Decontamination of main coolant pumps

International Nuclear Information System (INIS)

Roofthooft, R.

1988-01-01

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

Transient two-phase performance of LOFT reactor coolant pumps

International Nuclear Information System (INIS)

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

1983-01-01

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

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

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

Impedance calculations for power cables to primary coolant pump motors

International Nuclear Information System (INIS)

Hegerhorst, K.B.

1977-01-01

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

Reactor coolant pump seals: improving their performance

International Nuclear Information System (INIS)

Pothier, N.E.; Metcalfe, R.

1986-06-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

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

Science.gov (United States)

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

2011-01-01

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

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

DEFF Research Database (Denmark)

Kolaei, Alireza Rezania; Rosendahl, Lasse

2012-01-01

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

Coolant cleanup system for BWR type reactor

International Nuclear Information System (INIS)

Kinoshita, Shoichiro; Araki, Hidefumi.

1993-01-01

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

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

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

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

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.)

Compartmentalized safety coolant injection system

International Nuclear Information System (INIS)

Johnson, F.T.

1983-01-01

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

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.)

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

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

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)

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)

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

Science.gov (United States)

Rezania, A.; Rosendahl, L. A.

2012-06-01

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

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

International Nuclear Information System (INIS)

Feng Junting; Xu Mi; Wang Guizeng

2003-01-01

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

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

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

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

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

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)

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

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

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.

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

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

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...

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-11-01

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

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

Science.gov (United States)

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

2014-01-01

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

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.)

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)

Tendency of nuclear pumps for PWR primary system

International Nuclear Information System (INIS)

Shibata, Takeshi

1976-01-01

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

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)

Coolant controls of a PEM fuel cell system

Science.gov (United States)

Ahn, Jong-Woo; Choe, Song-Yul

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-15

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

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).

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

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

International Nuclear Information System (INIS)

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

1983-03-01

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

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)

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

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

International Nuclear Information System (INIS)

Saito, Michio.

1981-01-01

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

Secondary coolant purification system

International Nuclear Information System (INIS)

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

1978-01-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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

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)

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)

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

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

Coolant clean up system in nuclear reactor

International Nuclear Information System (INIS)

Tajima, Fumio; Iwami, Hiroshi.

1981-01-01

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

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

DEFF Research Database (Denmark)

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

2012-01-01

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

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)

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.)

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

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

1983-03-01

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

Full reactor coolant system chemical decontamination qualification programs

Energy Technology Data Exchange (ETDEWEB)

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

1995-03-01

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

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

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

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

International Nuclear Information System (INIS)

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

1982-09-01

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

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)

Primary coolant recycling device for FBR type reactor

International Nuclear Information System (INIS)

Kanbe, Mitsuru; Tokiwai, Moriyasu

1998-01-01

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

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.)

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)

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

International Nuclear Information System (INIS)

Siamidis, John; Mason, Lee

2006-01-01

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

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.

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

International Nuclear Information System (INIS)

Hermansky, Jozef; Zavodsky, Martin

2014-01-01

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

The empirical intensity of PWR primary coolant pumps failure and repair

International Nuclear Information System (INIS)

Milivojevicj, S.; Riznicj, J.

1988-01-01

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-10-01

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

A system for cooling electronic elements with an EHD coolant flow

International Nuclear Information System (INIS)

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

2014-01-01

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

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

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

International Nuclear Information System (INIS)

Sun Yanfei; Zhong Shuoping; Huang Xiaojin

2014-01-01

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

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)

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

International Nuclear Information System (INIS)

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

1982-06-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Triboengineering problems of lead coolant in innovative fast reactors

International Nuclear Information System (INIS)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

Andreychek, T.S.

1987-01-01

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

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)

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)

Pumps for German pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Dernedde, R.

1984-01-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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

International Nuclear Information System (INIS)

Malinauskas, A.P.

1990-01-01

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

Coolant system decontamination

International Nuclear Information System (INIS)

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

1981-01-01

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

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

International Nuclear Information System (INIS)

Knop, W.; Schreiner, P.

2003-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-05-01

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

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

International Nuclear Information System (INIS)

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

1995-05-01

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

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

International Nuclear Information System (INIS)

Luo Bangqi

2012-01-01

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

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.

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

Nuclear reactor coolant and cover gas system

International Nuclear Information System (INIS)

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

1976-01-01

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

Analysis of the VVER-1000 coolant transient benchmark phase 1 with the code system RELAP5/PARCS

International Nuclear Information System (INIS)

Victor Hugo Sanchez Espinoza

2005-01-01

Full text of publication follows: As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during

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

Reactor auxiliary cooling facility and coolant supplying method therefor

International Nuclear Information System (INIS)

Ando, Koji; Kinoshita, Shoichiro.

1996-01-01

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

Reactor auxiliary cooling facility and coolant supplying method therefor

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1996-06-07

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

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

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

Design of the coolant system for the Large Coil Test Facility pulse coils

International Nuclear Information System (INIS)

Bridgman, C.; Ryan, T.L.

1983-01-01

The pulse coils will be a part of the Large Coil Test Facility in Oak Ridge, Tennessee, which is designed to test six large tokamak-type superconducting coils. The pulse coil set consists of two resistive coaxial solenoid coils, mounted so that their magnetic axis is perpendicular to the toroidal field lines of the test coil. The pulse coils provide transient vertical fields at test coil locations to simulate the pulsed vertical fields present in tokamak devices. The pulse coils are designed to be pulsed for 30 s every 150 s, which results in a Joule heating of 116 kW per coil. In order to provide this capability, the pulse coil coolant system is required to deliver 6.3 L/s (100 gpm) of subcooled liquid nitrogen at 10-atm absolute pressure. The coolant system can also cool down each pulse coil from room temperature to liquid nitrogen temperature. This paper provides details of the pumping and heat exchange equipment designed for the coolant system and of the associated instrumentation and controls

Coolant monitoring systems for PWR reactors

International Nuclear Information System (INIS)

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

1987-01-01

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

Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

International Nuclear Information System (INIS)

Sanchez-Espinoza, Victor Hugo

2008-07-01

As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

Energy Technology Data Exchange (ETDEWEB)

Sanchez-Espinoza, Victor Hugo

2008-07-15

As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

Role of system characteristics in evolution of pump hydraulic design

International Nuclear Information System (INIS)

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

1994-01-01

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

Real-time reactor coolant system pressure/temperature limit system

International Nuclear Information System (INIS)

Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

1991-01-01

This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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

International Nuclear Information System (INIS)

Finn, P.A.

1991-03-01

Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs

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

Energy Technology Data Exchange (ETDEWEB)

Finn, P.A.

1991-03-01

Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs.

Flow rate control systems for coolants for BWR type reactors

International Nuclear Information System (INIS)

Igarashi, Yoko; Kato, Naoyoshi.

1981-01-01

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

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)

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

International Nuclear Information System (INIS)

Brachet, Alain.

1981-01-01

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

Browns Ferry Nuclear Plant: variation in test intervals for high-pressure coolant injection (HPCI) system

International Nuclear Information System (INIS)

Christie, R.F.; Stetkar, J.W.

1985-01-01

The change in availability of the high-pressure coolant injection system (HPCIS) due to a change in pump and valve test interval from monthly to quarterly was analyzed. This analysis started by using the HPCIS base line evaluation produced as part of the Browns Ferry Nuclear Plant (BFN) Probabilistic Risk Assessment (PRA). The base line evaluation showed that the dominant contributors to the unavailability of the HPCI system are hardware failures and the resultant downtime for unscheduled maintenance. The effect of changing the pump and valve test interval from monthly to quarterly was analyzed by considering the system unavailability due to hardware failures, the unavailability due to testing, and the unavailability due to human errors that potentially could occur during testing. The magnitude of the changes in unavailability affected by the change in test interval are discussed. The analysis showed a small increase in the availability of the HPCIS to respond to loss of coolant accidents (LOCAs) and a small decrease in the availability of the HPCIS to respond to transients which require HPCIS actuation. In summary, the increase in test interval from monthly to quarterly does not significantly impact the overall HPCIS availability

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

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

International Nuclear Information System (INIS)

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

1991-08-01

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-07-01

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

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

International Nuclear Information System (INIS)

Elina Syrjaelahti; Anitta Haemaelaeinen

2005-01-01

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

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)

Numerical FEM Analyses of primary coolant system at NPP Temelin

International Nuclear Information System (INIS)

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

2003-01-01

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

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

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.

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

Device for preventing coolant outflow in a reactor

International Nuclear Information System (INIS)

Nemoto, Kiyomitsu; Mochizuki, Keiichi.

1975-01-01

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

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

International Nuclear Information System (INIS)

Tira, Z.

1993-02-01

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

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)

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

Wesley, D.A.

1991-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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.

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

Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

Directory of Open Access Journals (Sweden)

Navid Bozorgan

2015-07-01

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

High temperature semiconductor diode laser pumps for high energy laser applications

Science.gov (United States)

Campbell, Jenna; Semenic, Tadej; Guinn, Keith; Leisher, Paul O.; Bhunia, Avijit; Mashanovitch, Milan; Renner, Daniel

2018-02-01

Existing thermal management technologies for diode laser pumps place a significant load on the size, weight and power consumption of High Power Solid State and Fiber Laser systems, thus making current laser systems very large, heavy, and inefficient in many important practical applications. To mitigate this thermal management burden, it is desirable for diode pumps to operate efficiently at high heat sink temperatures. In this work, we have developed a scalable cooling architecture, based on jet-impingement technology with industrial coolant, for efficient cooling of diode laser bars. We have demonstrated 60% electrical-to-optical efficiency from a 9xx nm two-bar laser stack operating with propylene-glycolwater coolant, at 50 °C coolant temperature. To our knowledge, this is the highest efficiency achieved from a diode stack using 50 °C industrial fluid coolant. The output power is greater than 100 W per bar. Stacks with additional laser bars are currently in development, as this cooler architecture is scalable to a 1 kW system. This work will enable compact and robust fiber-coupled diode pump modules for high energy laser applications.

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

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Loss-of-coolant and loss-of-flow accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

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

1994-07-01

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

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)

LWR primary coolant pipe rupture test rig

International Nuclear Information System (INIS)

Yoshitoshi, Shyoji

1978-01-01

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

Multiphysics Modeling of an Annular Linear Induction Pump With Applications to Space Nuclear Power Systems

Science.gov (United States)

Kilbane, J.; Polzin, K. A.

2014-01-01

An annular linear induction pump (ALIP) that could be used for circulating liquid-metal coolant in a fission surface power reactor system is modeled in the present work using the computational COMSOL Multiphysics package. The pump is modeled using a two-dimensional, axisymmetric geometry and solved under conditions similar to those used during experimental pump testing. Real, nonlinear, temperature-dependent material properties can be incorporated into the model for both the electrically-conducting working fluid in the pump (NaK-78) and structural components of the pump. The intricate three-phase coil configuration of the pump is implemented in the model to produce an axially-traveling magnetic wave that is qualitatively similar to the measured magnetic wave. The model qualitatively captures the expected feature of a peak in efficiency as a function of flow rate.

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

Science.gov (United States)

Rathbun, R. J.

1966-01-01

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

LOFT pump speed controller stability and accuracy analysis

International Nuclear Information System (INIS)

Good, R.R.

1978-01-01

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

AGING MANAGMENT OF REACTOR COOLANT SYSTEM MECHANICAL COMPONENTS FOR LICENSE RENEWAL

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Science.gov (United States)

Morrison, Russell H.; Holt, Mike

2005-01-01

The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

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

International Nuclear Information System (INIS)

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

1981-01-01

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

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.)

THE PROBLEM OF ENERGY EFFICIENCY OF THE GEOTHERMAL CIRCULATION SYSTEM IN DIFFERENT MODES OF REINJECTION OF THE COOLANT

OpenAIRE

D. K. Djavatov; A. A. Azizov

2017-01-01

Aim. Advanced technologies are crucial for widespread use of geothermal energy to ensure its competitiveness with conventional forms of energy. To date, the basis for the development of geothermal energy is the technology of extracting the heat transfer fluids from the subsoil. There are the following ways to extract the coolant: freeflow; pumping and circular methods. Of greatest interest is the technology to harness the geothermal energy based on geothermal circulatory system (GCS). There i...

Report on measurements at the pump Avala - Annex 7

International Nuclear Information System (INIS)

Nikolic, M.

1963-01-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Liquid metal coolants for fusion-fission hybrid system: A neutronic analysis

Energy Technology Data Exchange (ETDEWEB)

Marques, Renato V.A.; Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L., E-mail: claubia@nuclear.ufmg.br [Universidade de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Barros, Graiciany P. [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

Based on a work already published by the UFMG Nuclear Engineering Department, it was suggested to use different coolant materials in a fusion-fission system after a fuel burnup simulation, including that one used in reference work. The goal is to compare the neutron parameters, such as the effect multiplication factor and actinide amounts in transmutation layer, for each used coolant and find the best(s) coolant material(s) to be applied in the considered system. Results indicate that the lead and lead-bismuth coolant are the most suitable choices to be applied to cool the system. (author)

THE PROBLEM OF ENERGY EFFICIENCY OF THE GEOTHERMAL CIRCULATION SYSTEM IN DIFFERENT MODES OF REINJECTION OF THE COOLANT

Directory of Open Access Journals (Sweden)

D. K. Djavatov

2017-01-01

Full Text Available Aim. Advanced technologies are crucial for widespread use of geothermal energy to ensure its competitiveness with conventional forms of energy. To date, the basis for the development of geothermal energy is the technology of extracting the heat transfer fluids from the subsoil. There are the following ways to extract the coolant: freeflow; pumping and circular methods. Of greatest interest is the technology to harness the geothermal energy based on geothermal circulatory system (GCS. There is the problem of the right choice of technological parameters for geothermal systems to ensure their effective functioning.Methods. We consider the development of geothermal energy technology based on geothermal circulatory system, as this technology solves the dumping of the waste water containing environmentally harmful substances. In addition to the environmental issues, this technology makes it possible to intensify the process of production and the degree of extraction of thermal resources, which significantly increases the potential for geothermal heat resources in terms of the fuel and energy balance.Findings. Were carried out optimization calculations for Ternairsky deposits of thermal waters. In the calculations, was taken into account the temperature dependence of important characteristics, such as the density and heat capacity of the coolant.Conclusions. There is the critical temperature of the coolant injected, depending on the flow rate and the diameter of the well, ensuring the effective functioning of the geothermal circulatory systems. 

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

Energy Technology Data Exchange (ETDEWEB)

Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

2017-08-15

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

Making effective use of rod pumping systems in coalbed methane applications

Energy Technology Data Exchange (ETDEWEB)

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

2003-07-01

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

Primary coolant circuits in FBR type reactors

International Nuclear Information System (INIS)

Kutani, Masushiro.

1985-01-01

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

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

Experiment data report for LOFT large-break loss-of-coolant experiment L2-5

International Nuclear Information System (INIS)

Bayless, P.D.; Divine, J.M.

1982-08-01

Selected pertinent and uninterpreted data from the third nuclear large break loss-of-coolant experiment (Experiment L2-5) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large [approx. 1000 MW(e)] commercial PWR operations. Experiment L2-5 simulated a double-ended offset shear of a cold leg in the primary coolant system. The primary coolant pumps were tripped within 1 s after the break initiation, simulating a loss of site power. Consistent with the loss of power, the starting of the high- and low-pressure injection systems was delayed. The peak fuel rod cladding temperature achieved was 1078 +- 13 K. The emergency core cooling system re-covered the core and quenched the cladding. No evidence of core damage was detected

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

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

Application of damage function analysis to reactor coolant circuits

International Nuclear Information System (INIS)

MacDonald, D.D.

2002-01-01

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

Application of damage function analysis to reactor coolant circuits

Energy Technology Data Exchange (ETDEWEB)

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

2002-07-01

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

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.)

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

Directory of Open Access Journals (Sweden)

Hwang Bae

2017-08-01

Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

Responses to Small Break Loss of Coolant Accidents for SMART

International Nuclear Information System (INIS)

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

2004-01-01

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

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

Science.gov (United States)

Morrison, Russell H.; Holt, Mike

2006-01-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Coolant cleanup method in a nuclear reactor

International Nuclear Information System (INIS)

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

1983-01-01

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

IEA-R1 primary and secondary coolant piping systems coupled stress analysis

International Nuclear Information System (INIS)

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

2013-01-01

The aim of this work is to perform the stress analysis of a coupled primary and secondary piping system of the IEA-R1 based on tridimensional model, taking into account the as built conditions. The nuclear research reactor IEA-R1 is a pool type reactor projected by Babcox-Willcox, which is operated by IPEN since 1957. The operation to 5 MW power limit was only possible after the conduction of life management and modernization programs in the last two decades. In these programs the components of the coolant systems, which are responsible for the water circulation into the reactor core to remove the heat generated inside it, were almost totally refurbished. The changes in the primary and secondary systems, mainly the replacement of pump and heat-exchanger, implied in piping layout modifications, and, therefore, the stress condition of the piping systems had to be reanalyzed. In this paper the structural stress assessment of the coupled primary and secondary piping systems is presented and the final results are discussed. (author)

Design and fabrication of magnetic coolant filter

Science.gov (United States)

Prashanth, B. N.

2017-07-01

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

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.

Performance evaluation of an integrated automotive air conditioning and heat pump system

International Nuclear Information System (INIS)

Hosoz, M.; Direk, M.

2006-01-01

This study deals with the performance characteristics of an R134a automotive air conditioning system capable of operating as an air-to-air heat pump using ambient air as a heat source. For this aim, an experimental analysis has been performed on a plant made up of original components from an automobile air conditioning system and some extra equipment employed to operate the system in the reverse direction. The system has been tested in the air conditioning and heat pump modes by varying the compressor speed and air temperatures at the inlets of the indoor and outdoor coils. Evaluation of the data gathered in steady state test runs has shown the effects of the operating conditions on the capacity, coefficient of performance, compressor discharge temperature and the rate of exergy destroyed by each component of the system for both operation modes. It has been observed that the heat pump operation provides adequate heating only in mild weather conditions, and the heating capacity drops sharply with decreasing outdoor temperature. However, compared with the air conditioning operation, the heat pump operation usually yields a higher coefficient of performance and a lower rate of exergy destruction per unit capacity. It is also possible to improve the heating mode performance of the system by redesigning the indoor coil, using another refrigerant with a higher heat rejection rate in the condenser and employing a better heat source such as the engine coolant or exhaust gases

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

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

International Nuclear Information System (INIS)

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

1981-08-01

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

Assessment of the integration of a He-cooled divertor system in the power conversion system for the dual-coolant blanket concept (TW2-TRP-PPCS12D8)

International Nuclear Information System (INIS)

Norajitra, P.; Kruessmann, R.; Malang, S.; Reimann, G.

2002-12-01

Application of a helium-cooled divertor together with the dual-coolant blanket concept is considered favourable for achieving a high thermal efficiency of the power plant due to its relatively high coolant outlet temperature. A new FZK He-cooled modular divertor concept with integrated pin arrays (HEMP) is introduced. Its main features and function are described in detail. The result of the thermalhydraulic analysis shows that the HEMP divertor concept has the potential of resisting, a heat flow density of at least 10-15 MW/m 2 at a reachable heat transfer coefficient of approx. 60 kW/m 2 K and a reasonable pumping power. Integration of this divertor concept into the power conversion system using a closed Brayton gas turbine system with three-stage compression leads to a net efficiency of the blanket/divertor cycle of about 43%. (orig.)

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

Performance Analysis of Thermoelectric Based Automotive Waste Heat Recovery System with Nanofluid Coolant

Directory of Open Access Journals (Sweden)

Zhi Li

2017-09-01

Full Text Available Output performance of a thermoelectric-based automotive waste heat recovery system with a nanofluid coolant is analyzed in this study. Comparison between Cu-Ethylene glycol (Cu-EG nanofluid coolant and ethylene glycol with water (EG-W coolant under equal mass flow rate indicates that Cu-EG nanofluid as a coolant can effectively improve power output and thermoelectric conversion efficiency for the system. Power output enhancement for a 3% concentration of nanofluid is 2.5–8 W (12.65–13.95% compared to EG-Water when inlet temperature of exhaust varies within 500–710 K. The increase of nanofluid concentration within a realizable range (6% has positive effect on output performance of the system. Study on the relationship between total area of thermoelectric modules (TEMs and output performance of the system indicates that optimal total area of TEMs exists for maximizing output performance of the system. Cu-EG nanofluid as coolant can decrease optimal total area of TEMs compared with EG-W, which will bring significant advantages for the optimization and arrangement of TEMs whether the system space is sufficient or not. Moreover, power output enhancement under Cu-EG nanofluid coolant is larger than that of EG-W coolant due to the increase of hot side heat transfer coefficient of TEMs.

Experimental investigation of biomimetic self-pumping and self-adaptive transpiration cooling.

Science.gov (United States)

Jiang, Pei-Xue; Huang, Gan; Zhu, Yinhai; Xu, Ruina; Liao, Zhiyuan; Lu, Taojie

2017-09-01

Transpiration cooling is an effective way to protect high heat flux walls. However, the pumps for the transpiration cooling system make the system more complex and increase the load, which is a huge challenge for practical applications. A biomimetic self-pumping transpiration cooling system was developed inspired by the process of trees transpiration that has no pumps. An experimental investigation showed that the water coolant automatically flowed from the water tank to the hot surface with a height difference of 80 mm without any pumps. A self-adaptive transpiration cooling system was then developed based on this mechanism. The system effectively cooled the hot surface with the surface temperature kept to about 373 K when the heating flame temperature was 1639 K and the heat flux was about 0.42 MW m -2 . The cooling efficiency reached 94.5%. The coolant mass flow rate adaptively increased with increasing flame heat flux from 0.24 MW m -2 to 0.42 MW m -2 while the cooled surface temperature stayed around 373 K. Schlieren pictures showed a protective steam layer on the hot surface which blocked the flame heat flux to the hot surface. The protective steam layer thickness also increased with increasing heat flux.

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

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

Directory of Open Access Journals (Sweden)

S. K. Pavlov

2014-01-01

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

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

Cooling device for leaking fluid from a centrifugal pump

International Nuclear Information System (INIS)

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

1978-01-01

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

Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

International Nuclear Information System (INIS)

Fletcher, C.D.; Callow, R.A.

1989-01-01

The USNRC recently identified a possible safety concern for PWR's. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: 1) describes the important aspects of the problem, 2) provides an initial analysis of the consequences, and 3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of WE design, the concern is greatest for those plants. There is less concern for most plants of CE design, and likely no concern for plants of BW design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a WE PWR. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed. (orig./HP)

Photovoltaic pump systems

Science.gov (United States)

Klockgether, J.; Kiessling, K. P.

1983-09-01

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

System of Thermal Balance Maintenance in Modern Test Benches for Centrifugal Pumps

Directory of Open Access Journals (Sweden)

A. I. Petrov

2015-01-01

Full Text Available The article “Systems of the heat balance maintenance in modern test benches for centrifugal pumps” makes the case to include cooling systems of a working fluid (heat setting in test bench for impeller pumps. It briefly summarizes an experience of bench building to test centrifugal pumps, developed at the BMSTU Department E-10 over the last 10 years. The article gives the formulas and the algorithm to calculate the heat capacity of different types of impeller pumps when tested at the bench as ell as to determine the heating time of the liquid in the bench without external cooling. Based on analysis of the power balance of a centrifugal pump, it is shown that about 90% of the pump unit-consumed electric power in terminals is used for heating up the working fluid in the loop of the test bench. The article gives examples of elementary heat calculation of the pump operation within the test bench. It presents the main types of systems to maintain thermal balance, their advantages, disadvantages and possible applications. The cooling system schemes for open and closed version of the benches both with built-in and with an independent cooling circuit are analysed. The paper separately considers options of such systems for large benches using the cooling tower as a cooling device in the loop, and to test the pumps using the hydraulic fluids other than water, including those at high temperatures of working fluids; in the latter case a diagram of dual-circuit cooling system "liquid-liquid-air" is shown. The paper depicts a necessity to use ethylene glycol coolant in the two-loop cooling bench. It provides an example of combining the functions of cooling and filtration in a single cooling circuit. Criteria for effectiveness of these systems are stated. Possible ways for developing systems to maintain a thermal balance, modern methods of regulation and control are described. In particular, the paper shows the efficiency of frequency control of the

Real-time analysis and display of reactor system mass inventory

International Nuclear Information System (INIS)

Dao, L.T.; Meachum, T.R.

1982-01-01

A mass inventory system (MIS) to evaluate, in real-time, the coolant distribution within the primary coolant system of the Loss-of-Fluid Test (LOFT) reactor has been developed. The computer-based system calculates and displays the coolant levels by two methods: using level measurements and performing a mass balance. The MIS is designed to provide up-to-date, intelligible information on the coolant distribution during any LOFT experiment. During LOFT experiments in which the primary coolant pumps are on, the method also provides void fraction information and the anticipated liquid level in the reactor vessel should the pumps be turned off

Artificial heart system thermal converter and blood pump component research and development

International Nuclear Information System (INIS)

Pouchot, W.D.; Bifano, N.J.; Hanson, J.P.

1975-01-01

A bench model version of a nuclear-powered artificial heart system to be used as a replacement for the natural heart was constructed and tested as a part of a broader U. S. ERDA program. The objective of the broader program has been to develop a prototype of a fully implantable nuclear-powered total artificial heart system powered by the thermal energy of plutonium-238 and having minimum weight and volume and a minimum life of ten years. As a forward step in this broader program, component research and development has been carried out directed towards a fully implantable and advanced version of the bench model (IVBM). Some of the results of the component research and development effort on a Stirling engine, blood pump drive mechanisms, and coupling mechanisms are presented. The Stirling-mechanical system under development is shown. There are three major subassemblies: the thermal converter, the coupling mechanism, and the blood pump drive mechanism. The thermal converter uses a Stirling cycle to convert the heat of the plutonium-238 fueled heat source to a rotary shaft power output. The coupling mechanism changes the orientation of the output shaft by 90 degrees and transmits the pumping power by wire-wound core flexible shafting to the pumping mechanism. The coupling mechanism also provides routing of the coolant lines which carry the cycle waste heat from the thermal converter to the blood pump. The change in orientation of the thermal converter output shaft is for convenience in implanting in a calf. This orientation of thermal converter to blood pump seemed to give the best overall system fit in a calf based on fit trials with wooden models in a calf cadaver

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Review on research of small break loss of coolant accident

International Nuclear Information System (INIS)

Bo Jinhai; Wang Fei

1998-01-01

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

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

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

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

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

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1979-08-01

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

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

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.)

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

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.

Research on coolant radiochemistry

International Nuclear Information System (INIS)

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

2004-04-01

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

Experiences in design up-gradation of mechanical seal cooling scheme of Dhruva PHT pumps

International Nuclear Information System (INIS)

Balakrishnan, K.T.P.

2002-01-01

Full text: Dhruva is a natural uranium fuelled high flux research reactor. Heavy water is used as coolant, moderator and reflector. Heat from the heavy water coolant is removed in heat exchangers by demineralised water. The heavy water coolant is re-circulated between the reactor core and the heat exchangers in three separate loops by three main coolant pumps (MCPs). The MCPs are high capacity centrifugal pumps and are rated for continuous service. The mechanical seal of the pump prevents leakage of the process fluid, which is heavy water, through the pump shaft. Continuous operation of the pump results in the heating up of the seal and necessitates sustained cooling. An integral cooling provision is made by tapping a 15 NB line from the discharge volute of the pump and feeding the process fluid itself as coolant to the seal. A non-indicating type flow-sensing device monitors flow through this line. Limiting values of flow are set and annunciated by a pair of magnetic reed type relays. This cooling line was a built in feature of the pumps as supplied by the manufacturer. This arrangement had the following inherent limitations: 1. There was no on line indication of the coolant flow. 2. The reed type magnetic relays initiated pump trips by spurious actuation, resulting in the interruption of reactor operation. Servicing a faulty flow switch involved lengthy procedures and necessitated draining, filling and venting of the pump. This entailed extended plant outages. Close proximity of these flow switches to a highly radioactive piping element imposed severe restrictions on the planned maintenance activity on them. Efforts were made to provide a suitable alternate cooling and flow measurement scheme to overcome the above-mentioned limitations. After evaluating the relative merits and demerits of several schemes, a turbine type flow sensor, on a modified cooling line was selected as the most suitable alternative. The alternate seal-cooling scheme was implemented for all

Experiences in design up-gradation of mechanical seal cooling scheme of Dhruva PHT pumps

International Nuclear Information System (INIS)

Balakrishnan, K.T.P.; Bharathan, R.

2002-01-01

Full text: Dhruva is a natural uranium fuelled high flux research reactor. Heavy water is used as coolant, moderator and reflector. Heat from the heavy water coolant is removed in heat exchangers by demineralised water. The heavy water coolant is re-circulated between the reactor core and the heat exchangers in three separate loops by three main coolant pumps (MCPs). The MCPs are high capacity centrifugal pumps and are rated for continuous service. The mechanical seal of the pump prevents leakage of the process fluid, which is heavy water, through the pump shaft. Continuous operation of the pump results in the heating up of the seal and necessitates sustained cooling. An integral cooling provision is made by tapping a 15 NB line from the discharge volute of the pump and feeding the process fluid itself as coolant to the seal. A non-indicating type flow-sensing device monitors flow through this line. Limiting values of flow are set and annunciated by a pair of magnetic reed type relays. This cooling line was a built in feature of the pumps as supplied by the manufacturer. This arrangement had the following inherent limitations : 1. There was no on line indication of the coolant flow. 2. The reed type magnetic relays initiated pump trips by spurious actuation, resulting in the interruption of reactor operation. Servicing a faulty flow switch involved lengthy procedures and necessitated draining, filling and venting of the pump. This entailed extended plant outages. Close proximity of these flow switches to a highly radioactive piping element imposed severe restrictions on the planned maintenance activity on them. Efforts were made to provide a suitable alternate cooling and flow measurement scheme to overcome the above-mentioned limitations. After evaluating the relative merits and demerits of several schemes, a turbine type flow sensor, on a modified cooling line was selected as the most suitable alternative. The alternate seal-cooling scheme was implemented for all

THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

International Nuclear Information System (INIS)

Muramatsu, Ken; Akimoto, Masayuki

1982-08-01

THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

Design of automotive engine coolant hoses

Directory of Open Access Journals (Sweden)

Hrishikesh D BACHCHHAV

2018-03-01

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

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.)

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

Science.gov (United States)

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

2015-08-01

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

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

Development of motors and drives for main coolant pump and CEDM

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

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

Coolant Design System for Liquid Propellant Aerospike Engines

Science.gov (United States)

McConnell, Miranda; Branam, Richard

2015-11-01

Liquid propellant rocket engines burn at incredibly high temperatures making it difficult to design an effective coolant system. These particular engines prove to be extremely useful by powering the rocket with a variable thrust that is ideal for space travel. When combined with aerospike engine nozzles, which provide maximum thrust efficiency, this class of rockets offers a promising future for rocketry. In order to troubleshoot the problems that high combustion chamber temperatures pose, this research took a computational approach to heat analysis. Chambers milled into the combustion chamber walls, lined by a copper cover, were tested for their efficiency in cooling the hot copper wall. Various aspect ratios and coolants were explored for the maximum wall temperature by developing our own MATLAB code. The code uses a nodal temperature analysis with conduction and convection equations and assumes no internal heat generation. This heat transfer research will show oxygen is a better coolant than water, and higher aspect ratios are less efficient at cooling. This project funded by NSF REU Grant 1358991.

Reactor coolant flow measurements at Point Lepreau

International Nuclear Information System (INIS)

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

1996-01-01

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

Nonlinear dynamic analysis of nuclear reactor primary coolant systems

International Nuclear Information System (INIS)

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

1979-01-01

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

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 for impurity control of the plasma of a fusion reactor has a major impact on the design of the tritium systems. To achieve a sustained burn, conventional limiters and divertors remove large quantities of unburnt tritium and deuterium from the plasma which must be then recycled using a plasma processing system. The self-pumped limiter which does not remove the hydrogen species, does not require any plasma processing equipment. The blanket system and the coolant processing systems acquire greater importance with the use of this unconventional impurity control system. 3 refs., 2 figs

Natural circulation in reactor coolant system

International Nuclear Information System (INIS)

Han, J.T.

1987-01-01

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

Use of microPCM fluids as enhanced liquid coolants in automotive EV and HEV vehicles. Final report

Energy Technology Data Exchange (ETDEWEB)

Mulligan, James C.; Gould, Richard D.

2001-10-31

Proof-of-concept experiments using a specific microPCM fluid that potentially can have an impact on the thermal management of automotive EV and HEV systems have been conducted. Samples of nominally 20-micron diameter microencapsulated octacosane and glycol/water coolant were prepared for testing. The melting/freezing characteristics of the fluid, as well as the viscosity, were determined. A bench scale pumped-loop thermal system was used to determine heat transfer coefficients and wall temperatures in the source heat exchanged. Comparisons were made which illustrate the enhancements of thermal performance, reductions of pumping power, and increases of heat transfer which occur with the microPCM fluid.

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

International Nuclear Information System (INIS)

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

1975-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

TFTR ultrahigh-vacuum pumping system incorporating mercury diffusion pumps

International Nuclear Information System (INIS)

Sink, D.A.; Sniderman, M.

1976-06-01

The TFTR vacuum vessel will have a system of four 61 cm diameter mercury diffusion pumps to provide a base pressure in the 10 -8 to 10 -9 Torr range as well as a low impurity level within the vessel. The system, called the Torus Vacuum Pumping System (TVPS), will be employed with the aid of an occasional 250 0 C bakeout in situ as well as periodic applications of aggressive discharge cleaning. The TVPS is an ultrahigh-vacuum (UHV) system using no elastomers as well as being a closed system with respect to tritium or any tritiated gases. The backing system employing approximately 75 all-metal isolation valves is designed with the features of redundancy and flexibility employed in a variety of ways to meet the fundamental requirements and functions enumerated for the TVPS. Since the design, is one which is a modification of the conceptual design of the TVPS, those features which have changed are discussed. Calculations are presented for the major performance parameters anticipated for the TVPS and include conductances, effective pumping speeds, base pressures, operating parameters, getter pump parameters, and calculations of time constants associated with leak checking. Modifications in the vacuum pumping system for the guard regions on the twelve bellows sections are presented so that it is compatible with the main TVPS. The bellows pumping system consists of a mechanical pump unit, a zirconium aluminum getter pump unit and a residual gas analyzer. The control and management of the TVPS is described with particular attention given to providing both manual and automatic control at a local station and at the TFTR Central Control. Such operations as testing, maintenance, leak checking, startup, bakeout, and various other operations are considered in some detail. Various aspects related to normal pulsing, discharge cleaning, non-tritium operations and tritium operations are also taken into consideration. A cost estimate is presented

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

Pihowicz, W.

1999-01-01

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

Method for controlling a coolant liquid surface of cooling system instruments in an atomic power plant

International Nuclear Information System (INIS)

Monta, Kazuo.

1974-01-01

Object: To prevent coolant inventory within a cooling system loop in an atomic power plant from being varied depending on loads thereby relieving restriction of varied speed of coolant flow rate to lowering of a liquid surface due to short in coolant. Structure: Instruments such as a superheater, an evaporator, and the like, which constitute a cooling system loop in an atomic power plant, have a plurality of free liquid surface of coolant. Portions whose liquid surface is controlled and portions whose liquid surface is varied are adjusted in cross-sectional area so that the sum total of variation in coolant inventory in an instrument such as a superheater provided with an annulus portion in the center thereof and an inner cylindrical portion and a down-comer in the side thereof comes equal to that of variation in coolant inventory in an instrument such as an evaporator similar to the superheater. which is provided with an overflow pipe in its inner cylindrical portion or down-comer, thereby minimizing variation in coolant inventory of the entire coolant due to loads thus minimizing variation in varied speed of the coolant. (Kamimura, M.)

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

Selection of an Alternate Biocide for the ISS Internal Thermal Control System Coolant, Phase 2

Science.gov (United States)

Wilson, Mark E.; Cole, Harold; Weir, Natalee; Oehler, Bill; Steele, John; Varsik, Jerry; Lukens, Clark

2004-01-01

The ISS (International Space Station) ITCS (Internal Thermal Control System) includes two internal coolant loops that utilize an aqueous based coolant for heat transfer. A silver salt biocide had previously been utilized as an additive in the coolant formulation to control the growth and proliferation of microorganisms within the coolant loops. Ground-based and in-flight testing demonstrated that the silver salt was rapidly depleted, and did not act as an effective long-term biocide. Efforts to select an optimal alternate biocide for the ITCS coolant application have been underway and are now in the final stages. An extensive evaluation of biocides was conducted to down-select to several candidates for test trials and was reported on previously. Criteria for that down-select included: the need for safe, non-intrusive implementation and operation in a functioning system; the ability to control existing planktonic and biofilm residing microorganisms; a negligible impact on system-wetted materials of construction; and a negligible reactivity with existing coolant additives. Candidate testing to provide data for the selection of an optimal alternate biocide is now in the final stages. That testing has included rapid biocide effectiveness screening using Biolog MT2 plates to determine minimum inhibitory concentration (amount that will inhibit visible growth of microorganisms), time kill studies to determine the exposure time required to completely eliminate organism growth, materials compatibility exposure evaluations, coolant compatibility studies, and bench-top simulated coolant testing. This paper reports the current status of the effort to select an alternate biocide for the ISS ITCS coolant. The results of various test results to select the optimal candidate are presented.

Pump system characterization and reliability enhancement

International Nuclear Information System (INIS)

Staunton, R.H.

1998-01-01

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

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

International Nuclear Information System (INIS)

Puri, R.K.; Singh, M.

2006-01-01

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

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

International Nuclear Information System (INIS)

Lee, C.B.

2002-01-01

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

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

International Nuclear Information System (INIS)

Omahen, P.; Gubina, F.

1992-01-01

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

SNR coolant system components

International Nuclear Information System (INIS)

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

1976-01-01

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

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

DEFF Research Database (Denmark)

Asadi, Meisam; Asadi, Amin; Aberoumand, Sadegh

2018-01-01

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

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

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)

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

International Nuclear Information System (INIS)

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

1979-10-01

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

Primary coolant feed and bleed operating regions for the Midland Plant

International Nuclear Information System (INIS)

Tsai, M.S.

1985-01-01

Operating regions for primary coolant feed and bleed cooling are developed for the Midland Plant using core decay heat, the high-pressure injection (HPI) system capacity, and flow rate relief through the power-operated relief valve (PORV). This mode of cooling is used for accident scenarios in which the normal core cooling means of a nuclear power plant is lost because of loss of water inventory in the steam generators. The HPI flow is based on the capacities of one and two pumps. Saturated steam, saturated water, and subcooled water are considered to be possible states of the fluid being relieved through the PORV. In estimating the PORV relief rate, flow equations are derived from the Electric Power Research Institute test data obtained from the same model and size valve that is used in the Midland Plant. For easy reference by operators, the operating region is displayed on a plane of reactor coolant system pressure and temperature. The technique developed for the Midland Plant provides a convenient method for examining the feed and bleed cooling capability for a nuclear power plant that employs a pressurized water reactor system

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

DEFF Research Database (Denmark)

Asadi, Meisam; Asadi, Amin; Aberoumand, Sadegh

2018-01-01

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

Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

OpenAIRE

Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

2017-01-01

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are s...

N13 - based reactor coolant pressure boundary leakage system

International Nuclear Information System (INIS)

Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

1980-05-01

A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

Vision system for precision alignment of coolant channels

International Nuclear Information System (INIS)

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

1997-01-01

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

Single-phase sodium pump model for LMFBR thermal-hydraulic analysis

International Nuclear Information System (INIS)

Madni, I.K.; Cazzoli, E.G.; Agrawal, A.K.

1979-01-01

A single-phase, homologous pump model has been developed for simulation of safety-related transients in LMFBR systems. Pump characteristics are modeled by homologous head and torque relations encompassing all regimes of operation. These relations were derived from independent model test results with a centrifugal pump of specific speed equal to 35 (SI units) or 1800 (gpm units), and are used to analyze the steady-state and transient behavior of sodium pumps in a number of LMFBR plants. Characteristic coefficients for the polynomials in all operational regimes are provided in a tabular form. The speed and flow dependence of head is included through solutions of the impeller and coolant dynamic equations. Results show the model to yield excellent agreement with experimental data in sodium for the FFTF prototype pump, and with vendor calculations for the CRBR pump. A sample pipe rupture calculation is also performed to demonstrate the necessity for modeling the complete pump characteristics

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.

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting systems. 50.46a Section 50.46a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND... function of these systems. High point vents are not required for the tubes in U-tube steam generators...

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

Experience on vibration analysis of primary coolant pumps in Cirus

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Analysis of a simulated small break in the semiscale system under loss-of-coolant accident conditions

International Nuclear Information System (INIS)

Cartmill, C.E.

1978-01-01

The Semiscale Mod-1 experimental program conducted by EG and G Idaho, Inc., is part of the overall U.S. Nuclear Regulatory Commission (NRC) and Department of Energy (DOE) sponsored research and development program to investigate the behavior of the pressurized water reactor (PWR) system during an hypothesized loss-of-coolant accident (LOCA). The Semiscale Mod-1 program is intended to provide transient thermal-hydraulic data from a simulated LOCA using a small-scale experimental nonnuclear system. The Semiscale Mod-1 program is a major contributor of experimental data that provide a means of evaluating the adequacy of overall system analytical models as well as the models of the individual system components. Selected experimental data produced by this program will also be used to aid other DOE and NRC sponsored experimental programs, such as the Loss-of-Fluid Test (LOFT) program in optimizing test series, selecting test parameters, and evaluating test results. The Semiscale Mod-1 tests are performed with an experimental system which simulates the principal features of a nuclear plant but which is smaller in volume. Nuclear heating is simulated in the tests by a core composed of an array of electrically heated rods. The core is contained in a pressure vessel which also includes a downcomer, lower plenum, and upper plenum. The Semiscale system piping is arranged such that the intact loop represents three loops of a four-loop nuclear plant, and the broken loop represents the fourth loop. In the present configuration the intact loop contains an active steam generator and pump, and the broken loop contains passive simulators for the steam generator and pump

Generic study on the relation between contamination if primary coolants and occupational radiation exposure in nuclear power plants with PWR. Final report; Generische Studie zum Zusammenhang zwischen Kontamination von Primaerkreislaufmedien und beruflicher Strahlenexposition bei Kernkraftwerken mit Druckwasserreaktor. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Artmann, Andreas; Bruhn, Gerd; Schneider, Sebastian [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany); Strub, Erik [Koeln Univ. (Germany)

2016-01-15

A generic model for the primary cooling system contamination in pressurized water reactors and the resulting radiological consequences has been developed. The functional capability was demonstrated by means of three examples concerning manipulation procedures during revision outages. Activities at the main reactor coolant pumps were studied and the influence of the coolant contamination on the resulting dose rates and collective doses were calculated. The effect of a Co-90 hot spot in a more remote area on the radiation exposure during the specific action at the reactor pumps was considered.

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.)

Damages on pumps and systems the handbook for the operation of centrifugal pumps

CERN Document Server

Merkle, Thomas

2014-01-01

Damage on Pumps and Systems. The Handbook for the Operation of Centrifugal Pumps offers a combination of the theoretical basics and practical experience for the operation of circulation pumps in the engineering industry. Centrifugal pumps and systems are extremely vulnerable to damage from a variety of causes, but the resulting breakdown can be prevented by ensuring that these pumps and systems are operated properly. This book provides a total overview of operating centrifugal pumps, including condition monitoring, preventive maintenance, life cycle costs, energy savings and economic aspects. Extra emphasis is given to the potential damage to these pumps and systems, and what can be done to prevent breakdown. Addresses specific issues about pumping of metal chips, sand, abrasive dust and other solids in fluidsEmphasis on economic and efficiency aspects of predictive maintenance and condition monitoring Uses life cycle costs (LCC) to evaluate and calculate the costs of pumping systems

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-15

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

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.)

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)

Coolant processing device for nuclear reactor

International Nuclear Information System (INIS)

Kizawa, Hideo; Funakoshi, Toshio; Izumoji, Yoshiaki

1981-01-01

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

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

International Nuclear Information System (INIS)

2003-01-01

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

Conceptual design of primary coolant purification system using cylindrical membrane for nuclear energy system base on HTGR

International Nuclear Information System (INIS)

Piping Supriatna

2011-01-01

The recent progress of reactor technology design for next generation reactor will be implemented on cogeneration reactor, which the aim of reactor operation not only for generating electrical energy, but also for other application like desalination, industrial manufacturing process, hydrogen production, Enhanced Oil Recovery (EOR), etc. The cogeneration reactor concept developed for generate energy effectively, efficiently and sustainable, which reserve of uranium and thorium nuclear fuel for cogeneration reactor is supply able for world energy demand until next thousand years. The cogeneration reactor produce temperature output higher than commonly Nuclear Power Plant (NPP), and need special Heat Exchanger with helium gas as coolant. In order to preserve heat transfer with high efficiency, constant purity of the gas must be maintained as well as possible, especially contamination from its impurities. In this research has been designed modeling and assessment of primary coolant gas purification system with purify and fill up helium gas continuously, by using Cylindrical Helium Splitting Membrane and helium gas inventory system. The result of flow rate helium assessment for the purification system is 0.844x10 -3 kg/sec, where helium flow rate of reactor primary coolant is 120 kg/sec. The result of study show that the Primary Coolant Gas Purification System is enable to be implemented on Cogeneration Reactor HTGR200C. (author)

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

International Nuclear Information System (INIS)

Almeida, G.L. de.

1986-01-01

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

Engine restart aid

Science.gov (United States)

Fedewa, Andrew

2018-01-09

A system is disclosed comprising an engine having coolant passages defined therethrough, a first coolant pump, and a first radiator. The system additionally comprises a second coolant pump, a second radiator, and a liquid-to-air heat exchanger configured to condition the temperature of intake air to the engine. The system further includes a coolant valve means. For a first configuration of the coolant valve means the first coolant pump is configured to urge coolant through the coolant passages in the engine and through the first radiator, and the second coolant pump is configured to urge coolant through the liquid-to-air heat exchanger and through the second radiator. For a second configuration of the coolant valve means the second coolant pump is configured to urge coolant through the coolant passages in the engine and through the liquid-to-air heat exchanger. A method for controlling the system is also disclosed.

Flow pumping system for physiological waveforms.

Science.gov (United States)

Tsai, William; Savaş, Omer

2010-02-01

A pulsatile flow pumping system is developed to replicate flow waveforms with reasonable accuracy for experiments simulating physiological blood flows at numerous points in the body. The system divides the task of flow waveform generation between two pumps: a gear pump generates the mean component and a piston pump generates the oscillatory component. The system is driven by two programmable servo controllers. The frequency response of the system is used to characterize its operation. The system has been successfully tested in vascular flow experiments where sinusoidal, carotid, and coronary flow waveforms are replicated.

Multi-objective optimization of the reactor coolant system

International Nuclear Information System (INIS)

Chen Lei; Yan Changqi; Wang Jianjun

2014-01-01

Background: Weight and size are important criteria in evaluating the performance of a nuclear power plant. It is of great theoretical value and engineering significance to reduce the weight and volume of the components for a nuclear power plant by the optimization methodology. Purpose: In order to provide a new method for the optimization of nuclear power plant multi-objective, the concept of the non-dominated solution was introduced. Methods: Based on the parameters of Qinshan I nuclear power plant, the mathematical models of the reactor core, the reactor vessel, the main pipe, the pressurizer and the steam generator were built and verified. The sensitivity analyses were carried out to study the influences of the design variables on the objectives. A modified non-dominated sorting genetic algorithm was proposed and employed to optimize the weight and the volume of the reactor coolant system. Results: The results show that the component mathematical models are reliable, the modified non-dominated sorting generic algorithm is effective, and the reactor inlet temperature is the most important variable which influences the distribution of the non-dominated solutions. Conclusion: The optimization results could provide a reference to the design of such reactor coolant system. (authors)

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.

Advances in heat pump systems: A review

International Nuclear Information System (INIS)

Chua, K.J.; Chou, S.K.; Yang, W.M.

2010-01-01

Heat pump systems offer economical alternatives of recovering heat from different sources for use in various industrial, commercial and residential applications. As the cost of energy continues to rise, it becomes imperative to save energy and improve overall energy efficiency. In this light, the heat pump becomes a key component in an energy recovery system with great potential for energy saving. Improving heat pump performance, reliability, and its environmental impact has been an ongoing concern. Recent progresses in heat pump systems have centred upon advanced cycle designs for both heat- and work-actuated systems, improved cycle components (including choice of working fluid), and exploiting utilisation in a wider range of applications. For the heat pump to be an economical proposition, continuous efforts need to be devoted to improving its performance and reliability while discovering novel applications. Some recent research efforts have markedly improved the energy efficiency of heat pump. For example, the incorporation of a heat-driven ejector to the heat pump has improved system efficiency by more than 20%. Additionally, the development of better compressor technology has the potential to reduce energy consumption of heat pump systems by as much as 80%. The evolution of new hybrid systems has also enabled the heat pump to perform efficiently with wider applications. For example, incorporating a desiccant to a heat pump cycle allowed better humidity and temperature controls with achievable COP as high as 6. This review paper provides an update on recent developments in heat pump systems, and is intended to be a 'one-stop' archive of known practical heat pump solutions. The paper, broadly divided into three main sections, begins with a review of the various methods of enhancing the performance of heat pumps. This is followed by a review of the major hybrid heat pump systems suitable for application with various heat sources. Lastly, the paper presents novel

Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

International Nuclear Information System (INIS)

Ware, A.G.

1994-07-01

Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177 degrees C (350 degrees F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program

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

Directory of Open Access Journals (Sweden)

Rasa Kanapienytė

2011-12-01

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

Replacement Saltwell Pumping System Document Bibliography

International Nuclear Information System (INIS)

BELLOMY, J.R.

2000-01-01

This document bibliography is prepared to identify engineering documentation developed during the design of the Replacement Saltwell Pumping System. The bibliography includes all engineering supporting documents and correspondence prepared prior to the deployment of the system in the field. All documents referenced are available electronically through the Records Management Information System (RMIS). Major components of the Replacement Saltwell Pumping System include the Sundyne Canned Motor Pump, the Water Filter Skid, the Injection Water Skid and the Backflow Preventer Assembly. Drawing H-14-104498 provides an index of drawings (fabrication details, PandIDs, etc.) prepared to support development of the Replacement Saltwell Pumping System. Specific information pertaining to new equipment can be found in Certified Vendor Information (CVI) File 50124. This CVI file has been established specifically for new equipment associated with the Replacement Saltwell Pumping System

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

International Nuclear Information System (INIS)

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

2018-01-01

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

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

International Nuclear Information System (INIS)

Cao Hongjun; Yan Xiuping

2013-01-01

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

Root cause analysis of pump valve failures of three membrane pump systems

NARCIS (Netherlands)

Buijs, L.J.; Eijk, A.; Hooft, L. van

2014-01-01

This paper will present the root cause analysis and the solution of fatigue failures of the pump valves of three membrane pump systems installed on a chemical plant of Momentive in Pernis, the Netherlands. The membrane pumps were installed approximately 30 years ago. Each system has encountered

System and method for determining coolant level and flow velocity in a nuclear reactor

Science.gov (United States)

Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

2013-09-10

A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

TFCX pumped limiter and vacuum pumping system design and analysis

International Nuclear Information System (INIS)

Haines, J.R.

1985-04-01

Impurity control system design and performance studies were performed in support of the Tokamak Fusion Core Experiment (TFCX) pre-conceptual design. Efforts concentrated on pumped limiter and vacuum pumping system design configuration, thermal/mechanical and erosion lifetime performance of the limiter protective surface, and helium ash removal performance. The reference limiter design forms a continuous toroidal belt at the bottom of the device and features a flat surface with a single leading edge. The vacuum pumping system features large vacuum ducts (diameter approximately 1 m) and high-speed, compound cryopumps. Analysis results indicate that the limiter/vacuum pumping system design provides adequate helium ash removal. Erosion, primarily by disruption-induced vaporization and/or melting, limits the protective surface lifetime to about one calendar year or only about 60 full-power hours of operation. In addition to evaluating impurity control system performance for nominal TFCX conditions, these studies attempt to focus on the key plasma physics and engineering design issues that should be addressed in future research and development programs

Polonium release from an ATW burner system with liquid lead-bismuth coolant

International Nuclear Information System (INIS)

Li, N.; Yefimov, E.; Pankratov, D.

1998-04-01

The authors analyzed polonium release hazards in a conceptual pool-type ATW burner with liquid lead-bismuth eutectic (LBE) coolant. Simplified quantitative models are used based on experiments and real NPP experience. They found little Po contamination outside the burner under normal operating conditions with nominal leakage from the gas system. In sudden gas leak and/or coolant spill accidents, the P contamination level can reach above the regulation limit but short exposure would not lead to severe health consequences. They are evaluating and developing mitigation methods

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 lead-bismuth coolant technology for nuclear device

International Nuclear Information System (INIS)

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

2004-01-01

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

OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the Third Workshop (V1000-CT3)

International Nuclear Information System (INIS)

2005-01-01

The overall objective of the VVER-1000 coolant transient (V1000CT) benchmark is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in analysis of reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 concerns calculation of coolant mixing tests and main steam line break (MSLB) scenarios. Each of the two phases contains three exercises. The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. This event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modelled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 - Point kinetics plant simulation; b) Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 - Best-estimate coupled 3-D core/plant system transient modelling. In addition to the measured (experiment) scenario, extreme calculation scenarios were defined in the frame of Exercise 3 for better testing 3-D neutronics/thermal-hydraulics techniques. The proposals concerned: rod ejection simulations with scram set points at two different power levels. The technical topics presented at this workshop were: Review of the benchmark activities after the 2. Workshop; - Discussion of participant's feedback and introduced modifications

Heat transfer properties of organic coolants containing high boiling residues

International Nuclear Information System (INIS)

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

1964-01-01

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

Four-quadrant characteristics of Psb-VVER pumps

International Nuclear Information System (INIS)

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

2005-01-01

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

Pump efficiency in solar-energy systems

Science.gov (United States)

1978-01-01

Study investigates characteristics of typical off-the-shelf pumping systems that might be used in solar systems. Report includes discussion of difficulties in predicting pump efficiency from manufacturers' data. Sample calculations are given. Peak efficiencies, flow-rate control, and noise levels are investigated. Review or theory of pumps types and operating characteristics is presented.

Primary pump vibration under accident conditions

International Nuclear Information System (INIS)

Guthrie, B.M.; Currie, T.C.

1984-06-01

This report presents the results of an international survey on the subject of vibration in nuclear primary coolant pumps due to two-phase flow, accident conditions. The literature search also revealed few Canadian references other than those of Ontario Hydro. Ontario Hydro's work has been extensive. Confidence in the mechanical integrity of the pumpsets is good, given the extent of the testing. However, conclusions with respect to piping integrity and thermal-hydraulic performance are difficult to determine due to the inexact geometry of the piping and the difficulties in estimating fluid conditions at the pump. The tests help to understand the phenomena and provide background information for analysis, but should be applied with caution to plant analyses. Much of the discussion in the report relates to pump head instability. This is perceived to be the most important flow regime causing vibration, as attested by the emphasis of the reviewed literature. A method for quantitative assessment of the forcing functions acting on the pump-piping system due to void generation and collapse is recommended. A relatively fundamental analytical approach is proposed, supplemented by reduced scale testing in the latter stages. 151 refs

Reactor coolant and associated systems in nuclear power plants

International Nuclear Information System (INIS)

1986-01-01

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

An analytical method for defining the pump`s power optimum of a water-to-water heat pump heating system using COP

Directory of Open Access Journals (Sweden)

Nyers Jozsef

2017-01-01

Full Text Available This paper analyzes the energy efficiency of the heat pump and the complete heat pump heating system. Essentially, the maximum of the coefficient of performance of the heat pump and the heat pump heating system are investigated and determined by applying a new analytical optimization procedure. The analyzed physical system consists of the water-to-water heat pump, circulation and well pump. In the analytical optimization procedure the "first derivative equal to zero" mathematical method is applied. The objective function is the coefficient of performance of the heat pump, and the heat pump heating system. By using the analytical optimization procedure and the objective function, as the result, the local and the total energy optimum conditions with respect to the mass flow rate of hot and cold water i. e. the power of circulation or well pump are defined.

First Study of Helium Gas Purification System as Primary Coolant of Co-Generation Reactor

International Nuclear Information System (INIS)

Piping Supriatna

2009-01-01

The technological progress of NPP Generation-I on 1950’s, Generation-II, Generation-III recently on going, and Generation-IV which will be implemented on next year 2025, concept of nuclear power technology implementation not only for generate electrical energy, but also for other application which called cogeneration reactor. Commonly the type of this reactor is High Temperature Reactor (HTR), which have other capabilities like Hydrogen production, desalination, Enhanced Oil Recovery (EOR), etc. The cogeneration reactor (HTR) produce thermal output higher than commonly Nuclear Power Plant, and need special Heat Exchanger with helium gas as coolant. In order to preserve heat transfer with high efficiency, constant purity of the gas must be maintained as well as possible, especially contamination from its impurities. In this report has been done study for design concept of HTR primary coolant gas purification system, including methodology by sampling He gas from Primary Coolant and purification by using Physical Helium Splitting Membrane. The examination has been designed in physical simulator by using heater as reactor core. The result of study show that the of Primary Coolant Gas Purification System is enable to be implemented on cogeneration reactor. (author)

BWR recirculation pump diagnostic expert system

International Nuclear Information System (INIS)

Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

2004-01-01

At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

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.)

A simplified heat pump model for use in solar plus heat pump system simulation studies

DEFF Research Database (Denmark)

Perers, Bengt; Andersen, Elsa; Nordman, Roger

2012-01-01

Solar plus heat pump systems are often very complex in design, with sometimes special heat pump arrangements and control. Therefore detailed heat pump models can give very slow system simulations and still not so accurate results compared to real heat pump performance in a system. The idea here...

Walking beam pumping unit system efficiency measurements

International Nuclear Information System (INIS)

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

1991-01-01

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

Calorimetric and reactor coolant system flow uncertainty

International Nuclear Information System (INIS)

Bates, L.; McLean, T.

1991-01-01

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

Recirculation system for nuclear reactors

International Nuclear Information System (INIS)

Braun, H. E.; Dollard, W. J.; Tower, S. N.

1980-01-01

A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core. During emergency coolant injection, the emergency coolant is prevented from bypassing the core through the recirculation system by check valves inserted into the recirculation system piping

Materials and fabrication requirements for APWR systems

International Nuclear Information System (INIS)

Boothby, R.M.; Hippsley, C.A.; Gorton, O.K.; Garwood, S.J.

1995-01-01

Materials specifications for advanced pressurized water-cooled reactor (APWR) systems are generally based on existing designs, with improved materials and fabrication procedures being developed to counter known degradation effects. In this paper, materials ageing and degradation mechanisms in PWR primary circuit pressure boundary components (i.e. the reactor pressure vessel (RPV), control rod drive mechanisms (CRDMs), coolant piping, coolant pump casing, pressurizer, and steam generators) are reviewed. Important degradation mechanisms include irradiation embrittlement of the RPV, thermal ageing embrittlement of ferritic (e.g. the pressurizer) and cast austenitic (e.g. coolant pump casing and pipe elbows) steel components and environmentally assisted cracking of steam generator tubing and CRDM penetrations. Improved materials specifications and component design and fabrication issues affecting the integrity of the pressure boundary are discussed in the light of these materials problems. Improved fabrication procedures adopted for Sizewell B, such as the utilization of ring forgings to eliminate axial welds in the RPV and steam generator shells and the use of one-piece castings for coolant pump casings, provide a benchmark against which other APWR designs may be judged. (author)

Standardized sampling system for reactor coolants

International Nuclear Information System (INIS)

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

1982-09-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Investigation of break location effects on thermal-hydraulics during intermediate break loss-of-coolant accident experiments at ROSA-III

International Nuclear Information System (INIS)

Koizumi, Yasuo; Tasaka, Kanji

1986-01-01

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25 % main recirculation pump suction line break (MRPS-B) experiments, the 21 % single-ended jet pump drive line break (JPD-B) experiment and the 15 % main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests. In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop. (author)

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

Science.gov (United States)

Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

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

International Nuclear Information System (INIS)

Ji Duan; Cao Xuewu

2006-01-01

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

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

International Nuclear Information System (INIS)

Yuan Chaolian; Shen Yuxian; Wang Chuan; Du Pengcheng

2014-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Solar PV energy for water pumping system

International Nuclear Information System (INIS)

Mahar, F.

1997-01-01

The paper provides an introduction into understanding the relative merits, characteristics, including economics, of photovoltically powered water pumping systems. Although more than 10,000 photovoltaic pumping systems are known to be operating through out the world, many potential users do not know how to decide weather feasibility assessment, and system procurement so that the reader can made an informed decision about water pumping systems, especially those powered with photovoltaics. (author)

Design optimization of photovoltaic powered water pumping systems

International Nuclear Information System (INIS)

Ghoneim, A.A.

2006-01-01

The use of photovoltaics as the power source for pumping water is one of the most promising areas in photovoltaic applications. With the increased use of water pumping systems, more attention has been paid to their design and optimum utilization in order to achieve the most reliable and economical operation. This paper presents the results of performance optimization of a photovoltaic powered water pumping system in the Kuwait climate. The direct coupled photovoltaic water pumping system studied consists of the PV array, DC motor, centrifugal pump, a storage tank that serves a similar purpose to battery storage and a maximum power point tracker to improve the efficiency of the system. The pumped water is desired to satisfy the domestic needs of 300 persons in a remote area in Kuwait. Assuming a figure of 40 l/person/day for water consumption, a volume of 12 m 3 should be pumped daily from a deep well throughout the year. A computer simulation program is developed to determine the performance of the proposed system in the Kuwait climate. The simulation program consists of a component model for the PV array with maximum power point tracker and component models for both the DC motor and the centrifugal pump. The five parameter model is adapted to simulate the performance of amorphous silicon solar cell modules. The size of the PV array, PV array orientation and the pump-motor-hydraulic system characteristics are varied to achieve the optimum performance for the proposed system. The life cycle cost method is implemented to evaluate the economic feasibility of the optimized photovoltaic powered water pumping system. At the current prices of PV modules, the cost of the proposed photovoltaic powered water pumping system is found to be less expensive than the cost of the conventional fuel system. In addition, the expected reduction in the prices of photovoltaic modules in the near future will make photovoltaic powered water pumping systems more feasible

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

Neutronic Analysis on Coolant Options in a Hybrid Reactor System for High Level Waste Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Seoul (Korea, Republic of)

2014-10-15

A fusion-fission hybrid reactor (FFHR) which is a combination of plasma fusion tokamak as a fast neutron source and a fission reactor as of fusion blanket is another potential candidate. In FFHR, fusion plasma machine can supply high neutron-rich and energetic 14.1MeV (D, T) neutrons compared to other options. Therefore it has better capability in HLW incineration. While, it has lower requirements compared to pure fusion. Much smaller-sized tokamak can be achievable in a near term because it needs relatively low plasma condition. FFHR has also higher safety potential than fast reactors just as ADSR because it is subcritical reactor system. FFHR proposed up to this time has many design concepts depending on the design purpose. FFHR may also satisfy many design requirement such as energy multiplication, tritium production, radiation shielding for magnets, fissile breeding for self-sustain ability also waste transmutation. Many types of fuel compositions and coolant options have been studied. Effect of choices for fuel and coolant was studied for the transmutation purpose FFHR by our team. In this study LiPb coolant was better than pure Li coolant both for neutron multiplication and tritium breeding. However, performance of waste transmutation was reduced with increased neutron absorption at coolant caused by tritium breeding. Also, LiPb as metal coolant has a problem of massive MHD pressure drop in coolant channels. Therefore, in a previous study, waste transmutation performance was evaluated with light water coolant option which may be a realistic choice. In this study, a neutronic analysis was done for the various coolant options with a detailed computation. One of solutions suggested is to use the pressure tubes inside of first wall and second wall In this work, performance of radioactive waste transmutation was compared with various coolant options. On the whole, keff increases with all coolants except for FLiBe, therefore required fusion power is decreased. In

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Pumps for medium sized solar systems

DEFF Research Database (Denmark)

Furbo, Simon

1996-01-01

The suitability of the electronically controlled circulation pump type UPE 2000 from Grundfos for large solar heating systems was elucidated.......The suitability of the electronically controlled circulation pump type UPE 2000 from Grundfos for large solar heating systems was elucidated....

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

Science.gov (United States)

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

2010-06-01

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

Primary coolant pipe rupture event in liquid metal cooled reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2004-08-01

In liquid-metal cooled fast reactors (LMFR) the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). However, the primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors (Indira Gandhi Centre for Atomic Research, Kalpakkam, India, 13-17 January 2003) was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the technical meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the technical meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

Modeling the spatial distribution of the parameters of the coolant in the reactor volume

International Nuclear Information System (INIS)

Nikonov, S.P.

2011-01-01

In this paper the approach to the question about the spatial distribution of the parameters of the coolant in-reactor volume. To describe the in-core space is used specially developed preprocessor. When the work of the preprocessor in the first place, is recreated on the basis of available information (mostly-the original drawings) with high accuracy three-dimensional description of the structures of the reactor volume and, secondly, are prepared on this basis blocks input to the nodal system code improved estimate ATHLET, allows to take into account the hydrodynamic interaction between the spatial control volumes. As an example the special case of solutions of international standard problem on the reconstruction of the transition process in the third unit of the Kalinin nuclear power plant, due to the shutdown of one of the four Main Coolant Pumps in operation at the rated capacity (first download). Model-core area consists of approximately 58 000 control volumes and spatial relationships. It shows the influence of certain structural units of the core to the distribution of the mass floe rate of its height. It is detected a strong cross-flow coolant in the area over the baffle. Moreover, we study the distribution of the coolant temperature at the assembly head of WWER-1000 reactor. It is shown that in the region of the top of the assembly head, where we have installation of thermocouples, the flow coolant for internal assemblies core is formed by only from guide channel Reactor control and protected system Control rod flow, or a mixture of the guide channel flow and flow from the area in front of top grid head assembly (the peripheral assemblies). It is shown that the magnitude of the flow guide channels affects not only the position of control rods, but also the presence of a particular type of measuring channels (Self powered neutron detector sensors or Temperature control sensors) in the cassette. (Author)

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

International Nuclear Information System (INIS)

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

1966-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1987-04-01

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

On-Line Coolant Chemistry Analysis

International Nuclear Information System (INIS)

LM Bachman

2006-01-01

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

Water quality control device and water quality control method for reactor primary coolant system

International Nuclear Information System (INIS)

Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

1995-01-01

The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

Heat-pump-centered integrated community energy systems

Energy Technology Data Exchange (ETDEWEB)

Schaetzle, W.J.; Brett, C.E.; Seppanen, M.S.

1979-12-01

The heat-pump-centered integrated community energy system (HP-ICES) supplies district heating and cooling using heat pumps and a thermal energy storage system which is provided by nature in underground porous formations filled with water, i.e., aquifers. The energy is transported by a two-pipe system, one for warm water and one for cool water, between the aquifers and the controlled environments. Each energy module contains the controlled environments, an aquifer, wells for access to the aquifer, the two pipe water distribution system and water source heat pumps. The heat pumps upgrade the energy in the distribution system for use in the controlled environments. Economically, the system shows improvement on both energy usage and capital costs. The system saves over 60% of the energy required for resistance heating; saves over 30% of the energy required for most air-source heat pumps and saves over 60% of the energy required for gas, coal, or oil heating, when comparing to energy input required at the power plant for heat pump usage. The proposed system has been analyzed as demonstration projects for a downtown portion of Louisville, Kentucky, and a section of Fort Rucker, Alabama. The downtown Louisville demonstration project is tied directly to major buildings while the Fort Rucker demonstration project is tied to a dispersed subdivision of homes. The Louisville project shows a payback of approximately 3 y, while Fort Rucker is approximately 30 y. The primary difference is that at Fort Rucker new heat pumps are charged to the system. In Louisville, either new construction requiring heating and cooling systems or existing chillers are utilized. (LCL)

Thermosyphon Phenomenon as an alternate heat sink of Shutdown Cooling System for the CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Jonghyun [GNEST, Seoul (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

2006-07-01

During the outage(overhaul) of the CANDU plant, there is a period when the coolant is partially drained to the reactor header level and the coolant is cooled and depressurized by Shutdown Cooling System(SDCS) other than PHTS pump. In the postulated accident of the loss of SDCS-the PHTS pump failure, the primary coolant system should be cooled by the alternate heat sink using the thermosyphon pheonomenon(TS) through the steam generator(SG) This study was aimed at verification and analyzing the core cooling ability of the TS. And the sensitivity analysis was done for the number of SGs used in the TS. As an analysis tool, RELAP5/CANDU was used.

Improvement to liquid metal pumps

International Nuclear Information System (INIS)

Pennell, W.E.

1981-01-01

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

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

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)

Phenomena occurring in the reactor coolant system during severe core damage accidents

International Nuclear Information System (INIS)

Malinauskas, A.P.

1989-01-01

The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. In the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. In contrast to normal operating conditions, severe core damage accidents are characterized by significant temporal and spatial variations in heat and mass fluxes, and by eventual geometrical changes within the RCS. Furthermore, the difficulties in describing the system in the severe accident mode are compounded by the occurrence of chemical reactions. These reactions can influence both the thermal and the mass transport behavior of the system. In addition, behavior of the reactor vessel internals and of materials released from the core region (especially the radioactive fission products) in the course of the accident likewise become of concern to the analyst. This report addresses these concerns. 9 refs., 1 tab

Validation of the mathematical model of the NAPS PHT system flow with test data

International Nuclear Information System (INIS)

Rajesh Kumar, K.; Vani, K.; Chakraborty, G.; Venkat Raj, V.

1994-01-01

A dynamic analysis code to predict the time dependent behaviour of the reactor coolant system flow following the tripping and starting of Primary Circulating Pumps in the different operating modes has been developed for Indian Pressurised Heavy Water Reactor (PHWR) type power plants. The model is comprised of reactor coolant momentum equation, Primary Heat Transport (PHT) pump dynamic equation and pump characteristics. This model forms one of the modules of the integrated system code being developed for transient analysis of 220 MWe PHWR power plants. The Narora Atomic Power Station (NAPS) PHT system flow transient results for different combinations of pump operation predicted by the model have been compared with the experimental data obtained from a test carried out in NAPS-2 for validation of the model. The predicted results are in good agreement with the experimental data. (author). 3 refs., 5 figs

Design of dry scroll vacuum pumping system for efficient pumping of corrosive gas at medium vacuum range

International Nuclear Information System (INIS)

Banerjee, I.; Chandresh, B.G.; Guha, K.C.; Sarkar, S.

2015-01-01

Dry vacuum pumping systems attracts many applications because of its inherent capability of corrosion free pumping. It becomes a common trait of application in Thermo Nuclear Fusion, Semi conductor, Isotope separation industries etc. Thermo nuclear fusion requires a train of specially sealed roots pump backed by suitable capacity dry screw or reciprocating pump. Similarly corrosive fluoride gas pumping requires hermetically sealed specially designed dry scroll vacuum pump. Plant emergency operation however involves train of specially sealed roots pump backed with scroll pump for faster evacuation. In our attempt an indigenously designed scroll pump and associated system are designed to pump corrosive gases in a way to confine the corrosion product within the system. In order to execute the design, a numerical code for low pressure application is developed

N-Springs pump and treat system optimization study

International Nuclear Information System (INIS)

1997-03-01

This letter report describes and presents the results of a system optimization study conducted to evaluate the N-Springs pump and treat system. The N-Springs pump and treat is designed to remove strontium-90 (90Sr) found in the groundwater in the 100-NR-2 Operable Unit near the Columbia River. The goal of the system optimization study was to assess and quantify what conditions and operating parameters could be employed to enhance the operating and cost effectiveness of the recently upgraded pump and treat system.This report provides the results of the system optimization study, reports the cost effectiveness of operating the pump and treat at various operating modes and 90Sr removal goals, and provides recommendations for operating the pump and treat

ITER - torus vacuum pumping system remote handling issues

International Nuclear Information System (INIS)

Stringer, J.

1992-11-01

This report describes further design issues concerning remote maintenance of torus vacuum pumping systems options for ITER. The key issues under investigation in this report are flask support systems for valve seal exchange operations for the compound cryopump scheme and remote maintenance of a proposed multiple turbomolecular pump (TMP) system, an alternative ITER torus exhaust pumping option. Previous studies have shown that the overhead support methods for seal exchange flask equipment could malfunction due to valve/flask misalignment. A floor-mounted support system is described in this report. This scheme provides a more rigid support system for seal exchange operations. An alternative torus pumping system, based on the use of multiple TMPs, is studied from a remote maintenance standpoint. In this concept, centre distance spacing for pump/valve assemblies is too restrictive for remote maintenance. Recommendations are made for adequate spacing of these assemblies based on commercially-available 0.8 m and 1.0 m diameter valves. Fewer pumps will fit in this arrangement, which implies a need for larger TMPs. Pumps of this size are not commercially available. Other concerns regarding the servicing and storage of remote handling equipment in cells are also identified. (9 figs.)

Structural analysis of the as-built IEA-R1 primary coolant piping system using a complete three dimensional model

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Martins, Lucas B.; Marcolin, Gabriel; Mattar Neto, Miguel

2011-01-01

IEA-R1 is an open pool type research reactor, moderated by light water and upgraded from 2 MW to 5 MW of operating power level. Heat generated in the reactor core is removed by a coolant system divided in two circuits, primary and secondary, composed by pumps, piping, heat exchangers, cooling tower, and some other auxiliary components. The 5 MW operating power level is now possible due to a modernization program started in 1996. As a part of the modernization program, ageing assessment studies recommend the replacement of one of the two heat exchangers in the circuit. To manage this replacement, modifications in the layout of the primary and secondary piping and supporting systems were performed, based on preliminary stress analysis study. Then, the aim of this work is to present the final stress analysis of the primary circuit. To reach this and taking the modifications of the primary into account, a 3D model of the whole circuit, in the as-built condition, was made. Stress results and discussions are shown. (author)

Investigation of pump and pump switch failures in rainwater harvesting systems

Science.gov (United States)

Moglia, Magnus; Gan, Kein; Delbridge, Nathan; Sharma, Ashok K.; Tjandraatmadja, Grace

2016-07-01

Rainwater harvesting is an important technology in cities that can contribute to a number of functions, such as sustainable water management in the face of demand growth and drought as well as the detention of rainwater to increase flood protection and reduce damage to waterways. The objective of this article is to investigate the integrity of residential rainwater harvesting systems, drawing on the results of the field inspection of 417 rainwater systems across Melbourne that was combined with a survey of householders' situation, maintenance behaviour and attitudes. Specifically, the study moves beyond the assumption that rainwater systems are always operational and functional and draws on the collected data to explore the various reasons and rates of failure associated with pumps and pump switches, leaving for later further exploration of the failure in other components such as the collection area, gutters, tank, and overflows. To the best of the authors' knowledge, there is no data like this in academic literature or in the water sector. Straightforward Bayesian Network models were constructed in order to analyse the factors contributing to various types of failures, including system age, type of use, the reason for installation, installer, and maintenance behaviour. Results show that a number of issues commonly exist, such as failure of pumps (5% of systems), automatic pump switches that mediate between the tank and reticulated water (9% of systems), and systems with inadequate setups (i.e. no pump) limiting their use. In conclusion, there appears to be a lack of enforcement or quality controls in both installation practices by sometimes unskilled contractors and lack of ongoing maintenance checks. Mechanisms for quality control and asset management are required, but difficult to promote or enforce. Further work is needed into how privately owned assets that have public benefits could be better managed.

Lead coolant test facility systems design, thermal hydraulic analysis and cost estimate

Energy Technology Data Exchange (ETDEWEB)

Khericha, Soli, E-mail: slk2@inel.gov [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Harvego, Edwin; Svoboda, John; Evans, Robert [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Dalling, Ryan [ExxonMobil Gas and Power Marketing, Houston, TX 77069 (United States)

2012-01-15

The Idaho National Laboratory prepared a preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed below: Bullet Develop and demonstrate feasibility of submerged heat exchanger. Bullet Develop and demonstrate open-lattice flow in electrically heated core. Bullet Develop and demonstrate chemistry control. Bullet Demonstrate safe operation. Bullet Provision for future testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimated. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 Degree-Sign C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included

Optically pumped laser systems

International Nuclear Information System (INIS)

DeMaria, A.J.; Mack, M.E.

1975-01-01

Laser systems which are pumped by an electric discharge formed in a gas are disclosed. The discharge is in the form of a vortex stabilized electric arc which is triggered with an auxiliary energy source. At high enough repetition rates residual ionization between successive pulses contributes to the pulse stabilization. The arc and the gain medium are positioned inside an optical pumping cavity where light from the arc is coupled directly into the gain medium

Coolant leakage detection device

International Nuclear Information System (INIS)

Ito, Takao.

1983-01-01

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

Heat pumping in nanomechanical systems

OpenAIRE

Chamon, Claudio; Mucciolo, Eduardo R.; Arrachea, Liliana; Capaz, Rodrigo B.

2010-01-01

We propose using a phonon pumping mechanism to transfer heat from a cold to a hot body using a propagating modulation of the medium connecting the two bodies. This phonon pump can cool nanomechanical systems without the need for active feedback. We compute the lowest temperature that this refrigerator can achieve.

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

International Nuclear Information System (INIS)

Austregesilo Filho, H.

1978-12-01

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

Numerical Simulation of Tubular Pumping Systems with Different Regulation Methods

Science.gov (United States)

Zhu, Honggeng; Zhang, Rentian; Deng, Dongsheng; Feng, Xusong; Yao, Linbi

2010-06-01

Since the flow in tubular pumping systems is basically along axial direction and passes symmetrically through the impeller, most satisfying the basic hypotheses in the design of impeller and having higher pumping system efficiency in comparison with vertical pumping system, they are being widely applied to low-head pumping engineering. In a pumping station, the fluctuation of water levels in the sump and discharge pool is most common and at most time the pumping system runs under off-design conditions. Hence, the operation of pump has to be flexibly regulated to meet the needs of flow rates, and the selection of regulation method is as important as that of pump to reduce operation cost and achieve economic operation. In this paper, the three dimensional time-averaged Navier-Stokes equations are closed by RNG κ-ɛ turbulent model, and two tubular pumping systems with different regulation methods, equipped with the same pump model but with different designed system structures, are numerically simulated respectively to predict the pumping system performances and analyze the influence of regulation device and help designers make final decision in the selection of design schemes. The computed results indicate that the pumping system with blade-adjusting device needs longer suction box, and the increased hydraulic loss will lower the pumping system efficiency in the order of 1.5%. The pumping system with permanent magnet motor, by means of variable speed regulation, obtains higher system efficiency partly for shorter suction box and partly for different structure design. Nowadays, the varied speed regulation is realized by varied frequency device, the energy consumption of which is about 3˜4% of output power of the motor. Hence, when the efficiency of variable frequency device is considered, the total pumping system efficiency will probably be lower.

Aging related degradation in turbine drives and governors for safety related pumps

International Nuclear Information System (INIS)

Cox, D.F.

1991-01-01

This study is being performed to examine the relationship between time dependent degradation, and current industry practices in the areas of maintenance, surveillance, and operation of stem turbine drive for safety related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized water reactor (PWR) plants, and the Reactor Core Isolation Cooking (RCIC) and High Pressure Coolant Injection (HPCI) systems for Boiling Water Reactor (BWR) facilities. This research has been conducted by examining current information in the Nuclear Plant Reliability Data System (NPRDS), reviewing Licensee Event Reports, thoroughly investigating contacts with operating plant personnel, and by personal observation. This information was reviewed to determine the cause of each reported event and the method of discovery. From this data attempts have been made at determining the predictability of events and possible preventive measures that may be implemented

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

High temperature heat pumps for industrial cooling; Hoejtemperatur varmepumper til industriel koeling

Energy Technology Data Exchange (ETDEWEB)

Rasmussen, Lars; Nielsen, Jacob [Advansor A/S, Aarhus (Denmark); Kronborg, H. [Cronborg, Holstebro (Denmark); Skouenborg, K. [Jensens Koekken, Struer (Denmark)

2013-03-15

This report deals with theoretical analysis of various types of integration of heat pumps in the industry, as well as a demonstration plant that serves the project's practical execution. The report describes the system integration between heat pumps and existing industrial cooling systems. Ammonia plants in industry are estimated to have an allocation of 85%, which is why only an analysis of this type of installation as surplus heat supplier is included in this report. In contrast, heat pumps with both CO{sub 2} and Isobutane as the refrigerant are analysed, since these are the interesting coolants for generating high temperature heat. It can be seen through the project that the combination of heat pump with existing cooling installations may produce favorable situations where the efficiency of the heat pump is extremely high while at the same time electricity and water consumption for the cooling system is reduced. The analysis reflects that CO{sub 2} is preferred over Isobutane in the cases where a high level of temperature boost is desired, whereas Isobutane is preferable at low level of temperature boost. In the demonstration project, the report shows that the heat pump alone has a COP of 4.1, while the achieved COP is 5.5 when by considering the system as a whole. In addition to increased performance the solution profits by having a reduction in CO{sub 2} emissions of 81 tons/year and a saving of 470,000 DKK/year. (LN)

Absorption heat pump system

Science.gov (United States)

Grossman, G.

1982-06-16

The efficiency of an absorption heat pump system is improved by conducting liquid from a second stage evaporator thereof to an auxiliary heat exchanger positioned downstream of a primary heat exchanger in the desorber of the system.

ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

Directory of Open Access Journals (Sweden)

Veijo KAUPPINEN

2012-11-01

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

Operation of pumps in two-phase steam-water flow

International Nuclear Information System (INIS)

Grison, P.; Lauro, J.F.

1978-01-01

Determining the two-phase flow (critical or not) through a pump is an esential element for a complete description of loss of coolant accident in a PWR reactor. This article descibes the theoretical and experimental research being done on this subject in France. The model of the pump is first described and its behaviour is examined in different possible cases, particularly that of critical flow. The analysis of the behaviour of the pump is then used to define the experimental conditions for the tests. Two test loops, EVA and EPOPEE, were built. The experimental results are then compared with the theoretical forecasts [fr

Experimental Investigation on The Electromagnetic Clutch Water pump and Pneumatic Compressor for Improving the Efficiency of an Engine

Science.gov (United States)

Kumarasubramanian, R.; Xavier, Goldwin; Nishanthi, W. Mary; Rajasekar, R.

2017-05-01

Considering the fuel crises today many work and research were conducted to reduce the fuel consumption of the internal combustion engine. The fuel consumption of an internal combustion engine can be relatively reduced by use of the electromagnetic clutch water pump and pneumatic compressor. Normally in an engine, the water pump is driven by the crankshaft, with an aid of belt, for the circulation of the water for the cooling process. The circulation of coolant is resisted by the thermostat valve, while the temperature inside the coolant jacket of the engine is below 375K the thermostat is closed only above 375K it tends to open. But water pump run continuously even when thermostat is closed. In pneumatic braking system, pneumatic or air compressor purpose is to compress the air and stored into the storage tank for the brake operation. When the air pressure of the storage tanks gets increases above its storage capacity pressure is regulated by governor, by passing them to atmosphere. Such unnecessary work of this water pump and air compressor can be minimized by use of the electromagnetic clutch water pump and air compressor. The European Driving Cycle is used to evaluate the performance of this water pump and air compressor when used in an engine. The result shows that the fuel economy of the engine while using electromagnetic water pump and pneumatic compressor were improved by 8.0% compared with conventional types which already exist. The application of these electromagnetic water pump and pneumatic compressor are expected to contribute for the improvement of engine performance because of their effect in reduction of the rate of fuel consumption.

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

International Nuclear Information System (INIS)

Kadalev, Stoyan

2014-01-01

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

Heat pumping in nanomechanical systems.

Science.gov (United States)

Chamon, Claudio; Mucciolo, Eduardo R; Arrachea, Liliana; Capaz, Rodrigo B

2011-04-01

We propose using a phonon pumping mechanism to transfer heat from a cold to a hot body using a propagating modulation of the medium connecting the two bodies. This phonon pump can cool nanomechanical systems without the need for active feedback. We compute the lowest temperature that this refrigerator can achieve. © 2011 American Physical Society

An analytical model for prediction of two-phase (noncondensable) flow pump performance

International Nuclear Information System (INIS)

Furuya, O.

1985-01-01

During operational transients or a hypothetical LOCA (loss of coolant accident) condition, the recirculating coolant of PWR (pressurized water reactor) may flash into steam due to a loss of line pressure. Under such two-phase flow conditions, it is well known that the recirculation pump becomes unable to generate the same head as that of the single-phase flow case. Similar situations also exist in oil well submersible pumps where a fair amount of gas is contained in oil. Based on the one dimensional control volume method, an analytical method has been developed to determine the performance of pumps operating under two-phase flow conditions. The analytical method has incorporated pump geometry, void fraction, flow slippage and flow regime into the basic formula, but neglected the compressibility and condensation effects. During the course of model development, it has been found that the head degradation is mainly caused by higher acceleration on liquid phase and deceleration on gas phase than in the case of single-phase flows. The numerical results for head degradations and torques obtained with the model favorably compared with the air/water two-phase flow test data of Babcock and Wilcox (1/3 scale) and Creare (1/20 scale) pumps

Recent developments in coolant systems for Indus Accelerator Complex at RRCAT, Indore

International Nuclear Information System (INIS)

Nanda, Dipankar; Tiwari, Bablu; Pandey, R.M.

2015-01-01

Scarcity of fresh water forces mankind to explore other possible water sources that can meet the increasing demand of coolants in industries, R and D sectors and other establishments where water is used as coolant. It also becomes a challenge for water chemist to control water chemistry to keep the equipments/devices intact during its operation using water as coolant. Deionised (DI) and soft water have been used as coolants for Indus Accelerator Complex, RRCAT, Indore. DI water is produced and its quality is maintained either by conventional ion exchange method or a hybrid method of membrane separation and ion exchange technique. This requires handling of corrosive chemicals, manpower, space for plant installation, and a long array of water treatment units. CSL has implemented the idea of rain water harvesting to produce DI water after systematic studies in laboratory. The concerning issues are reduced to almost one-fourth by using rain water to produce DI water. The harvesting system has been in use for last three years. Heat is dissipated into air by evaporation of soft water in cooling tower. Requirement of soft water makeup has been estimated to be about 40,000 ltrs. / day (max.) if the machine is operated at its designed specifications. Non-availability of soft water (which circulates in open loop) may lead to shut down like situation and looking for alternate source becomes quite essential. Laboratory studies (water analysis and treatment) on sewage water (available 1,00,000 ltrs/day) from RRCAT colony as a possible source of producing soft water show promising result. (author)

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

Science.gov (United States)

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

1988-01-01

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

Estimation of free-hydrocarbon recovery from dual-pump systems

International Nuclear Information System (INIS)

Charbeneau, R.J.

1995-01-01

Free-product hydrocarbon which floats on the water table may be recovered using single-pump and dual-pump systems. The factors that affect the long-term free-product recovery using dual-pump systems include the free-product thickness as measured in monitoring wells, the ground-water pumping rate, hydrocarbon density and viscosity, and the soil permeability. This paper presents a simple model for prediction of free-product recovery using dual-pump systems. The model predicts the long-term rather than short-term recovery rates, and lends itself to spreadsheet calculations on microcomputers. A particularly simple form arises for cases where the drawdown is small. An application for estimating recovery from a dual-pump system is presented, and limitations of the model are summarized

Energy Efficient Pump Control for an Offshore Oil Processing System

DEFF Research Database (Denmark)

Yang, Zhenyu; Soleiman, Kian; Løhndorf, Bo

2012-01-01

The energy efficient control of a pump system for an offshore oil processing system is investigated. The seawater is lifted up by a pump system which consists of three identical centrifugal pumps in parallel, and the lifted seawater is used to cool down the crude oil flowing out of a threephase...... separator on one of the Danish north-sea platform. A hierarchical pump-speed control strategy is developed for the considered system by minimizing the pump power consumption subject to keeping a satisfactory system performance. The proposed control strategy consists of online estimation of some system...... operating parameters, optimization of pump configurations, and a real-time feedback control. Comparing with the current control strategy at the considered system, where the pump system is on/off controlled, and the seawater flows are controlled by a number of control valves, the proposed control strategy...

Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor coolant system and systems decontamination. Summary status report. Volume 1

International Nuclear Information System (INIS)

Doerge, D.H.; Miller, R.L.; Scotti, K.S.

1986-05-01

This document summarizes information relating to the decontamination and restoration of the Three Mile Island Unit 2 reactor coolant system and other plant systems. Data have been collected from activity reports, reactor containment entry records, and other sources and entered in a computerized data system which permits extraction/manipulation of specific data which can be used in planning for recovery from a loss of coolant event similar to that experienced by the Three Mile Island Unit 2 on March 28, 1979. This report contains a summary of radiation exposures, manpower, and time spent in radiation areas during the referenced period. Support activities conducted outside of radiation areas are not included. Computer reports included are: A chronological listing of all activities related to decomtamination and restoration of the reactor coolant system and other plant systems for the period of April 5, 1979, through December 19, 1984; a summary of labor and exposures by department for the same time period; and summary reports for each major task undertaken in connection with this specific work scope during the referenced time period

PWR systems transient analysis

International Nuclear Information System (INIS)

Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

1985-01-01

Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

Device for extracting steam or gas from the primary coolant line leading from a reactor pressure vessel to a straight through boiler or from the top primary boiler chamber of a water-cooled nuclear reactor

International Nuclear Information System (INIS)

Schatz, K.

1982-01-01

In such a nuclear reactor, a steam or gas cushion can form when the primary system is refilled, which can cause blocking of the natural circulation or filling of the system in the area of the hot primary coolant pipe or in the top primary boiler chamber. In order to remove such a steam or gas cushion, a ventilation pipe starting from the bend of the primary coolant line is connected to the feed pipe for introducing water into the primary system. The feed pipe is designed on the principle of the vacuum pump in the area of the opening of the ventilation pipe. There is a sub-pressure in the ventilation pipe, which makes it possible to extract the steam or gas. After mixing in the area of the opening, the steam condenses or is distributed with the gas in the primary coolant. (orig.) [de

Turbomolecular pump vacuum system for the Princeton Large Torus

International Nuclear Information System (INIS)

Dylla, H.F.

1977-10-01

A turbomolecular pump vacuum system has been designed and installed on the Princeton Large Torus (PLT). Four vertical shaft, oil-bearing, 1500 l/s turbomolecular pumps have been interfaced to the 6400 liter PLT Vacuum vessel to provide a net pumping speed of 3000 l/s for H 2 . The particular requirements and problems of tokamak vacuum systems are enumerated. A vacuum control system is described which protects the vacuum vessel from contamination, and protects the turbomolecular pumps from damage under a variety of possible failure modes. The performance of the vacuum system is presented in terms of pumping speed measurements and residual gas behavior

Research on Coolant Radiochemistry

International Nuclear Information System (INIS)

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

2007-06-01

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

Coolant circulation system for a liquid metal nuclear reactor

International Nuclear Information System (INIS)

DeLuca, R.A.; Garabedian, G.

1988-01-01

This patent describes a liquid metal circulation system comprising an electromagnetic pump comprised of: (a) an elongated cylindrical pump support housing; (b) a cylindrical pressure dome structure coaxially situated and supported within the pump support housing, having a closed, hemispherical upper end and an open, cylindrical lower end; (c) a cylindrical pump coaxially situated within the pressure dome structure including: (1) a central core body of laminated transformer steel having six peripherally equally spaced helical grooves on its outer surface extending the entire length of the central core body, (2) a multiplicity of square, ceramic insulated copper wires situated in the helical grooves, (3) electrical leads extending from the terminal ends of the square copper wires through the upper end of the pressure dome structure and to a three-phase low voltage/high amperage power source, (4) an austenitic stainless steel jacket covering the outer surface of the central core body and covering the helically coiled square copper wires, the outer stainless steel jacket and the inner surface of the pressure dome structure defining an annular flow passage

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

International Nuclear Information System (INIS)

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

2017-01-01

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

HANARO secondary coolant management

International Nuclear Information System (INIS)