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)

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

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

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

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.

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.)

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

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

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

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

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

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.)

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

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

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

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

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

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)

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)

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

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.)

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.

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.

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.)

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

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

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.)

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.).

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)

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

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.

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

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.)

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

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.

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)

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

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

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

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

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.)

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)

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

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

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

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

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

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

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)

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

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.

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

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

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)

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)

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

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)

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

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.

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

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)

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)

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

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)

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

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

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

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)

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

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.

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

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

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

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.

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)

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)

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

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.

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.

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...

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

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

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

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.)

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

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)

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.)

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

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

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

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

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)

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)

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)

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

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

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)

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

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)

Control of reactor coolant flow path during reactor decay heat removal

Science.gov (United States)

Hunsbedt, Anstein N.

1988-01-01

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

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

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)

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

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

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

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)

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

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.

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

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)

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)

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).

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

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

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)

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

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

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)

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

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)

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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.

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.

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.

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

Method and apparatus for protection of pump systems

International Nuclear Information System (INIS)

Youngborg, L.H.

1987-01-01

This patent describes a nuclear power plant having a fluid-filler reactor vessel with a vapor outflow line for removing vapor from the reactor vessel, and liquid inflow means for injecting liquid to the reactor vessel. The inflow means includes an inflow line, a centrifugal pump disposed along the inflow line having an inlet and an outlet, an induction motor to drive the pump, flow control means along the inflow line between the pump and the reactor vessel from the pump. A means is included for generating a first control signal in response to liquid level in the reactor vessel and net vapor outflow versus liquid inflow with respect to the reactor vessel, the first control signal generating means being effective to generate a first signal to open and a second signal to close the flow control means to maintain liquid level in the vessel within predetermined limits. A pump and pump motor protection apparatus is described comprising: means for measuring the pressure of the liquid in the inlet of the pump; means for measuring the temperature of the liquid in the inlet of the pump; means for determining a required subcooling for the pump at the instantaneous temperature of the liquid in the inlet of the pump; and means for determining the enthalpy of the liquid in the inlet of the pump from the pressure and temperature of the liquid

Detection of cavitation inception by acoustic technique in centrifugal pumps for nuclear application

International Nuclear Information System (INIS)

Prakash, V.; Prabhakar, R.; Rao, A.S.L.K.; Kale, R.D.

1994-01-01

The primary centrifugal pumps in a pool type reactor like the proposed Prototype Fast Breeder Reactor (PFBR) are required to operate at low values of available net positive suction head due to the limited submergence available in the pool. Pump hydraulics are designed to ensure that there is no cavitation or only minimum cavitation in the pump impeller in order to minimise long term erosion damage. Rigorous cavitation tests are usually carried out during development and final testing phase and a promising cavitation detection technique lies in acoustic noise measurements on the pump. As part of PFBR pump development programme, cavitation noise measurements were initially carried out on an experimental sodium pump in a water rig to establish detection procedures. Recently cavitation noise measurements were carried out on a 1/3 scale model impeller of PFBR pump along with visual observation of impeller passages to establish a correlation between visual and acoustic technique. Accelerometer responding to structure borne noise seems to give the best result. (author). 4 refs., 6 figs

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

Feasibility study on the type of KALIMER coolant circulation pump

International Nuclear Information System (INIS)

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

1997-07-01

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

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)

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

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.

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

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

The effect of pump cavitation on the design of the primary pumps for C.F.R

International Nuclear Information System (INIS)

Worster, R.C.

1976-01-01

In the design appraisal of the sodium pumps for the primary circuit of the proposed 1300 MW(e) CFR it has been recognised that cavitation, its effects and its control, is the outstanding hydraulic design problem. Careful consideration of this problem and of the possible effects of pump cavitation on the performance of other reactor systems has led to the conclusion that it is more prudent at present to specify pumps with zero cavitation at normal full speed operating conditions. Under abnormal operation it may be necessary to reduce the pumps' speed to prevent cavitation in the pumps or associated equipment. The principal reasons for this decision were uncertainties concerning the possibility of erosion due to limited cavitation in sodium and the possibility of pump cavitation noise interfering with acoustic detection of malfunctioning of reactor components or of boiling in the reactor core

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

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.)

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.

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

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

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)

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)

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

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.

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

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

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

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

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

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

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)

Residual heat removal pump retrofit program

International Nuclear Information System (INIS)

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

1990-01-01

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

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.

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.

Pump Coastdown with the Submerged Flywheel

Energy Technology Data Exchange (ETDEWEB)

Yoon, Hyun-Gi; Seo, KyoungWoo; Kim, Seong Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Many research reactors are generally designed as open pool types in consideration of the heat removal of the nuclear fuels, reactor operation and accessibility. Reactor structure assembly is generally placed at the pool bottom as shown in Fig. 1. Primary cooling system pump circulates the coolant from the reactor structure to the heat exchanger in order to continuously remove the heat generated from the reactor core in the research reactor as shown in Fig. 1. The secondary cooling system releases the transferred heat to the atmosphere by the cooling tower. Coastdown flow rate of the primary cooling system pump with the submerged flywheel are calculated analytically in case of the accident situation. Coastdown flow rate is maintained until almost 80 sec when the pump stops normally. But, coastdown flow rate is rapidly decreased when the flywheel is submerged because of the friction load on the flywheel surface.

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)

Reactor system

International Nuclear Information System (INIS)

Miyano, Hiroshi; Narabayashi, Naoshi.

1990-01-01

The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

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

Magnetohydrodynamic generator and pump system

International Nuclear Information System (INIS)

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

1983-01-01

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

Preliminary design of a tandem mirror reactor

International Nuclear Information System (INIS)

Strohmayer, J.N.

1984-04-01

The purpose of this thesis is to examine the TARA mirror experiment as a possible tandem mirror reactor configuration. This is a preliminary study to size the coil structure based on using the smallest end cell axial length that physics and engineering allow, zeroing the central cell parallel currents and having interchange stability. The input powers are estimated for the final reactor design so a Q value may be estimated. The Q value is defined as the fusion power divided by the total injected power absorbed by the plasma. A computer study was performed on the effect of the transition size, the transition vertical spacing and transition current. These parameters affect the central cell parallel currents, the recircularization of the flux tube and the ratio of central cell beta to anchor beta needed for marginal stability. Two designs were identified. The first uses 100 keV and 13 keV neutral beams to pump the ions that trap in the thermal barrier. The Q value of this reactor is 11.3. The second reactor uses a pump beam at 40 keV. This energy is chosen because there is a resonance for the charge exchange cross section between D 0 and He 2+ at this energy, thus the alpha ash will be pumped along with the deuterium and tritium. The Q value of this reactor is 11.6

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)

Final hazard classification for the 183-C D ampersand D project

International Nuclear Information System (INIS)

Zimmer, J.J.

1995-01-01

The intent of this document is to provide a final Hazard Classification for the Decontamination and Decommissioning (D ampersand D) activities associated with the 183-C Filter Plant/Pump Room facility. The Hazard Classification was determined based upon DOE-EM-STD-5502-94, ''DOE Limited Standard, Hazard Baseline Documentation,'' issued by the US Department of Energy. The 183-C Filter Plant/Pump Room facility was constructed to support operations of the 105-B and 105-C Reactors at the Hanford Site. Since shutdown of the 105-C Reactor in April 1969, the 183-C facility has been kept in a safe storage condition

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)

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.

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

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.

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

Coolant clean-up system in nuclear reactor

International Nuclear Information System (INIS)

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

1987-01-01

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

Nuclear power plant safety related pump issues

Energy Technology Data Exchange (ETDEWEB)

Colaccino, J.

1996-12-01

This paper summarizes of a number of pump issues raised since the Third NRC/ASME Symposium on Valve and Pump Testing in 1994. General issues discussed include revision of NRC Inspection Procedure 73756, issuance of NRC Information Notice 95-08 on ultrasonic flow meter uncertainties, relief requests for tests that are determined by the licensee to be impractical, and items in the ASME OM-1995 Code, Subsection ISTB, for pumps. The paper also discusses current pump vibration issues encountered in relief requests and plant inspections - which include smooth running pumps, absolute vibration limits, and vertical centrifugal pump vibration measurement requirements. Two pump scope issues involving boiling water reactor waterlog and reactor core isolation cooling pumps are also discussed. Where appropriate, NRC guidance is discussed.

Nuclear power plant safety related pump issues

International Nuclear Information System (INIS)

Colaccino, J.

1996-01-01

This paper summarizes of a number of pump issues raised since the Third NRC/ASME Symposium on Valve and Pump Testing in 1994. General issues discussed include revision of NRC Inspection Procedure 73756, issuance of NRC Information Notice 95-08 on ultrasonic flow meter uncertainties, relief requests for tests that are determined by the licensee to be impractical, and items in the ASME OM-1995 Code, Subsection ISTB, for pumps. The paper also discusses current pump vibration issues encountered in relief requests and plant inspections - which include smooth running pumps, absolute vibration limits, and vertical centrifugal pump vibration measurement requirements. Two pump scope issues involving boiling water reactor waterlog and reactor core isolation cooling pumps are also discussed. Where appropriate, NRC guidance is discussed

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.

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

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

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.

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

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

Development of ANC-type empirical two-phase pump model for full size CANDU primary heat transport pump

International Nuclear Information System (INIS)

Chan, A.M.C.; Huynh, H.M.

2004-01-01

The development of an ANC-type empirical two-phase pump model for CANDU (Canadian Deuterium) reactor primary heat transport pumps is described in the present paper. The model was developed based on Ontario Hydro Technologies' full scale Darlington pump first quadrant test data. The functional form of the ANC model which is widely used was chosen to facilitate the implementation of the model into existing computer codes. The work is part of a bigger test program with the aims: (1) to produce high quality pump performance data under off-normal operating conditions using both full-size and model scale pumps; (2) to advance our basic understanding of the dominant mechanisms affecting pump performance based on more detailed local measurements; and (3) to develop a 'best-estimate' or improved pump model for use in reactor licensing and safety analyses. (author)

Sodium-cooled nuclear reactor

International Nuclear Information System (INIS)

Hammers, H.W.

1982-01-01

The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

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

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

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.

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.)

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

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)

Nuclear power/water pumping-up composite power plant

International Nuclear Information System (INIS)

Okamura, Kiyoshi.

1995-01-01

In a nuclear power/water pumping-up composite power plant, a reversible pump for pumping-up power generation connected to a steam turbine is connected to an upper water reservoir and a lower water reservoir. A pumping-up steam turbine for driving the turbine power generator, a hydraulic pump for driving water power generator by water flowing from the upper water reservoir and a steam turbine for driving the pumping-up pump by steams from a nuclear reactor are disposed. When power demand is small during night, the steam turbine is rotated by steams of the reactor, to pump up the water in the lower water reservoir to the upper water reservoir by the reversible pump. Upon peak of power demand during day time, power is generated by the steams of the reactor, as well as the reversible pump is rotated by the flowing water from the upper water reservoir to conduct hydraulic power generation. Alternatively, hydraulic power generation is conducted by flowing water from the upper reservoir. Since the number of energy conversion steps in the combination of nuclear power generation and pumping-up power generation is reduced, energy loss is reduced and utilization efficiency can be improved. (N.H.)

Molten salt reactors. Synthesis of studies realized between 1973 and 1983. Experimental loop file

International Nuclear Information System (INIS)

1983-03-01

Four test loops were developed for the experimental study of a molten salt reactor with lead salt direct contact. A molten salt loop, completely in graphite, including the pump, showed that this material is convenient for salt containment and circulation. Reactor components like flowmeters, electromagnetic pumps, pressure gauge, valves developed for liquid sodium, were tested with liquid lead. A water-mercury loop was built for lead-molten salt simulation studies. Finally a lead-salt loop (COMPARSE) was built to study the behaviour of salt particles carried by lead in the heat exchanger. [fr

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

Study on cavitation in centrifugal sodium pumps for FBTR and PFBR

International Nuclear Information System (INIS)

Rao, A.S.L.K.; Prabhakar, R.; Prakash, V.; Paranjpe, S.R.

2002-01-01

Fast Breeder Test Reactor (FBTR) which is expected to become critical shortly is a loop type reactor of 40 MW thermal capacity and has two primary and two secondary centrifugal pumps for heat removal. During the initial periods of reactor operation, the steam generator is bypassed and the secondary sodium pumps are required to operate at flows less than that at best efficiency point. This paper deals with the cavitation problems associated with operation at partial f lows, theoretical estimations and experimental cavitation measurements carried out on FBTR secondary sodium pumps. These investigations revealed that operation of FBTR pumps at this off design condition is free from cavitation damage. Cavitation experiments on a model pump for the development of large sodium pumps for a 500 MWe Prototype Fast Breeder Reactor (PFBR) are described in this paper

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

Vacuum problems of thermonuclear reactor design

International Nuclear Information System (INIS)

Paty, L.

1981-01-01

A thermonuclear reactor can be considered to be a vacuum system in which constant concentration should be maintained of reacting particles while permanently discharging the undesirable particles using a system of pumps. The discharging proceeds in two stages: in the former, the reactor is degassed using external pumps connected to the reactor chamber through a pumping pipe. The latter in which hydrogen is admitted, uses high pump-rate machines based on the principle of the binding of the gas to the pump surface and must not introduce molecules of higher atomic mass in the system. Turbomolecular pumps of diffusion oil pumps are most suitable for the former stage while condensation, cryosorption, titanium pumping machines and special pumping methods are most suitable for the latter stage. Examples are shown of the pump system design for Tokamak 10 and for facilities at the Euratom laboratory in Fontenay-aux-Roses. (M.D.)

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).

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)

Feed pumps for a NPP

International Nuclear Information System (INIS)

Rzhebaev, Eh.E.; Zhukov, V.M.; Evtushenko, A.A.

1977-01-01

Given is a brief description of results of designing and testing the PE850-65 and SPE1650-75 pilot feed pumps for 440 MWe pressurized water reactors and for 1000 MWe boiling water reactors. Described is a bench-mark facility for the SPE1650-75 pump with full-scale parameters. The adjustment results of the preconnected axial wheel of the 1PE1650-75 pump during the service life tests of a pilot pump have been given. Practical results on new methods of the preconnected axial wheel protection have been obtained. The design criteria developed provide for the long life time within the proven range of circular velocities. Confirmed is a possibility of carrying out advanced tests for estimating the cavitation destruction intensity using the vibration due to cavitation

LWR type reactor

International Nuclear Information System (INIS)

Kato, Kiyoshi.

1993-01-01

A water injection tank in an emergency reactor core cooling system is disposed at a position above a reactor pressure vessel. A liquid phase portion of the water injection tank and an inlet plenum portion in the reactor pressure vessel are connected by a water injection pipe. A gas phase portion of the water injection tank and an upper portion in the reactor pressure vessel are connected by a gas ventilation pipe. Hydraulic operation valves are disposed in the midway of the water injection pipe and the gas ventilation pipe respectively. A pressure conduit is disposed for connecting a discharge port of a main recycling pump and the hydraulic operation valve. In a case where primary coolants are not sent to the main recycling pump by lowering of a liquid level due to loss of coolants or in a case where the main recycling pump is stopped by electric power stoppage or occurrence of troubles, the discharge pressure of the main recycling pump is lowered. Then, the hydraulic operation valve is opened to release the flow channel, then, boric acid water in the water injection tank is sent into the reactor by a falling head, to lead the reactor to a scram state. (I.N.)

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

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.

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

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.)

Transient behaviour of main coolant pump in nuclear power plants

International Nuclear Information System (INIS)

Delja, A.

1986-01-01

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

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

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

International Nuclear Information System (INIS)

Veronesi, Luciano; Francis, W.R.

1978-01-01

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

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

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.)

Nuclear reactor recyclation device

International Nuclear Information System (INIS)

Takigawa, Yukio; Chuma, Kazuto

1987-01-01

Purpose: To prevent the unevenness for the coolant flow rate even when abnormality occurs to one of recycling pumps. Constitution: A plurality of jet pumps disposed at an interval around the reactor core are divided circumferentially into two sets, and a pipeway is disposed to the outside of each pair including recycling pumps corresponding to each of the sets. The pipeway is connected to the recycling inlet of the jet pump by way of a manifold. The discharge portion of the recycling pumps of the loop pipeway are connected with each other by way of communication pipes, and a normally closed valve is disposed to the communication pipe and the normally closed valve of the communication pipe is opened upon detecting abnormality for one of the recycling pumps. Thus, if either one of the pair of recycling pumps shows abnormal state, coolants flows from the other of pipeway to the outside of the loop pipeway and coolants are supplied from all the jet pumps to the reactor core portion and, accordingly, the not-uniform flow rate can be prevented to eliminate undesired effect on the reactor core. (Kamimura, M.)

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

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

Terrestrial and extraterrestrial superresonators as drivers for an inertial confinement fusion reactor

International Nuclear Information System (INIS)

Seifritz, W.; Vath, W.

1992-01-01

This paper reports on the recirculating power fraction of a laser-driven inertial confinement fusion (ICF) reactor which can be reduced by using laser diodes to pump a neodymium solid-state laser. To overcome the high costs of two-dimensional arrays of laser diodes, two types of superresonators are proposed: a terrestrially based one and an extraterrestrially based one on a geostationary orbit. Both are designed in such a way that a sequence of short laser pulses (10 to 20 ns wide), each with an energy of 5 to 10 MJ and a frequency of 10 Hz, are produced to trigger a deuterium-tritium ICF reactor. The terrestrial superresonator needs a much smaller number of two-dimensional laser diode arrays than a conventionally pumped once-through solid-state laser system, and the extraterrestrial resonator is pumped by means of concentrated solar radiation. In practice, at least an order of magnitude fewer laser diodes and crystalline calcium fluoride gain media are needed to meet the requirements of a laser driver for an ICF reactor. If, finally, a liquid neodymium laser system could be used for an ICF reactor, the cooling of the gain slabs would be facilitated substantially

Analyses of Decrease in Reactor Coolant Flow Rate in SMART

International Nuclear Information System (INIS)

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

2011-01-01

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

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

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

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

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Nuclear reactor apparatus

International Nuclear Information System (INIS)

Braun, H.E.; Bonnet, H.P.

1978-01-01

The reactor and its containment, instead of being supported on a solid concrete pad, are supported on a truss formed of upper and lower reinforced horizontal plates and vertical walls integrated into a rigid structure. The plates and walls from chambers within which the auxiliary components of the reactor, such as valves, pumping equipment and various tanks, are disposed. Certain of the chambers are also access passages for personnel, pipe chases, valve chambers and the like. In particular the truss includes an annular chamber. This chamber is lined and sealed by a corrosion-resistant liner and contains coolant and serves as a refueling cooling storage tank. This tank is directly below the primary-coolant conductor loops which extend from the reactor above the upper plate. The upper plate includes a sump connected to the tank through which coolant flows into the tank in the event of the occurrence of a loss-of-coolant accident. The truss extends beyond the containment and has chambers in the extending annulus. Pumps for circulating the coolant between the refueling coolant storage tank and the reactor are provided in certain of these chambers. The pumps are connected to the reactor by relatively short coolant conductors. Access to these pumps is readily afforded through hatches in the extending annulus

Condition monitoring of main coolant pumps, Dhruva

International Nuclear Information System (INIS)

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

2002-01-01

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

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)

Safety features of TR-2 reactor

International Nuclear Information System (INIS)

Tuerker, T.

2001-01-01

TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

Fast-acting nuclear reactor control device

International Nuclear Information System (INIS)

Kotlyar, O.M.; West, P.B.

1993-01-01

A fast-acting nuclear reactor control device is described for controlling a safety control rod within the core of a nuclear reactor, the reactor controlled by a reactor control system, the device comprising: a safety control rod drive shaft and an electromagnetic clutch co-axial with the drive shaft operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core during reactor operation, the safety rod being oriented in a substantially vertical position to allow the rod to fall into the reactor core under the influence of gravity during shutdown of the reactor; the safety control rod drive shaft further operatively connected to a hydraulic pump such that operation of the drive shaft simultaneously drives and positions the safety control rod and operates the hydraulic pump such that a hydraulic fluid is forced into an accumulator, filling the accumulator with oil for the storage and supply of primary potential energy for safety control rod insertion such that the release of potential energy in the accumulator causes hydraulic fluid to flow through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core; a solenoid valve interposed between the hydraulic pump and the accumulator, said solenoid valve being a normally open valve, actuated to close when the safety control rod is out of the reactor during reactor operation; and further wherein said solenoid opens in response to a signal from the reactor control system calling for shutdown of the reactor and rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid releases the potential energy in the accumulator to place the safety control rod in a safe shutdown position

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

Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

Energy Technology Data Exchange (ETDEWEB)

Otra Otra, F; Leira Rey, G

1971-07-01

In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

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)

Pumps for nuclear power stations

International Nuclear Information System (INIS)

Ogura, Shiro

1979-01-01

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

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)

Recycling systems for BWR type reactors

International Nuclear Information System (INIS)

Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

1986-01-01

Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

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

Alternative water injection device to reactor equipment facility

International Nuclear Information System (INIS)

Yamashita, Masahiro.

1995-01-01

The device of the present invention injects water to the reactor and the reactor container continuously for a long period of time for preventing occurrence of a severe accident in a BWR type reactor and maintaining the integrity of the reactor container even if the accident should occur. Namely, diesel-driven pumps disposed near heat exchangers of a reactor after-heat removing system (RHR) are operated before the reactor is damaged by the after heat to cause reactor melting. A sucking valve disposed to a pump sucking pipeline connecting a secondary pipeline of the RHR heat exchanger and the diesel driving pump is opened. A discharge valve disposed to a pump discharge pipeline connecting a primary pipeline of the RHR heat exchanger and the diesel driving pump is opened. With such procedures, sea water is introduced from a sea water taking port through the top end of the secondary pipeline of the RHR heat exchanger and water is injected into the inside of the pressure vessel or the reactor container by way of the primary pipeline of the RHR heat exchanger. As a result, the reactor core is prevented from melting even upon occurrence of a severe accident. (I.S.)

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

Emergency reactor cooling device

International Nuclear Information System (INIS)

Arakawa, Ken.

1993-01-01

An emergency nuclear reactor cooling device comprises a water reservoir, emergency core cooling water pipelines having one end connected to a water feeding sparger, fire extinguishing facility pipelines, cooling water pressurizing pumps, a diesel driving machine for driving the pumps and a battery. In a water reservoir, cooling water is stored by an amount required for cooling the reactor upon emergency and for fire extinguishing, and fire extinguishing facility pipelines connecting the water reservoir and the fire extinguishing facility are in communication with the emergency core cooling water pipelines connected to the water feeding sparger by system connection pipelines. Pumps are operated by a diesel power generator to introduce cooling water from the reservoir to the emergency core cooling water pipelines. Then, even in a case where AC electric power source is entirely lost and the emergency core cooling system can not be used, the diesel driving machine is operated using an exclusive battery, thereby enabling to inject cooling water from the water reservoir to a reactor pressure vessel and a reactor container by the diesel drive pump. (N.H.)

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

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)

Reactor cooling apparatus

International Nuclear Information System (INIS)

Ogura, Kenji.

1983-01-01

Purpose: To increase natural convection flowrate in the reactor core upon interruption of a recycling pump by remarkably decreasing the flow resistance. Constitution: By-pass lines are disposed to a recycling pump in a primary coolant system and a second recycling pump in a secondary coolant system respectively, and a check valve and an isolation valve are attached to each of them. Each of the isolation valves is closed during normal operation and automatically opened when the number of rotation for each of the recycling pumps goes lower than a predetermined value. This can significantly decrease the flow resistance in the primary and secondary coolant systems upon interruption of the recycling pumps due to the entire loss of AC power source or the like to thereby increase the natural convection flowrate in the reactor core. (Sekiya, K.)

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)

Part-load pumping operation, control and behaviour

International Nuclear Information System (INIS)

1988-01-01

Twenty one papers are printed. Their main concern is with pump performance under abnormal operating conditions or when operating at part-load. The effect of part-load operation on pump performance, pump internal flow and pump cavitation and noise were considered. The pumps considered are used in a variety of situations and some case studies were discussed. One paper about part-load operation of the boiler pumps for French pressurized water reactors and one paper on pressure pulsations of centrifugal pumps at very low flowrate, are indexed separately. (U.K.)

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

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)

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)

Speed control device for coolant recycling pump

International Nuclear Information System (INIS)

Kageyama, Takao.

1992-01-01

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

FBR type reactors

International Nuclear Information System (INIS)

Suzuoki, Akira; Yamakawa, Masanori.

1985-01-01

Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

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)

Nuclear reactor auxiliary heat removal system

International Nuclear Information System (INIS)

Thompson, R.E.; Pierce, B.L.

1977-01-01

An auxiliary heat removal system to remove residual heat from gas-cooled nuclear reactors is described. The reactor coolant is expanded through a turbine, cooled in a heat exchanger and compressed by a compressor before reentering the reactor coolant. The turbine powers both the compressor and the pump which pumps a second fluid through the heat exchanger to cool the reactor coolant. A pneumatic starter is utilized to start the turbine, thereby making the auxiliary heat removal system independent of external power sources

Development of advanced boiling water reactor for medium capacity

International Nuclear Information System (INIS)

Kazuo Hisajima; Yutaka Asanuma

2005-01-01

This paper describes a result of development of an Advanced Boiling Water Reactor for medium capacity. 1000 MWe was selected as the reference. The features of the current Advanced Boiling Water Reactors, such as a Reactor Internal Pump, a Fine Motion Control Rod Drive, a Reinforced Concrete Containment Vessel, and three-divisionalized Emergency Core Cooling System are maintained. In addition, optimization for 1000 MWe has been investigated. Reduction in thermal power and application of the latest fuel reduced the number of fuel assemblies, Control Rods and Control Rod Drives, Reactor Internal Pumps, and Safety Relief Valves. The number of Main Steam lines was reduced from four to two. As for the engineered safety features, the Flammability Control System was removed. Special efforts were made to realize a compact Turbine Building, such as application of an in line Moisture Separator, reduction in the number of pumps in the Condensate and Feedwater System, and change from a Turbine-Driven Reactor Feedwater Pump to a Motor-Driven Reactor Feedwater Pump. 31% reduction in the volume of the Turbine Building is expected in comparison with the current Advanced Boiling Water Reactors. (authors)

CFD results for temperature dependence water cooling pump NPSH calculations - 15425

International Nuclear Information System (INIS)

Strongin, M.P.

2015-01-01

In this work the possibility to model the pump for water cooling reactors behavior in the critical situation was considered for cases when water temperature suddenly increases. In cases like this, cavitation effects may cause pump shutoff and consequently stop the reactor cooling. Centrifugal pump was modeled. The calculations demonstrate strong dependence of NPSH (net-positive-suction-head) on the water temperature on the pump inlet. The water temperature on the inlet lies between 25 and 180 C. degrees. The pump head performance curve has a step-like slope below NPSH point. Therefore, if the pressure on the pump inlet is below than NPSH, it leads to the pump shutoff. For high water temperature on the pump inlet, NPSH follows the vapor saturated pressure for given temperature with some offset. The results clearly show that in case of accidental increase of temperature in the cooling loop, special measures are needed to support the pressure on the pump inlet to prevent pump shutoff. (author)

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.)

A water inner circulation device for a reactor vessel

International Nuclear Information System (INIS)

Eriksson, O.

1976-01-01

A water inner circulation device for a reactor vessel comprising a pump mounted in the reactor vessel and driven by a water-cooled electric motor mounted in a housing outside the reactor vessel, the shaft of the pump passing through the reactor-vessel bottom and being coupled to the motor shaft in a member mechanically connected to the bottom of the reactor vessel in the vicinity of the motor housing, the pump shaft being surrounded by a resilient sealing ring, the reactor vessel communicating with the cooling channels of the pump, when the latter is operating, via a slot surrounding the pump hollow cylindrical shaft, characterized in that the slot inner end is used for/forming a circular space surrounding the pump shaft and surrounded by the motorhousing, in which is coaxially mounted a separating cylindral wall, the upper edge of which is tightly applied against the inner wall of the motor-housing to which it is fastened vertically, the inner surface of said wall being turned towards the outer surface of a circular packing-box, the outer surface of said separating wall constituting a separating radical inner surface for a circular chamber through which flow the motor cooling water. (author)

Nuclear-driven flashlamp pumping of the atomic iodine laser

International Nuclear Information System (INIS)

Miley, G.H.

1992-03-01

This report is a study of the atomic iodine laser pumped with nuclear- excited XeBr fluorescence. Preliminary experiments, conducted in the TRIGA reactor investigated the fluorescence of the excimer XeBr under nuclear pumping with 10 B and 3 He, for use as a flashlamp gas to stimulate the laser. These measurements included a determination of the fluorescence efficiency (light emitted in the wavelength region of interest, divided by energy deposited in the gas) of XeBr under nuclear pumping, with varying excimer mixtures. Maximum fluorescence efficiencies were approximately 1%. In order to better understand XeBr under nuclear excitation, a kinetics model of the system was prepared. The model generated the time-dependant concentrations of 20 reaction species for three pulse sizes, a TRIGA pulse, a fast burst reactor pulse, and an e-beam pulse. The modeling results predicted fluorescence efficiencies significantly higher (peak efficiencies of approximately 10%) than recorded in the fluorescence experiments. The cause of this discrepancy was not fully determined. A ray tracing computer model was also prepared to evaluate the efficiency with which nuclear-induced fluorescence generated in one cavity of a laser could be coupled into another cavity containing an iodine lasant. Finally, an experimental laser cell was constructed to verify that nuclear-induced XeBr fluorescence could be used to stimulate a laser. Lasing was achieved at 1.31 micron in the TRIGA using C 3 F 7 I, a common iodine lasant. Peak laser powers were approximately 20 mW. Measured flashlamp pump powers at threshold agreed well with literature values, as did lasant pressure dependency on laser operation

Countermeasure against thermal fatigue crack of primary loop recirculation pump in BWR

International Nuclear Information System (INIS)

Noda, Hiroshi; Narabayashi, Tadashi; Takahashi, Yuuji

2008-01-01

The reactor water was fed to the purge water of the mechanical seal on the original design of the primary loop recirculation pump. Because the mechanical seal had a short life due to the cruds in the reactor water, the clean purge water was adopted instead of the reactor water. After this modification, the shallow cracks were found on the surface of the pump shaft and casing cover due to the temperature fluctuation between the cold purge water and the hot pump discharge water. The fundamental mechanism and countermeasure were investigated by scale test, mock-up test and so on. The flow barrier with a heater was contrived through these tests. It has been introduced gradually in operating and constructing PLR pumps after its completion in 1995. The PLR pumps are overhauled around every 10 years in Japan. The first overhaul of the PLR pumps showed no cracks around the pump shaft and casing over after 10 years' operation. This paper presents both its development process and inspection results. (author)

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

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

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.)

Mechanical pumping at low temperature

Energy Technology Data Exchange (ETDEWEB)

Perin, J.P.; Claudet, G.; Disdier, F.

1994-12-31

This new concept consists of a mechanical pump able to run at low temperature (25 K). Since gas density varies inversely with temperature, the pump could deliver much higher mass flow rate than at room temperature for a given size. Advantages of this concept are reduction of an order of magnitude in size and weight when compared to a conventional pump scaled to perform the same mass flow rate at room temperature. Results obtained at 80 K and 25 K with a Holweck type molecular drag pump of 100 mm diameter and with few stages of a turbomolecular pump running at the same temperatures, are given. This pump would be a solution to allow continuous tritium extraction and minimize the mass inventory for the ITER (International Tokamak Experiment Reactor). 5 figs., 2 tabs., 4 refs.

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

International Nuclear Information System (INIS)

Fostier, L.; Howell, D.

2005-01-01

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

Structural evaluation of IEA-R1 primary system pump nozzles

Energy Technology Data Exchange (ETDEWEB)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel, E-mail: gfainer@ipen.br, E-mail: afaloppa@ipen.br, E-mail: calberto@ipen.br, E-mail: mmattar@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-11-01

The IEA-R1 pumps of the primary coolant system may be required to withstand design and operational conditions. IEA-R1 nuclear research reactor is an open pool type reactor operated by IPEN since 1957. The reactor can be operated up to 5MW heating power since it was upgraded in a modernization program conducted by IPEN. The primary coolant system is composed by the piping system, decay tank, two heat pumps and two heat exchangers. In the latest arrangement upgrade of the primary system, conducted in 2014 as part of an aging management program, a partial replacement of the coolant piping and total replacement of piping and pump supports were done. As consequence, reviewed loads in the pump nozzles were obtained demanding a new evaluation of them. The aim of this report is to present the structural evaluation of the pump nozzles, considering the new loads coming from the new piping layout, according to: API 610 code verification, Supplier loads and structural analysis applying finite element method, by using the ANSYS computer program, regarding ASME VIII Div 1 & 2 recommendations. (author)

Structural evaluation of IEA-R1 primary system pump nozzles

International Nuclear Information System (INIS)

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

2017-01-01

The IEA-R1 pumps of the primary coolant system may be required to withstand design and operational conditions. IEA-R1 nuclear research reactor is an open pool type reactor operated by IPEN since 1957. The reactor can be operated up to 5MW heating power since it was upgraded in a modernization program conducted by IPEN. The primary coolant system is composed by the piping system, decay tank, two heat pumps and two heat exchangers. In the latest arrangement upgrade of the primary system, conducted in 2014 as part of an aging management program, a partial replacement of the coolant piping and total replacement of piping and pump supports were done. As consequence, reviewed loads in the pump nozzles were obtained demanding a new evaluation of them. The aim of this report is to present the structural evaluation of the pump nozzles, considering the new loads coming from the new piping layout, according to: API 610 code verification, Supplier loads and structural analysis applying finite element method, by using the ANSYS computer program, regarding ASME VIII Div 1 & 2 recommendations. (author)

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.

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)

PWR type reactor plant

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi.

1993-01-01

A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

Reactor water sampling device

International Nuclear Information System (INIS)

Sakamaki, Kazuo.

1992-01-01

The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

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)

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)

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

Review of magnetohydrodynamic pump applications

Directory of Open Access Journals (Sweden)

O.M. Al-Habahbeh

2016-06-01

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

Emergency recirculation pump driving mechanism

International Nuclear Information System (INIS)

Morooka, Shin-ichi.

1980-01-01

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

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

Reactor containing facility

International Nuclear Information System (INIS)

Akagawa, Katsuhiko.

1992-01-01

A cooling space having a predetermined capacity is formed between a reactor container and concrete walls. A circulation loop disposed to the outside of the concrete walls is connected to the top and the bottom of the cooling space. The circulation loop has a circulation pump and a heat exchanger, and a cooling water supply pipe is connected to the upstream of the circulation pump for introducing cooling water from the outside. Upon occurrence of loss of coolant accident, cooling water is introduced from the cooling water supply pipe to the cooling space between the reactor container and the concrete walls after shut-down of the reactor operation. Then, cooling water is circulated while being cooled by the heat exchanger, to cool the reactor container by cooling water flown in the cooling space. This can cool the reactor container in a short period of time upon occurrence of the loss of coolant accident. Accordingly, a repairing operation for a ruptured portion can be conducted rapidly. (I.N.)

Nuclear reactor with makeup water assist from residual heat removal system

International Nuclear Information System (INIS)

Schulz, T.L.; Corletti, M.M.

1994-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit by pumping water from an in-containment refueling water storage tank during staged depressurization of the coolant circuit, the final stage including passive emergency cooling by gravity feed from the refueling water storage tank to the coolant circuit and to flood the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and avoids the final stage of depressurization with its flooding of the containment when such action is not necessary, but does not prevent the final stage when it is necessary. A high pressure makeup water storage tank coupled to the reactor coolant circuit holds makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal system can also be coupled in a loop with the refueling water supply tanks for cooling the tank. (Author)

Design capability of CANDU heat transport pump shafts against cracking

International Nuclear Information System (INIS)

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

1993-01-01

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

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

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

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.

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

Reactor-core isolation cooling system with dedicated generator

International Nuclear Information System (INIS)

Nazareno, E.V.; Dillmann, C.W.

1992-01-01

This patent describes a nuclear reactor complex. It comprises a dual-phase nuclear reactor; a main turbine for converting phase-conversion energy stored by vapor into mechanical energy for driving a generator; a main generator for converting the mechanical energy into electricity; a fluid reservoir external to the reactor; a reactor core isolation cooling system with several components at least some of which require electrical power. It also comprises an auxiliary pump for pumping fluid from the reservoir into the reactor pressure vessel; an auxiliary turbine for driving the pump; control means for regulating the rotation rate of the auxiliary turbine; cooling means for cooling the control means; and an auxiliary generator coupled to the auxiliary turbine for providing at least a portion of the electrical power required by the components during a blackout condition

Reactor water clean-up device

International Nuclear Information System (INIS)

Tanaka, Koji; Egashira, Yasuo; Shimada, Fumie; Igarashi, Noboru.

1983-01-01

Purpose: To save a low temperature reactor water clean-up system indispensable so far and significantly simplify the system by carrying out the reactor water clean-up solely in a high temperature reactor water clean-up system. Constitution: The reactor water clean-up device comprises a high temperature clean-up pump and a high temperature adsorption device for inorganic adsorbents. The high temperature adsorption device is filled with amphoteric ion adsorbing inorganic adsorbents, or amphoteric ion adsorbing inorganic adsorbents and anionic adsorbing inorganic adsorbents. The reactor water clean-up device introduces reactor water by the high temperature clean-up pump through a recycling system to the high temperature adsorption device for inorganic adsorbents. Since cations such as cobalt ions and anions such as chlorine ions in the reactor water are simultaneously removed in the device, a low temperature reactor water clean-up system which has been indispensable so far can be saved to realize the significant simplification for the entire system. (Seki, T.)

Reassessment of debris ingestion effects on emergency core cooling-system pump performance

International Nuclear Information System (INIS)

Sciacca, F.W.; Rao, D.V.

2004-01-01

A study sponsored by the United States (US) Nuclear Regulatory Commission (NRC) was performed to reassess the effects of ingesting loss of coolant accident (LOCA) generated materials into emergency core cooling system (ECCS) pumps and the subsequent impact of this debris on the pumps' ability to provide long-term cooling to the reactor core. ECCS intake systems have been designed to screen out large post-LOCA debris materials. However, small-sized debris can penetrate these intake strainers or screens and reach critical pump components. Prior NRC-sponsored evaluations of possible debris and gas ingestion into ECCS pumps and attendant impacts on pump performance were performed in the early 1980's. The earlier study focused primarily on pressurised water reactor (PWR) ECCS pumps. This issue was revisited both to factor in our improved knowledge of LOCA generated debris and to address specifically both boiling water reactor (BWR) and PWR ECCS pumps. This study discusses the potential effects of ingested debris on pump seals, bearing assemblies, cyclone debris separators, and seal cooling water subsystems. This assessment included both near-term (less than one hour) and long-term (greater than one hour) effects introduced by the postulated LOCA. The work reported herein was performed during 1996-1997. (authors)

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

Design of a Mechanical NaK Pump for Fission Space Power

Science.gov (United States)

Mireles, Omar R.; Bradley, David E.; Godfroy, Thomas

2011-01-01

Alkali liquid metal cooled fission reactor concepts are under development for spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid, which has specific pumping requirements. Traditionally, electromagnetic linear induction pumps have been used to provide the required flow and pressure head conditions for NaK systems but they can be limited in performance, efficiency, and number of available vendors. The objective of the project was to develop a mechanical NaK centrifugal pump that takes advantages of technology advances not available in previous liquid metal mechanical pump designs. This paper details the design, build, and performance test of a mechanical NaK pump developed at NASA Marshall Space Flight Center. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

Alignment and operability analysis of a vertical sodium pump

International Nuclear Information System (INIS)

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

1981-01-01

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

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

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

A review of pumped energy storage schemes

International Nuclear Information System (INIS)

Unsworth, G.N.

1975-07-01

The comparative advantages and costs of pumped water storage, steam accumulators, storage of boiler feedwater , and compressed air storage in caverns are described. Boiler feedwater storage in caverns and pumped water storage are most economical. All systems are costly enough to justify developing reactors with load following capabilities. (E.C.B.)

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-02-01

This report describes the engineering conceptual design of Fusion Experimental Reactor (FER) which is to be built as a next generation tokamak machine. This design covers overall reactor systems including MHD equilibrium analysis, mechanical configuration of reactor, divertor, pumped limiter, first wall/breeding blanket/shield, toroidal field magnet, poloidal field magnet, cryostat, electromagnetic analysis, vacuum system, power handling and conversion, NBI, RF heating device, tritium system, neutronics, maintenance, cooling system and layout of facilities. The engineering comparison of a divertor with pumped limiters and safety analysis of reactor systems are also conducted. (author)

The Performance test of Mechanical Sodium Pump with Water Environment

International Nuclear Information System (INIS)

Cho, Chungho; Kim, Jong-Man; Ko, Yung Joo; Jeong, Ji-Young; Kim, Jong-Bum; Ko, Bock Seong; Park, Sang Jun; Lee, Yoon Sang

2015-01-01

As contrasted with PWR(Pressurized light Water Reactor) using water as a coolant, sodium is used as a coolant in SFR because of its low melting temperature, high thermal conductivity, the high boiling temperature allowing the reactors to operate at ambient pressure, and low neutron absorption cross section which is required to achieve a high neutron flux. But, sodium is violently reactive with water or oxygen like the other alkali metal. So Very strict requirements are demanded to design and fabricate of sodium experimental facilities. Furthermore, performance testing in high temperature sodium environments is more expensive and time consuming and need an extra precautions because operating and maintaining of sodium experimental facilities are very difficult. The present paper describes performance test results of mechanical sodium pump with water which has been performed with some design changes using water test facility in SAM JIN Industrial Co. To compare the hydraulic characteristic of model pump with water and sodium, the performance test of model pump were performed using vender's experimental facility for mechanical sodium pump. To accommodate non-uniform thermal expansion and to secure the operability and the safety, the gap size of some parts of original model pump was modified. Performance tests of modified mechanical sodium pump with water were successfully performed. Water is therefore often selected as a surrogate test fluid because it is not only cheap, easily available and easy to handle but also its important hydraulic properties (density and kinematic viscosity) are very similar to that of the sodium. Normal practice to thoroughly test a design or component before applied or installed in reactor is important to ensure the safety and operability in the sodium-cooled fast reactor (SFR). So, in order to estimate the hydraulic behavior of the PHTS pump of DSFR (600 MWe Demonstraion SFR), the performance tests of the model pump such as performance

Alignment analysis of a vertical sodium pump

International Nuclear Information System (INIS)

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

1981-01-01

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

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)

Emergency reactor container cooling facility

International Nuclear Information System (INIS)

Suzuki, Hiroaki; Matsumoto, Tomoyuki.

1992-01-01

The present invention concerns an emergency cooling facility for a nuclear reactor container having a pressure suppression chamber, in which water in the suppression chamber is effectively used for cooling the reactor container. That is, the lower portion of a water pool in the pressure suppression chamber and the inside of the reactor container are connected by a pipeline. The lower end of the pipeline and a pressurized incombustible gas tank disposed to the outside of the reactor container are connected by a pipeline by way of valves. Then, when the temperature of the lower end of the pressure vessel exceeds a predetermined value, the valves are opened. If the valves are opened, the incombustible gas flows into the lower end of the pipeline connecting the lower portion of the water pool in the pressure suppression chamber and the inside of the reactor container. Since the inside of the pipeline is a two phase flow comprising a mixture of a gas phase and a liquid phase, the average density is decreased. Therefore, the water level of the two phase flow is risen by the level difference between the inside and the outside of the pipeline and, finally, the two phase mixture is released into the reactor container. As a result, the reactor container can be cooled by water in the suppression chamber by a static means without requiring pumps. (I.S.)

CENTAR code for extended nonlinear transient analysis of extraterrestrial reactor systems

International Nuclear Information System (INIS)

Nassersharif, B.; Peer, J.S.; DeHart, M.D.

1987-01-01

Current interest in the application of nuclear reactor-driven power systems to space missions has generated a need for a systems simulation code to model and analyze space reactor systems; such a code has been initiated at Texas A and M, and the first version is nearing completion; release was anticipated in the fall of 1987. This code, named CENTAR (Code for Extended Nonlinear Transient Analysis of Extraterrestrial Reactor Systems), is designed specifically for space systems and is highly vectorizable. CENTAR is composed of several specialized modules. A fluids module is used to model fluid behavior throughout the system. A wall heat transfer module models the heat transfer characteristics of all walls, insulation, and structure around the system. A fuel element thermal analysis module is used to predict the temperature behavior and heat transfer characteristics of the reactor fuel rods. A kinetics module uses a six-group point kinetics formulation to model reactivity feedback and control and the ANS 5.1 decay-heat curve to model shutdown decay-heat production. A pump module models the behavior of thermoelectric-electromagnetic pumps, and a heat exchanger module models not only thermal effects in thermoelectric heat exchangers, but also predicts electrical power production for a given configuration. Finally, an accumulator module models coolant expansion/contraction accumulators

Isolation colling device for reactor

International Nuclear Information System (INIS)

Ikehara, Morihiko; Arai, Shigeki.

1982-01-01

Purpose: To prevent undesired operation of an emergency core cooling system due to excess lowering of water level in a reactor. Constitution: In an emergency facility adapted to drive a turbine, upon reactor isolation, with the excess steams of the reactor to operate a pump and thereby inject cooling water to the reactor, a water level detector is provided and connected to a pump exhaust valve control circuit, a turbine inlet valve control circuit and a by-pass valve control circuit. Valve ON-OFF is automatically controlled depending on the water level to thereby render the level constant. A by-pass pipe is branched from a pump exhaust pipe and connected to a condensate storage tank. When the water level rises due to water injection, the injecting water is returned to circulate by way of the by-pass pipe to the condensate storage tank under the ON-OFF for each of the valves while the turbine being kept to drive. Then, if the water level is lowered, water injection is started again by the ON-OFF for each of the valves. (Ikeda, J.)

Supercritical Water Reactor Cycle for Medium Power Applications

International Nuclear Information System (INIS)

BD Middleton; J Buongiorno

2007-01-01

and pipes were modeled with realistic assumptions using the PEACE module of Thermoflex. A three-dimensional layout of the plant was also generated with the SolidEdge software. The results of the engineering design are as follows: (i) The cycle achieves a net thermal efficiency of 24.13% with 350/460 C reactor inlet/outlet temperatures, ∼250 bar reactor pressure and 0.75 bar condenser pressure. The steam quality at the turbine outlet is 90% and the total electric consumption of the pumps is about 2500 kWe at nominal conditions. (ii) The overall size of the plant is attractively compact and can be further reduced if a printed-circuit-heat-exchanger (vs shell-and-tube) design is used for the feedwater heater, which is currently the largest component by far. Finally, an analysis of the plant performance at off-nominal conditions has revealed good robustness of the design in handling large changes of thermal power and seawater temperature

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

Gas cooled reactors

International Nuclear Information System (INIS)

Kojima, Masayuki.

1985-01-01

Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

Hydrostatic radial bearing of centrifugal pump

International Nuclear Information System (INIS)

Skalicky, A.

1976-01-01

A hydrostatic radial pump is described characterized by the fact that part of the medium off-taken from delivery is used as a lubricating medium. Two additional bodies are placed alongside a hydrostatic bearing with coils in between them and the pump shaft; the coils have an opposite pitch. The feed channel for the hydrostatic bearing pocket is linked to delivery. The coil outlets are connected to the pump suction unit. Two rotating coils placed alongside the hydrostatic bearing will considerably simplify the communication channel design and reduce the dependence on the pump shaft deflections. The addition of another rotating coil in the close vicinity of the pump shaft or directly on the shaft further increases the efficiency. The bearing can be used in designing vertical circulating pumps for the cooling circuits of nuclear reactors. (J.B.)

Recycling flow rate control device in BWR type reactor

International Nuclear Information System (INIS)

Fujiwara, Tadashi; Koda, Yasushi

1988-01-01

Purpose: To reduce the recycling pump speed if the pressure variation width and the variation ratio in the nuclear reactor exceed predetermined values, to thereby avoid the shutdown of the plant. Constitution: There has been proposed a method of monitoring the neutron flux increase thereby avoiding unnecessary plant shutdown, but it involves a problems of reactor scram depending on the state of the plant and the set values. In view of the above, in the plant using internal pumps put under the thyristor control and having high response to recycling flow rate, the reactor pressure is monitored and the speed of the internal pump is rapidly reduced when the pressure variation width and variation ratio exceed predetermined values to reduce the reactor power and avoid the plant shutdown. This can reduce the possibility of unnecessary power reduction due to neutron flux noises or the possibility of plant shutdown under low power conditions. Further, since the reactor operation can be continued without stopping the recycling pump, the operation upon recovery can be made rapid. (Horiuchi, T.)

Savannah River Site reactor hardware design modification study

International Nuclear Information System (INIS)

Fisher, J.E.

1990-01-01

A study was undertaken to assess the merits of proposed design modifications to the Savannah River Site (SRS) reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. System recovery potential was evaluated for break locations at the pump suction, the pump discharge, and the plenum inlet. The code version used was RELAP5/MOD2.5 version 3d3, a preliminary version of RELAP5/MOD3. The model was a three-dimensional representation of the K-Reactor water plenum and moderator tank. It included explicit representations of all six loops, which were based on the configuration of L-Reactor. A combination of features is recommended to ensure liquid inventory recovery for all break locations. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 7 refs., 10 figs., 2 tabs

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

The flow measurement methods for the primary system of integral reactors

International Nuclear Information System (INIS)

Lee, J.; Seo, J. K.; Lee, D. J.

2001-01-01

It is the common features of the integral reactors that the main components of the primary system are installed within the reactor vessel, and so there are no any flow pipes connecting the reactor coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the primary system of the integral reactors, and it also makes impossible measure the primary coolant flow rate. The objective of the study is to draw up the flow measurement methods for the primary system of integral reactors. As a result of the review, we have made a selection of the flow measurement method by pump speed, bt HBM, and by pump motor power as the flow measurement methods for the primary system of integral reactors. Peculiarly, we did not found out a precedent which the direct pump motor power-flow rate curve is used as the flow measurement method in the existing commercial nuclear power reactors. Therefore, to use this method for integral reactors, it is needed to bear the follow-up measures in mind. The follow-up measures is included in this report

Design of a Mechanical NaK Pump for Fission Space Power Systems

Science.gov (United States)

Mireles, Omar R.; Bradley, David; Godfroy, Thomas

2010-01-01

Alkali liquid metal cooled fission reactor concepts are under development for mid-range spaceflight power requirements. One such concept utilizes a sodium-potassium eutectic (NaK) as the primary loop working fluid. Traditionally, linear induction pumps have been used to provide the required flow and head conditions for liquid metal systems but can be limited in performance. This paper details the design, build, and check-out test of a mechanical NaK pump. The pump was designed to meet reactor cooling requirements using commercially available components modified for high temperature NaK service.

Main coolant pump testing at Ontario Hydro

International Nuclear Information System (INIS)

Hartlen, R.

1991-01-01

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-08-01

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

A high-flow holweck pump for fusion applications

International Nuclear Information System (INIS)

Iseli, M.; Dinner, P.J.; Murdoch, D.K.

1995-01-01

Present concepts for power reactors require high pumping speed for the torus exhaust (10 5 -10 6 1/s) with low tritium inventories. Conventional approaches using Compound Cryopumps necessitate high tritium inventories and Turbomolecular pumps require large scale-up in throughput and are sensitive to eddy current heating of the rotor and sudden venting thrust. Cooling the gas to low temperature (20K) increases the gas density at the pump-entrance enough to obtain high throughputs from compact mechanical devices such as molecular drag pumps. A numerical model of such a pump and experimental results confirm the high pumping speed achievable with this concept. The model is used for extrapolation and optimisation of the design of a prototype. (orig.)

On-line monitoring of main coolant pump seals

International Nuclear Information System (INIS)

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

1984-06-01

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

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

Feedwater processing method in a boiling water reactor

Energy Technology Data Exchange (ETDEWEB)

Izumitani, M; Tanno, K

1976-09-06

The purpose of the invention is to decrease a quantity of corrosion products moving from the feedwater system to the core. Water formed into vapor after heated in a reactor is fed to the turbine through a main steam line to drive a generator to return it to liquid-state water in a condenser. The water is then again cycled into the reactor via the condensate pump, desalting unit, low pressure feedwater heater, medium pressure feedwater heater, and high pressure feedwater heater. The reactor water is recycled by a recycling pump. At this time, the reactor water recycled by the recycling pump is partially poured into a middle point between the desalting unit and the low pressure feedwater heater through a reducing valve or the like. With the structure described above, the quantity of the corrosion products from the feedwater system may be decreased by the function of a large quantity of active oxygen contained in the reactor water.

Reactor core flow rate control system

International Nuclear Information System (INIS)

Sakuma, Hitoshi; Tanikawa, Naoshi; Takahashi, Toshiyuki; Miyakawa, Tetsuya.

1996-01-01

When an internal pump is started by a variable frequency power source device, if magnetic fields of an AC generator are introduced after the rated speed is reached, neutron flux high scram occurs by abrupt increase of a reactor core flow rate. Then, in the present invention, magnetic fields for the AC generator are introduced at a speed previously set at which the fluctuation range of the reactor core flow rate (neutron flux) by the start up of the internal pump is within an allowable value. Since increase of the speed of the internal pump upon its start up is suppressed to determine the change of the reactor core flow rate within an allowable range, increase of neutron fluxes is suppressed to enable stable start up. Then, since transition boiling of fuels caused by abrupt decrease of the reactor core flow rate upon occurrence of abnormality in an external electric power system is prevented, and the magnetic fields for the AC generator are introduced in such a manner to put the speed increase fluctuation range of the internal pump upon start up within an allowable value, neutron flux high scram is not caused to enable stable start-up. (N.H.)

Emergency core cooling system in BWR type reactors

International Nuclear Information System (INIS)

Takizawa, Yoji

1981-01-01

Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

Emergency water supply facility for nuclear reactor

International Nuclear Information System (INIS)

Karasawa, Toru

1998-01-01

Water is stored previously in an equipment storage pit disposed on an operator floor of a reactor building instead of a condensate storage vessel. Upon occurrence of an emergency, water is supplied from the equipment storage pit by way of a sucking pipeline to a pump of a high pressure reactor core water injection circuit and a pump of a reactor-isolation cooling circuit to supply water to a reactor. The equipment storage pit is arranged in a building so that the depth thereof is determined to keep the required amount of water by storing water at a level lower than the lower end of a pool gate during normal operation. Water is also supplied from the equipment storage pit by way of a supply pipeline to a spent fuel storage pool on the operation floor of the reactor building. Namely, water is supplied to the spent fuel storage pool by a pump of a fuel pool cooling and cleaning circuit. This can eliminate a suppression pool cleaning circuit. (I.N.)

Operating states of PFBR-primary sodium pump

International Nuclear Information System (INIS)

Sreedharan, K.V.; Rao, A.S.L.K.; Raju, C.; Kale, R.D.

1994-01-01

The proposed 500 MWe Prototype Fast Breeder Reactor (PFBR), the design of which is in progress at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam will be equipped with four centrifugal sodium pumps each in its primary and secondary circuits. This paper discusses the various operating states of the primary pumps viz. (i) normal steady state (ii) transient state operation. The paper also discusses the safe limits on the pump operating speed and a few design basis incidents. (author). 1 ref., 2 figs., 3 tabs

Vacuum pumping by the halo plasma

International Nuclear Information System (INIS)

Barr, W.L.

1985-01-01

An estimate is made of the effective vacuum pumping speed of the halo plasma in a tandem mirror fusion reactor, and it is shown that, if the electron temperature and line density are great enough, the halo can be a very good vacuum pump. One can probably obtain the required density by recycling the ions at the halo dumps. An array of small venting ports in the dump plates allows local variation of the recycle fraction and local removal of the gas at a conveniently high pressure. This vented-port concept could introduce more flexibility in the design of pumped limiters for tokamaks

Heat Radiators for Electromagnetic Pumps

Science.gov (United States)

Campana, R. J.

1986-01-01

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

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Auxiliary water supply device for BWR type reactor

International Nuclear Information System (INIS)

Sasagawa, Hiroshi.

1994-01-01

In the device of the present invention, a cooling condensation means is disposed to a steam discharge channel of a turbine for driving pumps to directly return condensates to the reactor, so that the temperature of the suppression pool water is not elevated. Namely, the cooling condensation means for discharged steams is disposed to the discharge channel of the turbine. The condensate channel from the cooling condensation means is connected to a sucking side of the turbine driving pump. With such a constitution, when the reactor is isolated from a main steam system, reactor scram is conducted. Although the reactor water level is lowered by the reactor scram, the lowering of the reactor water level is prevented by supplementing cooling water by the turbine driving pump using steams generated in the reactor as a power source. The discharged steams after driving the turbine are cooled and condensated by the cooling condensation means by way of the discharge channel and returned to the reactor again by way of the condensate channel. With such procedures, since the temperature of suppression pool water is not elevated, there is no need to operate other cooling systems. In addition, auxiliary water can be supplied for a long period of time. (I.S.)

Fission fragment assisted reactor concept for space propulsion: Foil reactor

International Nuclear Information System (INIS)

Wright, S.A.

1991-01-01

The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures

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

Heat removing device for reactor container

International Nuclear Information System (INIS)

Hisamochi, Kohei; Matsumoto, Tomoyuki; Matsumoto, Masayoshi; Sato, Ken-ichi.

1996-01-01

A recycling loop for reactor water is disposed in a reactor pressure vessel of a BWR type reactor. Extracted reactor water from the recycling loop passes through a extracted reactor water pipeline and flows into a reactor coolant cleanup system. A pipeline for connecting the extracted reactor water pipeline and a suppression pool is disposed, and a discharged water pressurizing pump is disposed to the pipeline. Upon occurrence of emergency, discharged water from the suppression pool is pressurized by a discharged water pressurizing pump and sent to a reactor coolant cleanup system. The discharged water is cooled while passing through a sucking water cooling portion of a regenerative heat exchanger and a non-regenerative heat exchanger. Then, it is sent to a feed water pipeline passing a bypass line of a filtering desalter and a bypass line of the sucked water cooling portion of the regenerative heat exchanger, injected to the inside of the pressure vessel to cool the reactor core and remove after-heat. Then, it cools the inside of the reactor container together with coolants flown out of the pressure vessel and then returns to the suppression pool. (I.N.)

Novel limiter pump topologies

International Nuclear Information System (INIS)

Schultz, J.H.

1981-01-01

The use of limiter pumps as the principle plasma exhaust system of a magnetic confinement fusion device promises significant simplification, when compared to previously investigating divertor based systems. Further simplifications, such as the integration of the exhaust system with a radio frequency heating system and with the main reactor shield and structure are investigated below. The integrity of limiters in a reactor environment is threatened by many mechanisms, the most severe of which may be erosion by sputtering. Two novel topolgies are suggested which allow high erosion without limiter failure

Novel limiter pump topologies

International Nuclear Information System (INIS)

Schultz, J.H.

1981-01-01

The use of limiter pumps as the principle plasma exhaust system of a magnetic confinement fusion device promises significant simplification, when compared to previously investigating divertor based systems. Further simplifications, such as the integration of the exhaust system with a radio frequency heating system and with the main reactor shield and structure are investigated below. The integrity of limiters in a reactor environment is threatened by many mechanisms, the most severe of which may be erosion by sputtering. Two novel topologies are suggested which allow high erosion without limiter failure

Savannah River Site reactor hardware design modification study

International Nuclear Information System (INIS)

Fisher, J.E.

1990-03-01

A study was undertaken to assess the merits of proposed design modifications to the SRS reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. For the elevated piping design, system recovery was predicted for breaks in the plenum inlet or pump suction piping; response to the pump discharge break location did not show improvement compared to the present system configuration. The rotovalve closure design improved system response to plenum inlet or pump discharge breaks; recovery was not predicted for pump suction breaks. The pump suction valve closure design demonstrated system recovery for all break locations downstream of the valve. A combination of features is recommended to ensure liquid inventory recovery for all break locations. The elevated piping design performance during pump discharge breaks would be improved with addition of a dc pump trip in the affected loop. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 12 refs., 10 figs., 2 tabs

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

Reactor protection system

International Nuclear Information System (INIS)

Fairbrother, D.B.; Lesniak, L.M.; Orgera, E.G.

1977-10-01

The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

Flow-induced vibration characteristics of the BWR/5-201 jet pump

International Nuclear Information System (INIS)

LaCroix, L.V.

1982-09-01

A General Electric boiling water reactor BWR/5-201 jet pump was tested for flow-induced vibration (FIV) characteristics in the Large Steam Water Test Facility at Moss Landing, CA, during the period June-July 1978. High level periodic FIV were observed at reactor operating conditions (1027 psia, 532 0 F and prototypical flow rates) for the specific single jet pump assembly tested. High level FIV of similar amplitude and character have been shown capable of damaging jet pump components and associated support hardware if allowed to continue unchecked. High level FIV were effectively suppressed in two special cases tested: (1) lateral load (>500 lb) at the mixer to diffuser slip joint; and (2) a labyrinth seal (5 small, circumferential grooves) on the mixer at the slip joint. Stability criteria for the particular jet pump tested were developed from test data. A cause-effect relationship between the dynamic pressure within the slip joint and the jet pump vibration was established

The Thermal-hydraulic Analysis for the Aging Effect of the Component in CANDU-6 Reactor

International Nuclear Information System (INIS)

Bae, Jun Ho; Jung, Jong Yeob

2014-01-01

CANDU reactor consists of a lot of components, including pressure tube, reactor pump, steam generator, feeder pipe, and so on. These components become to have the aging characteristics as the reactor operates for a long time. The aging phenomena of these components lead to the change of operating parameters, and it finally results to the decrease of the operating safety margin. Actually, due to the aging characteristics of components, CANDU reactor power plant has the operating license for the duration of 30 years and the plant regularly check the plant operating state in the overhaul period. As the reactor experiences the aging, the reactor operators should reduce the reactor power level in order to keep the minimum safety margin, and it results to the deficit of economical profit. Therefore, in order to establish the safety margin for the aged reactor, the aging characteristics for components should be analyzed and the effect of aging of components on the operating parameter should be studied. In this study, the aging characteristics of components are analyzed and revealed how the aging of components affects to the operating parameter by using NUCIRC code. Finally, by scrutinizing the effect of operating parameter on the operating safety margin, the effect of aging of components on the safety margin has been revealed

Mechanical development for reliable reactor components

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Metcalfe, R.

1983-09-01

The CANDU reactor has achieved worldwide distinction because of its reliable performance. To achieve this, special attention was given to the reliability and maintainability of components in the heavy water circuits. Development programs were initiated early in the history of the CANDU reactor to improve the effectiveness of pump seals, valves, and static seals because of unacceptable performance of the commercial equipment then available. As a result, pump seals with a five year life now appear achievable, and valves and static seals are no longer a significant concern in CANDU reactors. Increasing effort is being given remotely operated tools and fabrication systems for radioactive environments

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.

Residual stress improving method for reactor structural component and residual stress improving device therefor

Energy Technology Data Exchange (ETDEWEB)

Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato

1996-09-03

The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

Residual stress improving method for reactor structural component and residual stress improving device therefor

International Nuclear Information System (INIS)

Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato.

1996-01-01

The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

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

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

Condition monitoring of pumps with co-relating field observations

International Nuclear Information System (INIS)

Mishra, S.K.; Prasad, V.; Sharma, R.B.

1994-01-01

The maintenance of 40 MWth research reactor, Cirus has been carried out for over 30 years following the time based maintenance schedule. With the commissioning of indigenously built 100 MWth nuclear research reactor Dhruva in the year 1985, a systematic work on condition monitoring has been commissioned. Apart from process parameters, which are recorded on hourly basis, vibration, noise, temperature, kurtosis etc. are measured for assessment of condition of pumps. The bearings of flywheel assembly of main pumps, Dhruva broke down almost abruptly during the initial years after first commissioning. The regular measurements of vibration level and kurtosis have greatly helped in avoiding breakdown. In a recent case one newly procured herringbone gear box (300 hp, 1475/1760 rpm) for the primary coolant pump was showing high vibration. In further checking using Fast Fourier Transform (FFT) analyser in a time domain plot the gear teeth damage was indicated. The pump was shut down for inspection and when the gear box was dismantled teeth were found broken. An attempt has been made in this paper to discuss a few interesting field experiences with condition monitoring and correlating field observations on pumps. (author). 3 figs

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)

An analytical model for neutral and charged particles in closed pump limiter

International Nuclear Information System (INIS)

Tokar, M.Z.

1990-01-01

In number of modern tokamaks pump limiters are used for control of the edge plasma parameters. In thermonuclear reactor projects pump limiters are considered as a possible alternative or additional for divertor configuration of magnetic field method for unburnt fuel, helium ash and other impurity exhaust. From the point of view of limiter operation efficiency the dependence of neutral particle pressure P g in pumping system on the plasma parameters in the scrape-off layer (SOL) is of importance since an increase of P g will permit to decrease needed pump rate and facilitate solution of reactor technological problems. In connection with this the experimentally observed mode of closed pump limiter operation with strong neutral compression is of great interest. In such a mode P g sharply grows up when the SOL plasma density n s exceeds some critial level. (author) 6 refs., 2 figs

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.

Design and experimental characterization of an EM pump

International Nuclear Information System (INIS)

Kim, Hee Reyoung; Hong, Sang Hee

1999-01-01

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

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

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)

Experimental Results and Model Calculations of a Hybrid Adsorption-Compression Heat Pump Based on a Roots Compressor and Silica Gel-Water Sorption

Energy Technology Data Exchange (ETDEWEB)

Van der Pal, M.; De Boer, R.; Wemmers, A.K.; Smeding, S.F.; Veldhuis, J.B.J.; Lycklama a Nijeholt, J.A.

2013-10-15

Thermally driven sorption systems can provide significant energy savings, especially in industrial applications. The driving temperature for operation of such systems limits the operating window and can be a barrier for market-introduction. By adding a compressor, the sorption cycle can be run using lower waste heat temperatures. ECN has recently started the development of such a hybrid heat pump. The final goal is to develop a hybrid heat pump for upgrading lower (<100C) temperature industrial waste heat to above pinch temperatures. The paper presents the first measurements and model calculations of a hybrid heat pump system using a water-silica gel system combined with a Roots type compressor. From the measurements can be seen that the effect of the compressor is dependent on where in the cycle it is placed. When placed between the evaporator and the sorption reactor, it has a considerable larger effect compared to the compressor placed between the sorption reactor and the condenser. The latter hardly improves the performance compared to purely heat-driven operation. This shows the importance of studying the interaction between all components of the system. The model, which shows reasonable correlation with the measurements, could proof to be a valuable tool to determine the optimal hybrid heat pump configuration.

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

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

Nuclear research reactor 0.5 to 3 MW

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-05-15

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor

Nuclear research reactor 0.5 to 3 MW

International Nuclear Information System (INIS)

1992-05-01

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation

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)

Construction of a Vibration Monitoring System for HANARO's Rotating Machinery and Analysis of Pump Vibration Signals

International Nuclear Information System (INIS)

Ryu, Jeong Soo; Yoon, Doo Byung

2005-01-01

HANARO is an open-tank-in-pool type research reactor with a thermal power of 30MW. In order to remove the heat generated by the reactor core and the reflector vessel, primary cooling pumps and reflector cooling pumps circulate coolant. These pumps are installed at the RCI(Reactor Concrete Island) which is covered by heavy concrete hatches. For the prevention of an abnormal operation of these pumps in the RCI, it is necessary to construct a vibration monitoring system that provides an alarm signal to the reactor control room when the rotating speed or the vibration level exceeds the allowable limit. The first objective of this work is to construct a vibration monitoring system for HANARO's rotating machinery. The second objective is to verify the possibility of condition monitoring of the rotating machinery. To construct a vibration monitoring system, as a first step, the standards and references related to the vibration monitoring system were investigated. In addition, to determine the number and the location of sensors that can effectively characterize the overall vibration of a pump, the vibration of the primary cooling pumps and the reflector cooling pumps were measured. Based on these results, detailed construction plans for the vibration monitoring system for HANARO were established. Then, in accordance with the construction plans, the vibration monitoring system for HANARO's rotating machinery was manufactured and installed at HANARO. To achieve the second objective, FFT analysis and bearing fault detection of the measured vibration signals were performed. The analysis results demonstrate that the accelerometers mounted at the bearing locations of the pumps can effectively monitor the pump condition

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

Vibration monitoring of large vertical pumps via a remote satellite station

International Nuclear Information System (INIS)

Cook, S.A.; Crowe, R.D.; Roblyer, S.P.; Toffer, H.

1985-01-01

The Hanford N Reactor is operated by UNC Nuclear Industries for the Department of Energy for the production of special isotopes and electric energy. The reactor has a unique design in which the equipment such as pumps, turbines, generators and diesel engines are located in separate buildings. This equipment arrangement has led to the conclusion that the most cost-effective implementation of a dedicated vibration monitoring system would be to install a computerized network system in lieu of a single analyzing station. In this approach, semi-autonomous micro processor based data collection stations referred to as satellite stations are located near each concentration of machinery to be monitored. The satellite stations provide near continuous monitoring of the machinery. They are linked to a minicomputer using voice grade telephone circuits and hardware and software specifically designed for network communications. The communications link between the satellite stations and the minicomputer permits data and programs to be transmitted between the units. This paper will describe the satellite station associated with large vertical pumps vibration monitoring. The reactor has four of these pumps to supply tertiary cooling to reactor systems. 4 figs

Qualification test of a main coolant pump for SMART pilot

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Nuclear-pumped lasers for large-scale applications

International Nuclear Information System (INIS)

Anderson, R.E.; Leonard, E.M.; Shea, R.F.; Berggren, R.R.

1989-05-01

Efficient initiation of large-volume chemical lasers may be achieved by neutron induced reactions which produce charged particles in the final state. When a burst mode nuclear reactor is used as the neutron source, both a sufficiently intense neutron flux and a sufficiently short initiation pulse may be possible. Proof-of-principle experiments are planned to demonstrate lasing in a direct nuclear-pumped large-volume system; to study the effects of various neutron absorbing materials on laser performance; to study the effects of long initiation pulse lengths; to demonstrate the performance of large-scale optics and the beam quality that may be obtained; and to assess the performance of alternative designs of burst systems that increase the neutron output and burst repetition rate. 21 refs., 8 figs., 5 tabs

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.

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

Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.

2003-01-01

Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

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

Vacuum engineering for fusion research and fusion reactors

International Nuclear Information System (INIS)

Pittenger, L.C.

1976-01-01

The following topics are described: (1) surface pumping by cryogenic condensation, (2) operation of large condensing cryopumps, (3) pumping for large fusion experiments, and (4) vacuum technology for fusion reactors

Nuclear reactor

International Nuclear Information System (INIS)

Garabedian, G.

1988-01-01

A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

Tokamak pump limiters

International Nuclear Information System (INIS)

Conn, R.W.

1984-05-01

Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described

Tokamak pump limiters

International Nuclear Information System (INIS)

Conn, R.W.; California Univ., Los Angeles

1984-01-01

Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scrape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this 'Z-mode' of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described. (orig.)

Reactor protection system. Revision 1

International Nuclear Information System (INIS)

Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

1975-04-01

The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

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.)

Dual throat NaK pump performance evaluation (conceptual design of prototype)

International Nuclear Information System (INIS)

Johnson, J.L.

1972-01-01

An evaluation of the performance of the dual throat NaK pump for the 5 Kwe Reactor TE System is discussed. Performance at nominal operating conditions and at startup conditions is predicted. The basis for selection of this pump design is presented. (U.S.)

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

Compact, Lightweight Electromagnetic Pump for Liquid Metal

Science.gov (United States)

Godfroy, Thomas; Palzin, Kurt

2010-01-01

A proposed direct-current electromagnetic pump for circulating a molten alkali metal alloy would be smaller and lighter and would demand less input power, relative to currently available pumps of this type. (Molten alkali metals are used as heat-transfer fluids in high-temperature stages of some nuclear reactors.) The principle of operation of this or any such pump involves exploitation of the electrical conductivity of the molten metal: An electric current is made to pass through the liquid metal along an axis perpendicular to the longitudinal axis of the flow channel, and a magnetic field perpendicular to both the longitudinal axis and the electric current is superimposed on the flowchannel region containing the electric current. The interaction between the electric current and the magnetic field produces the pumping force along the longitudinal axis. The advantages of the proposed pump over other such pumps would accrue from design features that address overlapping thermal and magnetic issues.

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

Reinforced confinement in a nuclear reactor

International Nuclear Information System (INIS)

Norman, H.

1988-01-01

The present invention concerns a nuclear reactor containing a reactor core, a swimming pool space that is filled and pressurized with a neutron-absorbing solution, a reactor tank, at least one heat exchanger, at least one inlet line, at least one return line and at least one circulation pump, where the said reactor tank is confined in the said swimming pool space and designed to be cooled with the aid of relatively pure water, which is fed by means of the said at least one circulating pump to the said reactor tank from the said heat exchanger via the said at least one inlet line and is returned to the heat exchanger via the said at least one return line. The problem that is to be solved by the invention is to design a reactor of the above type in such a way that a complete confinement of the primary circuit of the reactor is achieved at relatively low extra cost. This problem is solved by providing the reactor with a special confinement space that confines the heat exchanger, but not the reactor tank, with the confinement space and the swimming pool space being fashioned in the same concrete body

Device for controlling water supply to nuclear reactor

International Nuclear Information System (INIS)

Iwasaki, Toshio.

1974-01-01

Object: To smoothly control automatic water supply for realizing stable operation of a nuclear reactor by providing a flow rate limiting signal selection circuit and a preferential circuit in a water supply control device for a nuclear reactor wherein the speed of a recirculation pump may be changed in two-steps. Structure: Opening angle signals for a water supply regulating valve are controlled by a nuclear reactor water level signal, a vapor flow rate signal and a supplied water flow rate signal through an adder and an adjuster in response to a predetermined water level setting signal. When the water in the reactor is maintained at a predetermined level, a selection circuit receives a water pump condition signal for selecting one of the signals from a supplied water rate limiting signal generator generating signals for indicating whether one or two water supply pumps are operated. A low value preferential circuit passes the lower of the values generated from the selection circuit and the adder. The selection circuit receives a recirculation pump condition signal and selects either one of the signals from the supplied water flow rate limiting signal generator operated at high speed or low speed. A high value preferential circuit passes the higher value

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.

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

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.)

Liquid metal flow control by DC electromagnetic pumps

International Nuclear Information System (INIS)

Borges, Eduardo Madeira; Braz Filho, Francisco Antonio; Guimaraes, Lamartine Nogueira Frutuoso

2006-01-01

The cooling system of high-density thermal power requires fluids of high thermal conductivity, such as liquid metals. Electromagnetic pumps can be used to liquid metal fluid flow control in cooling circuits. The operation of electromagnetic pumps used to flow control is based on Lorentz force. This force can be achieved by magnetic field and electric current interaction, controlled by external independent power supplies. This work presents the electromagnetic pump operational principles, the IEAv development scheme and the BEMC-1 simulation code. The theoretical results of BEMC-1 simulation are compared to electromagnetic pump operation experimental data, validating the BEMC-1 code. This code is used to evaluate the DC electromagnetic pump performance applied to Mercury flow control and others liquid metal such as Sodium, Lead and Bismuth, used in nuclear fast reactors. (author)

Parallel operation of primary sodium pumps in FBTR

International Nuclear Information System (INIS)

Athmalingam, S.; Ellappan, T.R.; Vaidyanathan, G.; Chetal, S.C.; Bhoje, S.B.

1994-01-01

Sodium pumps used in the primary main circuit of Fast Breeder Test Reactor (FBTR) are centrifugal pumps. These pumps have a free level of sodium with a cover gas above it to simplify the pump seal arrangement. The sodium level in the pumps will vary based on the flow. The minimum level is governed by consideration of gas entrainment and net positive suction head (NPSH) to the pump while the maximum level is limited by sodium entering the pump tank gas line. There is a special feature in these pumps in that a small portion of the pump outlet sodium flow is led back into the suction chamber to maintain level and avoid gas entrainment. A control valve in this line helps in controlling the level at the desired value. With parallel operation of two sodium pumps a study was conducted to find the regions of safe operation of the two pumps. The purpose of this paper is to give the various design features and methodology of the analysis to arrive at the limiting condition of operation for the different operating states of the two pumps and the effect of pump speed variations on the fluctuations in sodium flows. (author). 6 figs

Leak detector for reactor pressure vessel

International Nuclear Information System (INIS)

Morimoto, Mikio.

1991-01-01

A branched pipe is disposed to a leak off pipeline led from a flange surface which connects the main body and the upper lid of a reactor pressure vessel. An exhaust pump is disposed to the branched pipe and a moisture gage is disposed on the side of the exhaustion and a dry air supplier is connected to the branched pipe. Upon conducting a pressure-proof leak test for the reactor pressure vessel, the exhaust pump is operated and an electromagnet valve disposed at the upstream of the dry air supplier is opened and closed repeatedly. The humidity of air sucked by the exhaust pump is detected by the moisture gage. If leaks should be caused in the joining surface of the flange, leaked water is diffused as steams. Accordingly, occurrence of leak can be detected instantly based on the comparison with the moisture level of the dry air as a standard. In this way, a leak test can be conducted reliably in a short period of time with no change of for the reactor pressure container itself. (I.N.)

85,000-GPM, single-stage, single-suction LMFBR intermediate centrifugal pump

International Nuclear Information System (INIS)

Fair, C.E.; Cook, M.E.; Huber, K.A.; Rohde, R.

1983-01-01

The mechanical and hydraulic design features of the 85,000-gpm, single-stage, single-suction pump test article, which is designed to circulate liquid-sodium coolant in the intermediate heat-transport system of a Large-Scale Liquid Metal Fast Breeder Reactor (LS-LMFBR), are described. The design and analytical considerations used to satisfy the pump performance and operability requirements are presented. The validation of pump hydraulic performance using a hydraulic scale-model pump is discussed, as is the featute test for the mechanical-shaft seal system

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.

Site specific health and safety plan, 100-HR-3 pump and treat. Revision 2

International Nuclear Information System (INIS)

St John, C.H.

1997-09-01

The 100-HR-3 Pump and Treat system is a groundwater remedial action to remove Hexavalent Chromium (Cr+6) from the groundwater underlying the 100-HR-3 Operable Unit. This plan covers operation, maintenance, repairs, resin exchange and equipment removal/installation. The 100-HR-3 Operable Unit addresses groundwater underlying the 100-D Area. The primary groundwater contaminant is Chromium +6. The chromium contamination resulted from the use of sodium dichromate during past reactor operations. Sodium dichromate was used to treat reactor coolant water during reactor operations. The purpose of this Pump and Treat system is to pump contaminated groundwater through above ground ion exchange resin and then return treated water to aquifer. Chromium levels extracted from the wells are anticipated to range in the low parts per billion (∼50 ppb) which is the drinking water limit for Cr+6

Absorption heat pumps

International Nuclear Information System (INIS)

Formigoni, C.

1998-01-01

A brief description of the difference between a compression and an absorption heat pump is made, and the reasons why absorption systems have spread lately are given. Studies and projects recently started in the field of absorption heat pumps, as well as criteria usually followed in project development are described. An outline (performance targets, basic components) of a project on a water/air absorption heat pump, running on natural gas or LPG, is given. The project was developed by the Robur Group as an evolution of a water absorption refrigerator operating with a water/ammonia solution, which has been on the market for a long time and recently innovated. Finally, a list of the main energy and cost advantages deriving from the use of absorption heat pumps is made [it

Liquid sodium pumps

International Nuclear Information System (INIS)

Allen, H.G.

1985-01-01

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

Failure cause and failure rate evaluation on pumps of BWR plants in PSA. Hypothesis testing for typical or plant specific failure rate of pumps

International Nuclear Information System (INIS)

Sanada, Takahiro; Nakamura, Makoto

2009-01-01

In support of domestic nuclear industry effort to gather and analyze failure data of components concerning nuclear power plants, Nuclear Information Archives (NUCIA) are published for useful information to help PSA. This report focuses on NUCIA pertaining to pumps in domestic nuclear power plants, and provides the reliable estimation on failure rate of pumps resulting from failure cause analysis and hypothesis testing of classified and plant specific failure rate of pumps for improving quality in PSA. The classified and plant specific failure rate of pumps are estimated by analyzing individual domestic nuclear power plant's data of 26 Boiling Water Reactors (BWRs) concerning functionally structurally classified pump failures reported from beginning of commercial operation to March 31, 2007. (author)

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

FTU pump limiter

International Nuclear Information System (INIS)

Alessandrini, C.; Ciotti, M.; Mattei, A. De; Maddaluno, G.; Mazzitelli, G.

1989-01-01

The control of the refuelling and recycling of the plasma is crucial in providing enhanced performances in tokamaks and steady-state operation in future reactors. In this paper, we report details of the design and analysis for the pump limiter to be incorporated into the FTU tokamak. The FTU, presently under commissioning, is a compact high field (B=8T), medium high density, circular cross section machine with small accesses. The dimensions of the equatorial port (width 8 cm) would reduce the length of the entrance throat to a few centimeters, which is unacceptable for efficient particle trapping. We have, therefore, designed a rotating blade of the pump limiter head that, in the working position, extends in the toroidal direction inside the vacuum chamber. (author) 8 refs., 4 figs

Continuous cryopump for steady state mirror fusion reactors

International Nuclear Information System (INIS)

Batzer, T.H.; Call, W.R.

1983-01-01

The characteristics of mirror fusion reactors, i.e., steady state operation, a low neutral gas density, and a large gas throughput require unique vacuum pumping capabilities. One approach that appears to meet these requirements is a liquid helium-cooled cryopump system in which a fixed portion can be isolated and degassed while the remainder continues to pump. The time to degas a rotating, fixed portion of the pumping area and the ratio of that area to the total area fixes the gas inventory in the chamber. It follows that the active pump area maintains the required neutral gas density and the time-averaged degassing rate equals the gas throughput. We have built such a cryopump whereby the gas condensed (deuterium) on the liquid helium-cooled panel can be transferred to a collector pump and subsequently to an exterior mechanical pump and exhausted. At panel loadings as high as 0.55 Torr-/lcm 2 the gas leakage during degassing is less than 8% and the degassing time is less than 10 min. Scaling to reactor size appears to be feasible

Effect of the self-pumped limiter concept on the tritium fuel cycle

International Nuclear Information System (INIS)

Finn, P.A.; Sze, D.K.; Hassanein, A.

1988-01-01

The self-pumped limiter concept was the impurity control system for the reactor in the Tokamak Power Systems Study (TPSS). The use of a self-pumped limiter had a major impact on the design of the tritium systems of the TPSS fusion reactor. The self-pumped limiter functions by depositing the helium ash under a layer of metal (vanadium). The majority of the hydrogen species are recycled at the plasma edge; a small fraction permeates to the blanket/coolant which was lithium in TPSS. Use of the self-pumped limiter results in the elimination of the plasma processing system. Thus, the blanket tritium processing system becomes the major tritium system. The main advantages achieved for the tritium systems with a self-pumped limiter are a reduction in the capital cost of tritium processing equipment as well as a reduction in building space, a reduced tritium inventory for processing and for reserve storage, an increase in the inherent safety of the fusion plant and an improvement in economics for a fusion world economy

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)

Development of the methodology for the MHD analysis in a linear induction electro-magnetic pump

International Nuclear Information System (INIS)

Seong, Seung Hwan; Hur, Seop; Kim, Seong O; Choi, Seok Ki; Wi, Myung Hwan; Jeon, Won Dae

2004-01-01

Generally, fast breeder reactors have adopted a liquid metal as a coolant for the heat transfer from the reactor to the heat exchangers. Since a liquid metal has an electrical conductivity, the pumping of the liquid metal may use an induction electro-magnetic (EM) pump which induces electrical current and body force on the metal flow. These linear induction pumps use a traveling magnetic field wave created by poly-phase currents and the induced currents and their associated magnetic field generate the Lorentz force whose effect can be actually the pumping of the liquid metal. The flow behaviors in the pump are very complex such as the existence of a rotational force, pulsation and so on, because the induction EM pump has time-varying magnetic fields and the induced convective currents which originate form the flow of the liquid metal. These phenomena generate a stability problem in the pump and depend on the changes of the magnetic field and fluid flow field due to the induced currents and the fluid flow of the liquid metal with time and complex pump geometry. Therefore, an exact flow analysis is required for designing and evaluating the stability of a pump