are both described. Material compatibility, x-ray absorption, thermal hydraulics, structural integrity, and other technical features of these components are analyzed to make a preliminary assessment of the feasibility of this concept

  15. Evaluation and prediction of neutron embrittlement in reactor pressure vessel materials. Final report

    International Nuclear Information System (INIS)

    Hawthorne, J.R.; Menke, B.H.; Loss, F.J.; Watson, H.E.; Hiser, A.L.; Gray, R.A.

    1982-12-01

    This study evaluates the effects of fast neutron irradiation on the mechanical properties of eight nuclear reactor vessel materials. The materials include submerged arc weldments, three plates, and one forging. The materials are in the unirradiated and irradiated conditions with regard to tensile, Charpy impact, and static and dynamic fracture toughness properties. Correlations between impact and fracture toughness parameters are developed from the experimental results. The observed shifts in transition temperature and the drop in upper-shelf energy are compared with predictions developed from the Regulatory Guide 1.99.1 trend curves

  16. Conceptual studies of toroidal field magnets for the tokamak experimental power reactor. Final report

    International Nuclear Information System (INIS)

    Buncher, B.R.; Chi, J.W.H.; Fernandez, R.

    1976-01-01

    This report documents the principal results of a Conceptual Design Study for the Superconducting Toroidal Field System for a Tokamak Experimental Power Reactor. Two concepts are described for peak operating fields at the windings of 8 tesla, and 12 tesla, respectively. The design and manufacturing considerations are treated in sufficient detail that cost and schedule estimates could be developed. Major uncertainties in the design are identified and their potential impact discussed, along with recommendations for the necessary research and development programs to minimize these uncertainties. The minimum dimensions of a sub-size test coil for experimental qualification of the full size design are developed and a test program is recommended

  17. Evaluation of improved light water reactor core designs. Final progress report, September 1979. LWRCD-20

    International Nuclear Information System (INIS)

    1979-01-01

    The work conducted under this research project has developed information which supports in all respects the U.S. position evolved under the NASAP/INFCE programs with respect to the near and intermediate term potential for ore conservation in LWRs on the once-through fuel cycle. Moreover, in the even longer term, it has been confirmed that contention by Edlund and others that tight-pitch Pu/UO 2 PWR cores can achieve conversion ratios which may allow these reactors to provide a competitive energy source far into the ore-scarce post-2000 era

  18. Mitigation of intergranular stress corrosion cracking in RBMK reactors. Final report of the programme's steering committee

    International Nuclear Information System (INIS)

    2002-09-01

    In 2000 the IAEA initiated an Extrabudgetary Programme on Mitigation of Intergranular Stress Corrosion Cracking in RBMK Reactors to assist countries operating RBMK reactors in addressing the issue in austenitic stainless steel 300 mm diameter piping. Intergranular stress corrosion cracking of austenitic stainless steel piping in BWRs has been a major safety concern since the early seventies. Similar degradation was found in RBMK reactor piping in 1997. Early in 1998 the IAEA responded to requests for assistance from RBMK operating countries on this issue through activities organized in the framework of Technical Co-operation Department regional projects and the Extrabudgetary Programme on the Safety of WWER and RBMK Nuclear Power Plants. Results of these activities were a basis for the formulation of the objective and scope of the Extrabudgetary Programme on Mitigation of Intergranular Stress Corrosion Cracking in RBMK reactors ('the Programme'). The scope of the Programme included in-service inspection, assessment, repair and mitigation, and water chemistry and decontamination. The Programme was pursued by means of exchange of experience, formulation of guidance, transfer of technology, and training, which will assist the RBMK operators to address related safety concerns. The Programme implementation relied on voluntary extrabudgetary financial contributions from Japan, Spain, the United Kingdom and the USA, and on in kind contributions from Finland, Germany and Sweden. The Programme was implemented in close co-ordination with ongoing national and bilateral activities and major inputs to the Programme were provided through the activities of the Swedish International Project Nuclear Safety and of the US DOE International Nuclear Safety Program. The RBMK nuclear power plants in Lithuania, Russian Federation and Ukraine hosted most of the Programme activities. Support of these Member States involved in the Programme was instrumental for its successful completion in

  19. Final design of the neutral beam lines for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Pittenger, L.C.; Valby, L.E.; Stone, R.R.; Pedrotti, L.R.; Denhoy, B.; Yoard, R.

    1979-01-01

    Final design of the neutral beam lines for TFTR has been completed. A prototype has been assembled at Lawrence Berkeley Laboratory and is undergoing testing as part of the Neutral Beam System Test Facility (NBSTF). The final neutral beam line (NBL) configuration differs in several details from that previously reported upon; certain components have been added; and testing of the cryopump system has led to some design simplification. It is these developments which are reported herein

  20. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  1. Statistical analysis of nuclear power plant pump failure rate variability: some preliminary results

    International Nuclear Information System (INIS)

    Martz, H.F.; Whiteman, D.E.

    1984-02-01

    In-Plant Reliability Data System (IPRDS) pump failure data on over 60 selected pumps in four nuclear power plants are statistically analyzed using the Failure Rate Analysis Code (FRAC). A major purpose of the analysis is to determine which environmental, system, and operating factors adequately explain the variability in the failure data. Catastrophic, degraded, and incipient failure severity categories are considered for both demand-related and time-dependent failures. For catastrophic demand-related pump failures, the variability is explained by the following factors listed in their order of importance: system application, pump driver, operating mode, reactor type, pump type, and unidentified plant-specific influences. Quantitative failure rate adjustments are provided for the effects of these factors. In the case of catastrophic time-dependent pump failures, the failure rate variability is explained by three factors: reactor type, pump driver, and unidentified plant-specific influences. Finally, point and confidence interval failure rate estimates are provided for each selected pump by considering the influential factors. Both types of estimates represent an improvement over the estimates computed exclusively from the data on each pump

  2. Lunar electric power systems utilizing the SP-100 reactor coupled to dynamic conversion systems. Final report

    International Nuclear Information System (INIS)

    Harty, R.B.; Durand, R.E.

    1993-03-01

    An integration study was performed by Rocketdyne under contract to NASA-LeRC. The study was concerned with coupling an SP-0100 reactor to either a Brayton or Stirling power conversion system. The application was for a surface power system to supply power requirements to a lunar base. A power level of 550 kWe was selected based on the NASA Space Exploration Initiative 90-day study. Reliability studies were initially performed to determine optimum power conversion redundancy. This study resulted in selecting three operating engines and one stand-by unit. Integration design studies indicated that either the Brayton or Stirling power conversion systems could be integrated with the PS-100 reactor. The Stirling system had an integration advantage because of smaller piping size and fewer components. The Stirling engine, however, is more complex and heavier than the Brayton rotating unit, which tends to off-set the Stirling integration advantage. From a performance consideration, the Brayton had a 9 percent mass advantage, and the Stirling had a 50 percent radiator advantage

  3. Improved control rod drive handling equipment for BWRs [boiling-water reactors]: Final report

    International Nuclear Information System (INIS)

    Turner, A.P.L.; Gorman, J.A.

    1987-08-01

    Improved equipment for removing and replacing control rod drives (CRDs) in BWR plants has been designed, built and tested. Control rod drives must be removed from the reactor periodically for servicing. Removal and replacement of CRDs using equipment originally supplied with the plant has long been recognized as one of the more difficult and highest radiation exposure maintenance operations that must be performed at BWR plants. The improved equipment was used for the first time at Quad Cities, Unit 2, during a Fall 1986 outage. The trial of the equipment was highly successful, and it was shown that the new equipment significantly improves CRD handling operations. The new equipment significantly simplifies the sequence of operations required to lower a CRD from its housing, upend it to a horizontal orientation, and transport it out of the reactor containment. All operations of the new equipment are performed from the undervessel equipment handling platform, thus, eliminating the requirement for a person to work on the lower level of the undervessel gallery which is often highly contaminated. Typically, one less person is required to operate the equipment than were used with the older equipment. The new equipment incorporates a number of redundant and fail safe features that improve operations and reduce the chances for accidents

  4. Optimization and implementation study of plutonium disposition using existing CANDU Reactors. Final report

    International Nuclear Information System (INIS)

    1996-09-01

    Since early 1994, the Department of Energy has been sponsoring studies aimed at evaluating the merits of disposing of surplus US weapons plutonium as Mixed Oxide (MOX) fuel in existing commercial Canadian Pressurized Heavy Water reactors, known as CANDU's. The first report, submitted to DOE in July, 1994 (the 1994 Executive Summary is attached), identified practical and safe options for the consumption of 50 to 100 tons of plutonium in 25 years in some of the existing CANDU reactors operating the Bruce A generating station, on Lake Huron, about 300 km north east of Detroit. By designing the fuel and nuclear performance to operate within existing experience and operating/performance envelope, and by utilizing existing fuel fabrication and transportation facilities and methods, a low cost, low risk method for long term plutonium disposition was developed. In December, 1995, in response to evolving Mission Requirements, the DOE requested a further study of the CANDU option with emphasis on more rapid disposition of the plutonium, and retaining the early start and low risk features of the earlier work. This report is the result of that additional work

  5. Evaluation of spectral shift controlled reactors operating on the uranium fuel cycle. Final report

    International Nuclear Information System (INIS)

    Matzie, R.A.; Sider, F.M.

    1979-08-01

    The performance of the spectral shift controlled reactor (SSCR) operating on uranium fuel cycles was evaluated and compared with the conventional pressurized water reactor (PWR). In order to analyze the SSCR, the PSR design methodology was extended to include systems moderated by mixtures of light water and heavy water and these methods were validated by comparison with experimental results. Once the design methods had been formulated, the resouce requirements and power costs were determined for the uranium-fueled SSCR. The ore requirements of the UO 2 once-through fuel cycle and the UO 2 fuel cycle with self-generated recycle (SGR) of plutonium were found to be 10% and 19% less than those of similarly fueled PWRs, respectively. A fuel cycle optimization study was performed for the UO 2 once-through SSCR and the SGR SSCR. By individually altering lattice parameters, discharge exposure or number of in-core batches, savings of less than 8% in resource requirements and less than 1% in power costs were obtained

  6. URI Program Final Report FY 2001 Grant for the University of Florida Training Reactor

    International Nuclear Information System (INIS)

    Vernetson, W.G.

    2004-01-01

    The purpose of the URI program is to upgrade and improve university nuclear research and training reactors and to contribute to strengthening the academic community's nuclear engineering infrastructure. It should be noted that the proposed UFTR facility instrumentation and equipment can generally be subdivided into three categories: (1) to improve reactor operations, (2) to improve existing facility/NAA Laboratory operations, and (3) to expand facility capability. All of these items were selected recognizing the objectives of the University Reactor Instrumentation Program to respond to the widespread needs in the academic reactor community for modernization and improvement of research and training reactor facilities, especially at large and diverse institutions such as the University of Florida. These needs have been particularly pressing at the UFTR which is the only such research and training reactor in the State of Florida which is undergoing rapid growth in a variety of technical areas. As indicated in Table 2, the first item is a security system control panel with associated wiring and detectors. The existing system is over 30 years old and has been the subject of repeated maintenance over the past 5 years. Some of its detection devices are no longer replaceable from stock. Modifications made many years ago make troubleshooting some parts of the system such as the backup battery charging subsystem essentially impossible, further increasing maintenance frequency to replace batteries. Currently, various parts of the system cable trays remain open for maintenance access further degrading facility appearance. In light of relicensing plans, this item is also a key consideration for housekeeping appearance considerations. The cost of a replacement ADEMCO Vista 20 security system including turnkey installation by a certified vendor was to be $2,206. Replacement of this system was expected to save up to 5 days of maintenance per year, decrease security alarm response

  7. Corrosion of aluminium-clad spent fuel in LVR-15 research reactor storage facilities. Final report

    International Nuclear Information System (INIS)

    Splichal, K.; Berka, J.; Keilova, E.

    2006-03-01

    The corrosion of the research reactor aluminium clad spent fuel in water was investigated in two storage facilities. The standard racks were delivered by the IAEA and consisted of two aluminium alloys AA 6061 and Szav-1 coupons. Bimetallic couples create aluminium alloy and stainless steel 304 coupons. Rolled and extruded AA 6061 material was also tested. Single coupons, bimetallic and crevice couples were exposed in the at-reactor basin (ARB) and the high-level wastage pool (HLW). The water chemistry parameters were monitored and sedimentation of impurities was measured. The content of impurities of mainly Cl and SO 4 was in the range of 2 to 15 μg/l in the HLW pool; it was about one order higher in ARB. The Fe content was below 2 μg/l for both facilities. After two years of exposure the pitting was evaluated as local corrosion damage. The occurrence of pits was evaluated predominantly on the surfaces of single coupons and on the outer and inner surfaces of bimetallic and crevices coupons. No correlation was found between the pitting initiation and the type of aluminium alloys and rolled and extruded materials. In bimetallic couples the presence of stainless coupons did not have any effect on local corrosion. The depth of pits was lower than 50 μm for considerable areas of coupons and should be compared with the results of other participating institutes. (author)

  8. Penis Pump

    Science.gov (United States)

    ... your appointment might be less involved. Choosing a penis pump Some penis pumps are available without a ... it doesn't get caught in the ring. Penis pumps for penis enlargement Many advertisements in magazines ...

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

    International Nuclear Information System (INIS)

    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. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-01

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO 2 -free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR designs and their ability to accept a

  11. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J; Van Den Durpel, L; Chauvet, V; Cerullo, N; Cetnar, J; Abram, T; Bakker, K; Bomboni, E; Bernnat, W; Domanska, J G; Girardi, E; De Haas, J B.M.; Hesketh, K; Hiernaut, J P; Hossain, K; Jonnet, J; Kim, Y; Kloosterman, J L; Kopec, M; Murgatroyd, J; Millington, D; Lecarpentier, D; Lomonaco, G; McEachern, D; Meier, A; Mignanelli, M; Nabielek, H; Oppe, J; Petrov, B Y; Pohl, C; Ruetten, H J; Schihab, S; Toury, G; Trakas, C; Venneri, F; Verfondern, K; Werner, H; Wiss, T; Zakova, J

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR designs and their ability to accept a

  12. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR

  13. Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

    Energy Technology Data Exchange (ETDEWEB)

    Woods, Brian; Gutowska, Izabela; Chiger, Howard

    2018-03-26

    Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion to ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with

  14. Instrumentation Needs for Integral Primary System Reactors (IPSRs) - Task 1 Final Report

    International Nuclear Information System (INIS)

    Gary D Storrick; Bojan Petrovic; Luca Oriani; Lawrence E Conway; Diego Conti

    2005-01-01

    This report presents the results of the Westinghouse work performed under Task 1 of this Financial Assistance Award and satisfies a Level 2 Milestone for the project. While most of the signals required for control of IPSRs are typical of other PWRs, the integral configuration poses some new challenges in the design or deployment of the sensors/instrumentation and, in some cases, requires completely new approaches. In response to this consideration, the overall objective of Task 1 was to establish the instrumentation needs for integral reactors, provide a review of the existing solutions where available, and, identify research and development needs to be addressed to enable successful deployment of IPSRs. The starting point for this study was to review and synthesize general characteristics of integral reactors, and then to focus on a specific design. Due to the maturity of its design and availability of design information to Westinghouse, IRIS (International Reactor Innovative and Secure) was selected for this purpose. The report is organized as follows. Section 1 is an overview. Section 2 provides background information on several representative IPSRs, including IRIS. A review of the IRIS safety features and its protection and control systems is used as a mechanism to ensure that all critical safety-related instrumentation needs are addressed in this study. Additionally, IRIS systems are compared against those of current advanced PWRs. The scope of this study is then limited to those systems where differences exist, since, otherwise, the current technology already provides an acceptable solution. Section 3 provides a detailed discussion on instrumentation needs for the representative IPSR (IRIS) with detailed qualitative and quantitative requirements summarized in the exhaustive table included as Appendix A. Section 3 also provides an evaluation of the current technology and the instrumentation used for measurement of required parameters in current PWRs. Section 4

  15. Study on an innovative fast reactor utilizing hydride neutron absorber - Final report of phase I study

    International Nuclear Information System (INIS)

    Konashi, K.; Iwasaki, T.; Itoh, K.; Hirai, M.; Sato, J.; Kurosaki, K.; Suzuki, A.; Matsumura, Y.; Abe, S.

    2010-01-01

    These days, the demand to use nuclear resources efficiently is growing for long-term energy supply and also for solving the green house problem. It is indispensable to develop technologies to reduce environmental load with the nuclear energy supply for sustainable development of human beings. In this regard, the development of the fast breeder reactor (FBR) is preferable to utilize nuclear resources effectively and also to burn minor actinides which possess very long toxicity for more than thousands years if they are not extinguished. As one of the FBR developing works in Japan this phase I study started in 2006 to introduce hafnium (Hf) hydride and Gadolinium-Zirconium (Gd-Zr) hydride as new control materials in FBR. By adopting them, the FBR core control technology is improved by two ways. One is extension of control rod life time by using long life Hf hydride which leads to reduce the fabrication and disposal cost and the other is reduction of the excess reactivity by adopting Gd-Zr hydride which leads to reduce the number of control rods and simplifies the core upper structure. This three year study was successfully completed and the following results were obtained. The core design was performed to examine the applicability of the Hf hydride absorber to Japanese Sodium Fast Reactor (JSFR) and it is clarified that the control rod life time can be prolonged to 6 years by adopting Hf hydride and the excess reactivity of the beginning of the core cycle can be reduced to half and the number of the control rods is also reduced to half by using the Gd-Zr hydride burnable poison. The safety analyses also certified that the core safety can be maintained with the same reliability of JSFR Hf hydride and Gd-Zr hydride pellets were fabricated in good manner and their basic features for design use were measured by using the latest devices such as SEM-EDX. In order to reduce the hydrogen transfer through the stainless steel cladding a new technique which shares calorizing

  16. Final hazard classification for the 100-HR-3 and 100-KR-4 pump-and-treat systems

    International Nuclear Information System (INIS)

    Oestreich, D.K.

    1996-09-01

    The purpose of this hazard classification report is to document the final hazard classification for two treatment facilities that will be located in the 100-HR-3 and 100-KR-4 Operable Units. These facilities will be sited at the 100-H and 100-K Areas for removal of chromium VI from the groundwater. The FHC is based on the hazardous material inventory (both chromium and radionuclides) that accumulates on the anion-exchange resin up to the point of resin saturation. Segmentation of the total inventory can be readily justified because the inventory resides in a number of separate resin vessels, and it is unreasonable to consider that a release could take place from more than one vessel at a time as a result of common-cause failure

  17. High-Temperature Salt Pump Review and Guidelines - Phase I Report

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hazelwood, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-05-01

    Fluoride salt cooled high-temperature reactor (FHR) concepts include pumps for forced circulation of the primary and secondary coolants. As part of a cooperative research and development agreement between the Shanghai Institute of Applied Physics and the Oak Ridge National Laboratory (ORNL), a research project was initiated to aid in the development of pumps for high-temperature salts. The objectives of the task included characterization of the behavior of an existing ORNL LSTL pump; design and test a modified impeller and volute for improved pump characteristics; and finally, provide lessons learned, recommendations, and guidelines for salt pump development and design. The pump included on the liquid salt test loop (LSTL) at ORNL served as a case study. This report summarizes the progress to date. The report is organized as follows. First, there is a review, focused on pumps, of the significant amount of work on salts at ORNL during the 1950s 1970s. The existing pump on the LSTL is then described. Plans for hot and cold testing of the pump are then discussed, including the design for a cold shakedown test stand and the required LSTL modifications for hot testing. Initial hydraulic and vibration modeling of the LSTL pump is documented. Later, test data from the LSTL will be used to validate the modeling approaches, which could then be used for future pump design efforts. Some initial insights and test data from the pump are then provided. Finally, some preliminary design goals and requirements for a future LSTL pump are provided as examples of salt pump design considerations.

  18. High-Temperature Salt Pump Review and Guidelines - Phase I Report

    International Nuclear Information System (INIS)

    Robb, Kevin R.; Jain, Prashant K.; Hazelwood, Thomas J.

    2016-01-01

    Fluoride salt cooled high-temperature reactor (FHR) concepts include pumps for forced circulation of the primary and secondary coolants. As part of a cooperative research and development agreement between the Shanghai Institute of Applied Physics and the Oak Ridge National Laboratory (ORNL), a research project was initiated to aid in the development of pumps for high-temperature salts. The objectives of the task included characterization of the behavior of an existing ORNL LSTL pump; design and test a modified impeller and volute for improved pump characteristics; and finally, provide lessons learned, recommendations, and guidelines for salt pump development and design. The pump included on the liquid salt test loop (LSTL) at ORNL served as a case study. This report summarizes the progress to date. The report is organized as follows. First, there is a review, focused on pumps, of the significant amount of work on salts at ORNL during the 1950s 1970s. The existing pump on the LSTL is then described. Plans for hot and cold testing of the pump are then discussed, including the design for a cold shakedown test stand and the required LSTL modifications for hot testing. Initial hydraulic and vibration modeling of the LSTL pump is documented. Later, test data from the LSTL will be used to validate the modeling approaches, which could then be used for future pump design efforts. Some initial insights and test data from the pump are then provided. Finally, some preliminary design goals and requirements for a future LSTL pump are provided as examples of salt pump design considerations.

  19. Iodine species in reactor effluents and in the environment. Final report

    International Nuclear Information System (INIS)

    Voilleque, P.G.

    1979-12-01

    The air-to-vegetation transfer velocity, an important parameter in evaluating dose to man from I-131 in the environment, has been used to establish an environmental reactivity ranking for the radioiodine species. In these terms, the hierarchy of reactivities is I 2 > particulate > HOI > organic iodides. Previous measurements showing that the least reactive forms, HOI and organic iodides, comprise 50% or more of the total I-131 release from boiling water reactors were confirmed in this study. Measurements of the chemical forms of I-131 and stable iodine in the environment were made using radioiodine species sampler media and a stable iodine species sampler. The compound HOI-131 was observed in the environment 1.5 km from the plant release point. Organic forms were prevalent in environmental samples. Organic iodide residence times calculated using the measured variability of ambient concentrations are > 80 days

  20. Examination of the use of the breeder reactor to enhance security of energy supply. Final report

    International Nuclear Information System (INIS)

    Jones, G.S.; Chow, B.G.

    1980-01-01

    Early commercialization of the breeder reactor has been justified in terms of its promise for lessening a nation's dependence on imported uranium. The relationship between breeder timing (assuming a vigorous but not instantaneous breeder penetration rate) and the reduction in uranium requirements is examined using Japan as an example. Comparing early and delayed breeder introductions (2006 and 2026 respectively in the high growth projections), the difference in cumulative uranium requirements is slight through the year 2045; for both scenarious, annual requirements remain high. The early breeder introduction strategy is more attractive when it incorporates a 30 percent improved LWR. However, given the unlikelihood of sufficient funding for the development of both designs, the optimal strategy for Japan is stockpiling uranium for an improved LWR and delaying introduction of the breeder

  1. Seismic safety review mission Almaty WWR 10 MW research reactor Almaty, Kazakhstan. Final report

    International Nuclear Information System (INIS)

    Gurpinar, A.; Slemmons, D.B.; David, M.; Masopust, R.

    1995-06-01

    On the request of the government of Kazakhstan and within the scope of the TC project KAZ/0/004, a seismic safety review mission was conducted in Almaty, 8-19 May 1995 for the WWR 10 Mw research reactor. This review followed the fact finding mission which visited Almaty in November 1993 together with an INSARR mission. At that time some information regarding the seismotectonic setting of the site as well as the seismic capacity of the facility was obtained. This document presents the results of further work carried out on both the issues. It discusses technical session findings on geology, seismology, structures and equipments. In the end conclusions and recommendations of the mission are given. 4 refs, figs, tabs, 18 photos

  2. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    Energy Technology Data Exchange (ETDEWEB)

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  3. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    International Nuclear Information System (INIS)

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding 233 U from 232 Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program

  4. Stress corrosion cracking studies of reactor pressure vessel steels. Final report

    International Nuclear Information System (INIS)

    Van Der Sluys, W.A.

    1996-10-01

    The objective of this project was to perform a critical review of the information available in open literature on stress corrosion cracking of reactor pressure vessel materials in simulated light-water-reactor (LWR) conditions, develop a test procedure for conducting stress corrosion crack growth experiments in simulated LWR environments, and conduct a test program in an effort to duplicate some of the data available from the literature. The authors concluded that stress corrosion crack growth has been observed in pressure vessel steels under laboratory test conditions. The composition of the water in most cases where growth was observed is outside of the composition specified for operating conditions. Crack growth was observed in the experiments performed in this program, and it was intermittent. The cracking would start and stop for no apparent reason. In most instances, it would not restart without the change of some external variable. In a few instances, it restarted on its own. Crack growth rates as high as 3.6 x 10 -9 m/sec were observed in pressure vessel steels in high-purity water with 8 ppm oxygen. These high crack growth rates were observed for extremely short bursts in crack extension. They could not be sustained for crack growth extensions greater than a few tenths of a millimeter. From the results of this project it appears highly unlikely that stress corrosion cracking will be observed in operating nuclear plants where the coolant composition is maintained within water chemistry guidelines. However, more work is needed to better define the contaminations that cause crack growth. The crack growth rates are so high and the threshold values for crack nucleation are so low that the conditions causing them need to be well defined and avoided

  5. Final environmental statement, Liquid Metal Fast Breeder Reactor Program. Volume 3

    International Nuclear Information System (INIS)

    1975-12-01

    Included are copies of thirty-four comment letters on the Proposed Final Environmental Statement together with the ERDA replies to these letters. The letters were received from Federal, State, and local agencies, environmental and public interest groups, members of the academic and industrial communities, and individual citizens

  6. Final environmental statement, Liquid Metal Fast Breeder Reactor Program. Volume 2

    International Nuclear Information System (INIS)

    1975-12-01

    Included are copies of fifty-six comment letters on the Proposed Final Environmental Statement together with the ERDA replies to these letters. The letters were received from Federal, State, and local agencies, environmental and public interest groups, members of the academic and industrial communities, and individual citizens

  7. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Somers, J.; Van Den Durpel, L.

    2013-01-01

    The PUMA project - the acronym stands for “Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors” - was a Specific Targeted Research Project (STREP) within the Euratom 6th Framework (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO2-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR designs and their ability to accept a variety

  8. Final Environmental Statement related to license renewal and power increase for the National Bureau of Standards Reactor: Docket No. 50-184

    International Nuclear Information System (INIS)

    1982-08-01

    This Final Environmental Statement contains an assessment of the environmental impact associated with renewal of Operating License No. TR-5 for the National Bureau of Standards (NBS) reactor for a period of 20 years at a power level of 20 MW. This reactor is located on the 576-acre NBS site near Gaithersburg in Montgomery County, Maryland, about 20 mi northwest of the center of Washington, DC. The reactor is a high-flux heavy-water-moderated, cooled and reflected test reactor, which first went critical on December 7, 1967. Though the reactor was originally designed for 20-MW operation, it has been operating for 14 years at a maximum authorized power level to 10 MW. Program demand is now great enough to warrant operation at a power level of 20 MW. No additional major changes to the physical plant are required to operate at 20 MW

  9. Final Report, Nuclear Energy Research Initiative (NERI) Project: An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model

    International Nuclear Information System (INIS)

    Anistratov, Dmitriy Y.; Adams, Marvin L.; Palmer, Todd S.; Smith, Kord S.; Clarno, Kevin; Hikaru Hiruta; Razvan Nes

    2003-01-01

    OAK (B204) Final Report, NERI Project: ''An Innovative Reactor Analysis Methodology Based on a Quasidiffusion Nodal Core Model'' The present generation of reactor analysis methods uses few-group nodal diffusion approximations to calculate full-core eigenvalues and power distributions. The cross sections, diffusion coefficients, and discontinuity factors (collectively called ''group constants'') in the nodal diffusion equations are parameterized as functions of many variables, ranging from the obvious (temperature, boron concentration, etc.) to the more obscure (spectral index, moderator temperature history, etc.). These group constants, and their variations as functions of the many variables, are calculated by assembly-level transport codes. The current methodology has two main weaknesses that this project addressed. The first weakness is the diffusion approximation in the full-core calculation; this can be significantly inaccurate at interfaces between different assemblies. This project used the nodal diffusion framework to implement nodal quasidiffusion equations, which can capture transport effects to an arbitrary degree of accuracy. The second weakness is in the parameterization of the group constants; current models do not always perform well, especially at interfaces between unlike assemblies. The project developed a theoretical foundation for parameterization and homogenization models and used that theory to devise improved models. The new models were extended to tabulate information that the nodal quasidiffusion equations can use to capture transport effects in full-core calculations

  10. Decommissioning of eight surplus production reactors at the Hanford Site, Richland, Washington. Addendum (Final Environmental Impact Statement)

    Energy Technology Data Exchange (ETDEWEB)

    1992-12-01

    The first section of this volume summarizes the content of the draft environmental impact statement (DEIS) and this Addendum, which together constitute the final environmental impact statement (FEIS) prepared on the decommissioning of eight surplus plutonium production reactors at Hanford. The FEIS consists of two volumes. The first volume is the DEIS as written. The second volume (this Addendum) consists of a summary; Chapter 9, which contains comments on the DEIS and provides DOE`s responses to the comments; Appendix F, which provides additional health effects information; Appendix K, which contains costs of decommissioning in 1990 dollars; Appendix L, which contains additional graphite leaching data; Appendix M, which contains a discussion of accident scenarios; Appendix N, which contains errata; and Appendix 0, which contains reproductions of the letters, transcripts, and exhibits that constitute the record for the public comment period.

  11. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels (Final Report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J. H.; Lee, B. S.; Chi, S. H.; Kim, J. H.; Oh, Y. J.; Yoon, J. H.; Kwon, S. C.; Park, D. G.; Kang, Y. H.; Choo, K. N.; Oh, J. M.; Park, S. J.; Kim, B. K.; Shin, Y. T.; Cho, M. S.; Sohn, J. M.; Kim, D. S.; Choo, Y. S.; Ahn, S. B.; Oh, W. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-05-01

    Reactor pressure vessel materials, which were produced by a domestic company, Doosan Heavy Industries and construction Co., Ltd., have been evaluated using the neutron irradiation facility HANARO. For this evaluation, instrumented capsules were used for neutron irradiation of various kinds of specimens made of different heats of steels, which are VCD(Y4), VCD+Al(U4), Si+Al(Y5), U4 weld metal, and U4 HAZ, respectively. The fast neutron fluence levels ranged 1 to 5 (x10{sup 19} n/cm{sup 2}, E>1MeV) depending on the specimens and the irradiation temperature was controlled within 290{+-}10 deg C. The test results showed that, in the ranking of the material properties of the base metals, both before and after neutron irradiation, Y5 is the best, U4 the next and Y4 the last. Y4 showed a substantial change by neutron irradiation as well as the properties was worse than others in the unirradiated state. However, Y5, which showed the best properties in unirradiated state, was also the best in the resistance for irradiation embrittlement and one can hardly detect the property change after irradiation. The weldment showed a reasonably good resistance to irradiation embrittlement while the unirradiated properties were worse than base metals. The RPV steels are all expected to meet the screening criteria of the USNRC codes and regulations during the end of plant life. 39 refs., 42 figs., 27 tabs. (Author)

  12. Suspended-Bed Reactor preliminary design, 233U--232Th cycle. Final report (revised)

    International Nuclear Information System (INIS)

    Karam, R.A.; Alapour, A.; Lee, C.C.

    1977-11-01

    The preliminary design Suspended-Bed Reactor is described. Coated particles about 2 mm in diameter are used as the fuel. The coatings consist of three layers: (1) low density pyrolytic graphite, 70 μ thick, (2) silicon carbide pressure vessel, 30 μ thick, and (3) ZrC layer, 50 μ thick, to protect the pressure vessel from moisture and oxygen. The fuel kernel can be either uranium-thorium dicarbide or metal. The coated particles are suspended by helium gas (coolant) in a cluster of pressurized tubes. The upward flow of helium fluidizes the coated particles. As the flow rate increases, the bed of particles is lifted upward to the core section. The particles are restrained at the upper end of the core by a suitable screen. The overall particle density in the core is just enough for criticality condition. Should the helium flow cease, the bed in the core section will collapse, and the particles will flow downward into the section where the increased physical spacings among the tubes brings about a safe shutdown. By immersing this section of the tubes in a large graphite block to serve as a heat sink, dissipation of decay heat becomes manageable. This eliminates the need for emergency core cooling systems

  13. Improvement of reference nuclear data for commercial power reactor analysis and design. Final report

    International Nuclear Information System (INIS)

    Levitt, L.; Magurno, B.; Rose, P.

    1977-10-01

    The National Neutron Cross Section Center (NNCSC) has participated in the Electric Power Research Institute (EPRI) program of improving the nuclear data base to be used as input for the analysis of power reactors. The NNCSC contribution consisted of organizing and participating in specialists' meetings on the Thermal Neutron Cross Sections for the Actinides and Fission Products, participating in a meeting on the recommended value for the 252 Cf anti nu, and keeping abreast of new measurements on the low energy resonance parameters of 238 U. In addition, complete evaluations for seven isotopes of gadolinium have been produced for the next version of the data library. Definitive benchmark calculations using the Monte Carlo code RECAP have been carried out for two homogeneous uranium nitrate critical spheres (ORNL 1 and 2), two water moderated uranium metal lattices (TRX-1 and 2), and three plutonium nitrate spheres (PNL 2, 3, and 5). Results for the uranium spheres gave good agreement with experiments and with previous calculations. The calculation of the TRX lattice parameters, however, casts doubt on the validity of analytic leakage corrections used previously. Eigenvalues of the plutonium nitrate spheres are found to increase with increasing ratio of total 240 Pu/ 239 Pu captures

  14. Elastic stability and vibration of toroidal magnets for fusion reactors. Final report

    International Nuclear Information System (INIS)

    Moon, F.C.; Swanson, C.

    1975-09-01

    The vibration and elastic stability of a set of discrete superconducting toroidal field magnets arranged to form a ''bumpy'' torus is examined. The mutual destabilizing magnetic forces between magnet pairs are calculated using a numerical differential inductance technique. It is shown that the mutual attractive magnetic forces can produce elastic buckling of the entire toroidal set. The vibration modes of the set are also found as functions of the coil current. The response of the set of magnets to an earthquake type motion of the toroidal base is calculated. The calculations have been incorporated in a computer code which accompanies the report. Measurements are made of the lateral stiffness of a flexible, planar, superconducting coil between two rigid coils in series. These tests show a dramatic decrease in the natural bending frequency with subsequent elastic instability or ''buckling'' at a critical value of the current in the coils. These observations support a magnetoelastic analysis which shows that proposed designs, of toroidal field coils for Tokamak fusion reactors, have insufficient lateral support for mechanical stability of the magnets

  15. GoAmazon2014/15. Oxidation Flow Reactor Final Campaign Report

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, J. L. [Univ. of Colorado, Boulder, CO (United States); Day, D. A. [Univ. of Colorado, Boulder, CO (United States); Hu, W. [Univ. of Colorado, Boulder, CO (United States); Palm, B. B. [Univ. of Colorado, Boulder, CO (United States); Campuzano-Jost, P. [Univ. of Colorado, Boulder, CO (United States)

    2016-03-01

    The primary goal of the Green Ocean Amazon (GoAmazon2014/5) field campaign was to measure and mechanistically understand the formation of particle number and mass in a region affected by large tropical rainforest biogenic emissions and sometimes anthropogenic influence from a large urban center. As part of the two intensive operational periods (IOPs) and in collaboration with Pacific Northwest National Laboratory (PNNL) and Harvard, the Jimenez Group proposed to deploy a High-Resolution Time-of-Flight Aerosol Mass Spectrometer (HR-ToF-AMS), Thermal Denuder (TD), Scanning Mobility Particle Size (SMPS), two oxidation flow reactors (OFR; including supporting O3, CO/CO2/CH4, RH analyzers), and a high volume filter sampler (MCV) for the measurement of gas and aerosol chemical, physicochemical, and volatility properties. The two IOPs were conducted during the wet season (February to March, 2014) and dry season (August to October, 2014). This proposal was part of a collaborative proposal involving other university and government laboratories.

  16. Gas-cooled fast reactor fuel-cost assessment. Final report, October 1978-September 1979

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, M.L.

    1979-01-01

    This program, contracted to provide a Gas Cooled Fast Reactor (GCFR) fuel assembly fabrication cost assessment, comprised the following basic activities: establish agreement on the ground rules for cost assessment, prepare a fuel factory flow sheet, and prepare a cost assessment for fuel assembly fabrication. Two factory sizes, 250 and 25 MTHM/year, were considered for fuel assembly fabrication cost assessment. The work on this program involved utilizing GE LMFBR cost assessment and fuel factory studies experience to provide a cost assessment of GCFR fuel assembly fabrication. The recent impact of highly sensitive safety and safeguards environment policies on fuel factory containment, safety, quality assurance and safeguards costs are significantly higher than might have been expected just a few years ago. Fuel assembly fabrication costs are significant because they represent an estimated 30 to 60% of the total fuel cycle costs. In light of the relative high cost of fabrication, changes in the core and assembly design may be necessary in order to enhance the overall fuel cycle economics. Fabrication costs are based on similar operations and experience used in other fuel cycle studies. Because of extrapolation of present technology (e.g., remote fuel fabrication versus present contact fabrication) and regulatory requirements, conservative cost estimates were made.

  17. Reactor Emergency Action Level Monitor: Volume 2, REALM user's reference guide: Final report

    International Nuclear Information System (INIS)

    Touchton, R.A.

    1988-09-01

    A User Manual for the Reactor Emergency Action Level Monitor (REALM) expert system prototype is provided in this volume. REALM has been designed to provide expert assistance in the identification of a nuclear power plant emergency situation and the determination of its severity. REALM has been developed as an expert system which can provide sensor interpretation and situation assessment in a real-time processing environment. In its state of development at project completion, these capabilities are used in an off-line (i.e., stand-alone, desktop) fashion to provide emergency preparedness assistance in the areas of emergency classification training and emergency exercise scenario generation. REALM also serves a prototype and stepping-stone for the possible connection to the plant for on-line use. In order to distinguish the off-line system (now complete) from the on-line system (now moving from a research prototype to an installed system), the term ''REALM'' is used to indicate the on-line version, with users in the control room, technical support center, and the emergency operations facility, The off-line version is referred to as ''uREALM.''

  18. Conceptual studies of toroidal field magnets for the tokamak (fusion) experimental power reactor. Final report

    International Nuclear Information System (INIS)

    1976-01-01

    This report presents the results of ''Conceptual Studies of Toroidal Field Magnets for the Tokamak Experimental Power Reactor'' performed for the Energy Research and Development Administration, Oak Ridge Operations. Two conceptual coil designs are developed. One design approach to produce a specified 8 Tesla maximum field uses a novel NbTi superconductor design cooled by pool-boiling liquid helium. For a highest practicable field design, a unique NbSn 3 conductor is used with forced-flow, single-phase liquid helium cooling to achieve a 12 Tesla peak field. Fabrication requirements are also developed for these approximately 7 meter horizontal bore by 11 meter vertical bore coils. Cryostat design approaches are analyzed and a hybrid cryostat approach selected. Structural analyses are performed for approaches to support in-plane and out-of-plane loads and a structural approach selected. In addition to the conceptual design studies, cost estimates and schedules are prepared for each of the design approaches, major uncertainties and recommendations for research and development identified, and test coil size for demonstration recommended

  19. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  20. Calculational-experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating water-cooled and moderated reactor WWER-type NPPs shake table investigation at Paks NPP. Final report from 15 June 1993 - 15 June 1994

    International Nuclear Information System (INIS)

    Kaznovsky, S.

    1995-01-01

    This final report involves the calculation and experimental examination and ensuring the seismic resistance of the reactor equipment and pipelines at start up and operation of WWER type nuclear power plants. Shake table experiments performed at the Paks NPP are included. Namely the following devices of the emergency cooling system were tested: pump of low pressure; valve of low pressure; intermediate heat exchanger. The following values were determined: natural frequencies and vibration decrements and the main modes of normal vibrations for the heat exchanger