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Sample records for main feedwater system

  1. Operation of the main feedwater system turbopump following plant trip with total failure of the auxiliary feedwater system

    International Nuclear Information System (INIS)

    Lucas Alvaro, A.M. de; Rosa Martinez, B. de la; Alcaide, F.; Toledano Camara, C.

    1993-01-01

    The Auxiliary Feedwater System (AF) is a safeguard system which has been designed to supply feedwater to the steam generators, cool the primary system and remove decay heat from the reactor when the main feedwater pumps fail due to loss of power or any other reason. Thus, when plant trip occurs, the AF system pumps start up automatically, allowing removal of decay heat from the reactor. However, even though this system (2 motor-driven pumps and 1 turbopump) is highly reliable, injection of water to the steam generators must be ensured when it fails completely. To do this, if plant trip has not been caused by loss of off site power or failure of the Main Feedwater System (FW) turbopumps, one of these turbopumps can be used to achieve removal of decay heat. Since a large amount of steam is consumed by these turbopumps, an analysis has been performed to determine whether one of these pumps can be used and what actions are necessary to inject water into the steam generators. Results show that, for the case in question, a FW turbopump can be used to remove decay heat from the reactor. (author)

  2. Factors analysis of water hammer in FLOWMASTER for main feedwater systems of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Wang Xin; Han Weishi

    2010-01-01

    The main feedwater system of a nuclear power plant (NPP) is an important part in ensuring the cooling of a steam generator. It is the main pipe section where water hammers frequently occur. Studying the regulator patterns of water hammers in the main feedwater systems is significant to the stable operation of the system. This article focuses on a parametric study to avoid the consequences of water hammer effect in PWR by employing a general purpose fluid dynamic simulation software-FLOWMASTER. Through FLOWMASTER's transient calculating functions, a mathematical model is established with boundary conditions such as feedwater pumps, control valves, etc., calculations of water hammer pressure when feedwater pumps and control valves shut down, and simulations during instantaneous changes in water hammer pressure. Combining a plethora of engineering practical examples, this research verified the viability of calculating water hammer pressure through FLOWMASTER's transient functions and we found out that, increasing the periods of closure of control valves and feedwater pumps control water hammers effectively. We also found out that changing the intervals of closing signals to feedwater pumps and control valves aid to relieve hydraulic impact. This could be a guideline for practical engineering design and system optimization. (author)

  3. Water hammer calculation and analysis in main feedwater system of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Wang Xin; Han Weishi

    2010-01-01

    The main feedwater system of a nuclear power plant is an important part in ensuring the cooling of the steam generator. Moreover, it is the main pipe section where water hammers frequently occur. Studying the regular patterns of water hammers to the main feedwater system is significant to the stable operation of the system. The paper focuses on the study of water hammers through Flowmaster's transient calculating function to establish a mathematical model with boundary conditions such as a feedwater pump, control valves, etc.; calculation of the water hammers pressure when feedwater pumps and control valves shut down; exporting the instantaneous change in solution of pressure. Combined with engineering practical examples, the conclusions verify the viability of calculating the water hammers pressure through Flowmaster's transient function, increasing the periods of closure of control valves and feedwater pumps control water hammers effectively, changing the intervals of closing signals to feedwater pumps and control valves to relieve hydraulic impact. This could be a guideline for practical engineering design and system optimization. (authors)

  4. Simulation of main steam and feedwater system of full scope simulator for Qinshan 300 MW Nuclear Power Unit

    International Nuclear Information System (INIS)

    Zhao Xiaoyu

    1996-01-01

    The simulation of main steam and feedwater system is the most important and maximal part in secondary circuit model, including all of main steam and feedwater's thermal-hydraulic properties, except heat-exchange of secondary side of steam generator. It simulates main steam header, steam power in each stage of turbine, moisture separator-reheater, deaerator, condenser, high pressure and low pressure heater, auxiliary feedwater and main steam bypass in full scope

  5. Reactor feedwater system

    International Nuclear Information System (INIS)

    Kagaya, Hiroyuki; Tominaga, Kenji.

    1993-01-01

    In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

  6. Feedwater control system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    Excessive swing of the feedwater in nuclear reactor power supply apparatus on the occurrence of a transient is suppressed by injecting an anticipatory compensating signal (δWsub(fw)) into the control for the feedwater. Typical overshoot occurs on removal of a large part of the load, the steam flow is reduced so that the conventional control system reduces the flow of feedwater. At the same time there is a reduction of feedwater level in the steam generator because of the collapse of the bubbles under increased steam pressure. By the time the control responds to the drop in level, the apparatus has begun to stabilize so that there is overshoot. The anticipatory signal is derived from the boiling power (BP) which is a function of the nuclear power (Qsub(N)) developed, the enthalpy of saturated water (hsub(s)) and the enthalpy of the feedwater injected into the steam generator (hsub(fw)). From the boiling power (BP) and the increment in steam pressure resulting from the transient an anticipatory increment of feedwater flow is derived. This increment is added to the other parameters controlling the feedwater. (author)

  7. Auxiliary feedwater system risk-based inspection guide for the Maine Yankee Nuclear Power Plant

    International Nuclear Information System (INIS)

    Gore, B.F.; Vo, T.V.; Moffitt, N.E.; Bumgardner, J.D.

    1992-10-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. The information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Maine Yankee was selected as one of a series of plants for study. ne product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Maine Yankee plant

  8. Monitor for reactor feedwater systems

    International Nuclear Information System (INIS)

    Takizawa, Yoji; Tomizawa, Teruaki

    1983-01-01

    Purpose: To improve the reliability of operator's procedures upon occurrence of the feedwater system abnormality in a BWR type reactor by presenting the operation with effective information to avoid such abnormality. Constitution: A feedwater temperature at the reactor inlet of a reactor feedwater system measured by a temperature detector and a predetermined value for the feedwater temperature at the reactor inlet determined depending on the reactor conditions are inputted to a start-up system. The start-up system outputs a start-up signal when the difference between the inputted values exceeds a predetermined value. Then, the start-up signal is inputted to a display device where information required for the operator is displayed in the device. Thus, the information required for the operator is rapidly provided upon abnormality of the feedwater system to thereby improve the reliability of the operator's procedures. (Moriyama, K.)

  9. Trend and pattern analysis of failures of main feedwater system components in United States commercial nuclear power plants

    International Nuclear Information System (INIS)

    Gentillon, C.D.; Meachum, T.R.; Brady, B.M.

    1987-01-01

    The goal of the trend and pattern analysis of MFW (main feedwater) component failure data is to identify component attributes that are associated with relatively high incidences of failure. Manufacturer, valve type, and pump rotational speed are examples of component attributes under study; in addition, the pattern of failures among NPP units is studied. A series of statistical methods is applied to identify trends and patterns in failures and trends in occurrences in time with regard to these component attributes or variables. This process is followed by an engineering evaluation of the statistical results. In the remainder of this paper, the characteristics of the NPRDS that facilitate its use in reliability and risk studies are highlighted, the analysis methods are briefly described, and the lessons learned thus far for improving MFW system availability and reliability are summarized (orig./GL)

  10. Feedwater system in a nuclear power plant

    International Nuclear Information System (INIS)

    Shimizu, Tadayuki.

    1975-01-01

    Object: To improve the control property of a steam turbine for a feedwater pump and plant operation characteristics where water is supplied at a low rate. Structure: In a nuclear power plant where feedwater pumps of the reactor are driven by a steam turbine, the main feedwater duct on the discharge side of the feedwater pumps is provided with a cut-off valve and is connected parallel with a bypass duct having a pressure compensated flow control valve. With this arrangement, at the time when the rate of feedwater is high the cut-off valve is open so that water supplied from the feedwater pumps driven by the steam turbine is supplied through the main feedwater duct to the reactor while in case when the rate of feedwater is low the flow control valve is opened to let the water be supplied through the bypass duct. (Kamimura, M.)

  11. Feedwater control system in nuclear power plants

    International Nuclear Information System (INIS)

    Masuyama, Hideo.

    1981-01-01

    Purpose: To enable switching operation for feedwater systems in a short time and with no fluctuations in the reactor water level by increasing or decreasing the flow rate in the feedwater systems during automatic operation by the amount of the fluctuations in the flow rate in the feedwater system during manual operation. Constitution: In a BWR type nuclear power plant having a plurality of feedwater systems to a nuclear reactor, a feedwater control system is constituted with a reactor water level controller, a M/A switcher for switching either of automatic flow rate demand signals or manual flow rate set signals from the reactor level controller to apply flow rate demand signals for each of the feedwater systems, a calculation device for calculating the flow rate set signals in the feedwater systems during manual operation and an adder for subtracting the flow rate set signals in the manual feedwater system calculated in the calculating device from the automatic flow rate demand signals for the feedwater systems during automatic operation. This enables rapid switching for the feedwater systems with no fluctuations in the reactor water level by increasing or decreasing the flow rate in the feedwater systems during automatic operation by the amount of fluctuations in the flow rate in the feedwater systems during manual operation and compensating the effects in upon manual and automatic switching by the M/A switcher. (Seki, T.)

  12. Auxiliary feedwater system aging study

    International Nuclear Information System (INIS)

    Kueck, J.D.

    1992-01-01

    The Phase 1 Auxiliary Feedwater (AFW) System Aging Study, NUREG/CR-5404 V1, focused on how and to what extent the various AFW system component types fail, how the failures have been and can be detected, and on the value of current testing requirements and practices. This follow-on study, which will be provided in full in NUREG/CR-5404 V2, provides a closure to the Phase 1 Study. For each of the component types and for the various sources of component failure identified in the Phase 1 Study, the methods of failure detection were designated and tabulated and the following findings became evident: Instrumentation and Control (I and C) related failures dominated the group of failures that were detected during demand conditions; many of the potential failure sources not detectable by the current monitoring practices were related to the I and C portion of the system; some component failure modes are actually aggravated by conventional test methods; and several important system functions did not undergo any function verification test. The goal of this follow-on study was to categorize and evaluate the deficiencies in testing identified by Phase 1 and to make specific recommendations for corrective action. In addition, this study presents discussions of alternate, state-of-the-art test methods, and provides a proposed Auxiliary Feedwater Pump test at normal operating pressure which should do much to verify system operability while eliminating degradation

  13. Auxiliary feedwater system aging study

    International Nuclear Information System (INIS)

    Kueck, J.D.

    1993-07-01

    This report documents the results of a Phase I follow-on study of the Auxiliary Feedwater (AFW) System that has been conducted for the US Regulatory Commission's Nuclear Plant Aging research Program. The Phase I study found a number of significant AFW System functions that are not being adequately tested by conventional test methods and some that are actually being degraded by conventional testing. Thus, it was decided that this follow-on study would focus on these testing omissions nd equipment degradation. The deficiencies in current monitoring and operating practice are categorized and evaluated. Areas of component degradation caused by current practice are discussed. Recommendations are made for improved diagnostic methods and test procedures

  14. Main feedwater valve diagnostics at Waterford 3 nuclear generating station

    International Nuclear Information System (INIS)

    Fitzgerald, W.V.

    1991-01-01

    Pneumatically-operated control valves are coming under increasing scrutiny in nuclear power plants because of their relatively high incident rate. The theory behind a device that could make performance evaluation of these valves simpler and more effective was first described at the original EPRI Power Plant Valve Symposium. The development of this Diagnostic System was completed in 1989, and it was recently used to troubleshoot two main feedwater valves at Louisiana Power and Light's Waterford 3 Power Station. During a cold snap last December, these valves failed to respond to the input signal and, as a result, the plant came off line. An incident report had to be filed, and the plant chose to contact the original equipment manufacturer (OEM) for assistance. This paper describes the original incident involving these valves and then gives a brief description of the diagnostic system and how it works. The balance of the paper then reviews how the OEM and plant personnel utilized the system to evaluate each component of the control valve assembly (I/P transducer, positioner, volume boosters, actuator, and valve body assembly). By simply stroking the valve and monitoring pneumatic signals and valve position, the problem was traced to a malfunctioning positioner and a volume booster that was leaking. The problems were corrected and new performance signatures run for the valves using the system to document their improved operation. This case study demonstrates how new Diagnostic Technology along with OEM involvement can effectively address problems with pneumatically-operated control valves so that root-cause solutions can be implemented

  15. Feedwater temperature control methods and systems

    Science.gov (United States)

    Moen, Stephan Craig; Noonan, Jack Patrick; Saha, Pradip

    2014-04-22

    A system for controlling the power level of a natural circulation boiling water nuclear reactor (NCBWR) is disclosed. The system, in accordance with an example embodiment of the present invention, may include a controller configured to control a power output level of the NCBWR by controlling a heating subsystem to adjust a temperature of feedwater flowing into an annulus of the NCBWR. The heating subsystem may include a steam diversion line configured to receive steam generated by a core of the NCBWR and a steam bypass valve configured to receive commands from the controller to control a flow of the steam in the steam diversion line, wherein the steam received by the steam diversion line has not passed through a turbine. Additional embodiments of the invention may include a feedwater bypass valve for controlling an amount of flow of the feedwater through a heater bypass line to the annulus.

  16. Aging assessment of auxiliary feedwater systems

    International Nuclear Information System (INIS)

    Casada, D.A.

    1989-01-01

    A study of Pressurized Water Reactor Auxiliary Feedwater (AFW) Systems has been conducted by Oak Ridge National Laboratory (ORNL) under the auspices of the Nuclear Regulatory Commission's Nuclear Plant Aging Research Program. The study has reviewed historical failure experience and current monitoring practices for the AFW System. This paper provides an overview of the study approach and results. 7 figs

  17. Considerations for surviving the loss of a main feedwater pump at full power

    International Nuclear Information System (INIS)

    Gaydos, K.A.; Calvo, R.; Conroy, P.W.; Klein, C.M.; Mellers, J.E.

    1990-01-01

    Today's economics dictate that nuclear power operational costs be contained by addressing frequently-occurring trips that might be minimized or avoided via specific upgrades. Much recent attention has focused on the significant percentage of plant trips related to feedwater flow regulation; however, another frequent feedwater-related trip stems from the loss of a single main feedwater pump while operating at high power levels, causing a plant trip on low steam generator water-level. This paper summarizes the results of several plant-specific studies that evaluate a unit's capabilities to consistently survive the loss of a main feedwater pump from full power, and outlines a methodology for analyzing this capability

  18. Analysis of containment parameters during the main steam line break with the failure of the feedwater control valves

    International Nuclear Information System (INIS)

    Fabjan, L.; Petelin, S.; Mavko, B.; Gortnar, O.; Tiselj, I.

    1992-01-01

    U.S. Nuclear Regulatory Commission (NRC) information notice 91-69: 'Errors in Main Steam Line Break Analyses for Determining Containment Parameters' shows the possibility of an accident which could lead to beyond design containment pressure and temperature. Such accident would be caused by the continuation of feedwater flow following a main stream line break (MSLB) inside the containment. Krsko power plant already experienced problems with main feedwater control valves. For that reason, analysis of MSLB has been performed taking into account continuous feedwater addition scenario and different containment safety systems capabilities availability. Steam and water released into the containment during MSLB was calculated using RELAP5/MOD2 computer code. The containment response to MSLB was calculated using CONTEMPT-LT/028 computer code. The results indicated that the continuous feedwater flow following a MSLB could lead to beyond design containment pressure. The peak pressure and temperature depend on isolation time for main- and auxiliary-feedwater supply. In the case of low boron concentration injection, the core recriticality is characteristic for this type of accidents. It was concluded that the presented analysis of MSLB with continuous feedwater addition scenario is the worst case for containment design

  19. Plant data comparisons for Comanche Peak 1/2 main feedwater pump trip transient

    Energy Technology Data Exchange (ETDEWEB)

    Boatwright, W.J.; Choe, W.G; Hiltbrand, D.W. [TU Electric, Dallas, TX (United States)] [and others

    1995-09-01

    A RETRAN-02 MOD5 model of Comanche Peak Steam Electric Station was developed by TU Electric for the purpose of performing core reload safety analyses. In order to qualify this model, comparisons against plant transient data from a partial loss of main feedwater flow were performed. These comparisons demonstrated that good representations of the plant response could be obtained with RETRAN-02 and the user-developed models of the primary-to-secondary heat transfer and plant control systems.

  20. Condensate and feedwater systems, pumps, and water chemistry. Volume seven

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Subject matter includes condensate and feedwater systems (general features of condensate and feedwater systems, condenser hotwell level control, condensate flow, feedwater flow), pumps (principles of fluid flow, types of pumps, centrifugal pumps, positive displacement pumps, jet pumps, pump operating characteristics) and water chemistry (water chemistry fundamentals, corrosion, scaling, radiochemistry, water chemistry control processes, water pretreatment, PWR water chemistry, BWR water chemistry, condenser circulating water chemistry

  1. System Study: Auxiliary Feedwater 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the auxiliary feedwater (AFW) system at 69 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the AFW results.

  2. Design and transient analyses of passive emergency feedwater system of CPR1000. Part 1. Air cooling condition

    International Nuclear Information System (INIS)

    Zhang Yapei; Qiu Suizheng; Su Guanghui; Tian Wenxi; Cao Jianhua; Lu Donghua; Fu Xiangang

    2011-01-01

    The steam generator secondary passive emergency feedwater system is a new design for traditional generation Ⅱ + reactor CPR1000. The passive emergency feedwater system is designed to supply water to the SG shell side and improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the feed line break (FLB) or loss of heat sink accident. The passive emergency feedwater system consists of steam generator (SG), heat exchanger (HX), air cooling tower, emergency makeup tank (EMT), and corresponding pipes and valves for air cooling condition. In order to improve the safety and reliability of CPR1000, the model of the primary loop system and the passive emergency feedwater system was developed to investigate residual heat removal capability of the passive emergency feedwater system and the transient characteristics of the primary loop system affected by the passive emergency feedwater system using RELAP5/MOD3.4. The transient characteristics of the primary loop system and the passive emergency feedwater system were calculated in the event of feed line break accident. Sensitivity studies of the passive emergency feedwater system were also conducted to investigate the response of the primary loop and the passive emergency feedwater system on the main parameters of the passive emergency feedwater system. The passive emergency feedwater system could supply water to the SG shell side from the EMT successfully. The calculation results showed that the passive emergency feedwater system could take away the decay heat from the primary loop effectively for air cooling condition, and that the single-phase and two-phase natural circulations were established in the primary loop and passive emergency feedwater system loop, respectively. (author)

  3. Review of the Shearon Harris Unit 1 auxiliary feedwater system reliability analysis

    International Nuclear Information System (INIS)

    Fresco, A.; Youngblood, R.; Papazoglou, I.A.

    1986-02-01

    This report presents the results of a review of the Auxiliary Feedwater System Reliability Analysis for the Shearon Harris Nuclear Power Plant (SHNPP) Unit 1. The objective of this report is to estimate the probability that the Auxiliary Feedwater System will fail to perform its mission for each of three different initiators: (1) loss of main feedwater with offsite power available, (2) loss of offsite power, (3) loss of all ac power except vital instrumentation and control 125-V dc/120-V ac power. The scope, methodology, and failure data are prescribed by NUREG-0611 for other Westinghouse plants

  4. Feedwater recycling system in BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To improve the reactor safety by preventing thermal stresses and cracks generated in structural materials due to the fluctuations in the temperature for high temperature water - low temperature water mixture near the feedwater nozzle. Method: Feedwater pipes are connected to a pressure vessel not directly but by way of a flow control valve. While the recycled water is circulated from an inlet nozzle to an outlet nozzle through a recycle pump, flow control valve and recycling pipeways, feedwater is fed from the feedwater pipes to the recycling pipeways by way of the flow control valve. More specifically, since the high temperature recycle water and the low temperature recycle water are mixed within the pipeways, the temperature fluctuations resulted from the temperature difference between the recycle water and the feedwater is reduced to prevent thermal fatigue and generation of cracks thereby securing the reactor safety. (Furukawa, Y.)

  5. Operating experiences and degradation detection for auxiliary feedwater systems

    International Nuclear Information System (INIS)

    Casada, D.; Farmer, W.S.

    1992-01-01

    A study of Pressurized Water Reactor Auxiliary Feedwater (AFW) Systems has been conducted by Oak Ridge National Laboratory (ORNL) under the auspices of the Nuclear Regulatory Commission's Nuclear Plant Aging Research Program. The results of the study are documented in NUREG/CR-5404, Vol. 1, Auxiliary Feedwater System Aging Study. The study reviewed historical failure experience and current monitoring practices for the AFW System. This paper provides an overview of the study approach and results

  6. Simulation of a passive auxiliary feedwater system with TRACE5

    Energy Technology Data Exchange (ETDEWEB)

    Lorduy, María; Gallardo, Sergio; Verdú, Gumersindo, E-mail: maloral@upv.es, E-mail: sergalbe@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Instituto Universitario de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), València (Spain)

    2017-07-01

    The study of the nuclear power plant accidents occurred in recent decades, as well as the probabilistic risk assessment carried out for this type of facility, present human error as one of the main contingency factors. For this reason, the design and development of generation III, III+ and IV reactors, which include inherent and passive safety systems, have been promoted. In this work, a TRACE5 model of ATLAS (Advanced Thermal- Hydraulic Test Loop for Accident Simulation) is used to reproduce an accidental scenario consisting in a prolonged Station BlackOut (SBO). In particular, the A1.2 test of the OECD-ATLAS project is analyzed, whose purpose is to study the primary system cooling by means of the water supply to one of the steam generators from a Passive Auxiliary Feedwater System (PAFS). This safety feature prevents the loss of secondary system inventory by means of the steam condensation and its recirculation. Thus, the conservation of a heat sink allows the natural circulation flow rate until restoring stable conditions. For the reproduction of the test, an ATLAS model has been adapted to the experiment conditions, and a PAFS has been incorporated. >From the simulation test results, the main thermal-hydraulic variables (pressure, flow rates, collapsed water level and temperature) are analyzed in the different circuits, contrasting them with experimental data series. As a conclusion, the work shows the TRACE5 code capability to correctly simulate the behavior of a passive feedwater system. (author)

  7. Expert system for nuclear power plant feedwater system diagnosis

    International Nuclear Information System (INIS)

    Meguro, R.; Kinoshita, Y.; Sato, T.; Yokota, Y.; Yokota, M.

    1987-01-01

    The Expert System for Nuclear Power Plant Feedwater System Diagnosis has been developed to assist maintenance engineers in nuclear power plants. This system adopts the latest process computer TOSBAC G8050 and the expert system developing tool TDES2, and has a large scale knowledge base which consists of the expert knowledge and experience of engineers in many fields. The man-machine system, which has been developed exclusively for diagnosis, improves the man-machine interface and realizes the graphic displays of diagnostic process and path, stores diagnostic results and searches past reference

  8. Excessive heat removal due to feedwater system malfunction

    International Nuclear Information System (INIS)

    Beader, D.; Peterlin, G.

    1986-01-01

    Excessive heat removal transient of the Krsko Nuclear Power Plant, caused by steam generators feedwater system malfunctions was simulated by RELAP5/MOD1 computer code. The results are increase of power and reactor scram caused by high-high steam generator level. (author)

  9. Influence of reactor vessel nodalization in the coupled code analysis of Asymmetric Main Feedwater Isolation

    International Nuclear Information System (INIS)

    Bencik, V.; Feretic, D.; Grgic, D.

    2001-01-01

    Asymmetric Main Feedwater Isolation (AMFWI) transient in one Steam Generator (SG) for NPP Krsko using RELAP5 standalone code and coupled code RELAP5- QUABOX/CUBBOX (R5QC) was analyzed. In the RELAP5 standalone calculation, a point kinetics model was used, while in the coupled code a three-dimensional (3D) neutronics model of QUABOX with different RELAP5 nodalization schemes of reactor vessel was used. Both code versions use best-estimate thermal-hydraulic system code for all components in the plant and include realistic description of plant protection and control systems. Two different types of calculations were performed: with and without automatic control rod system available. The AMFWI transient causes the great asymmetry of the transferred heat in the SGs and subsequently the asymmetry of the power produced across the core due to different reactivity feedback resulting from the thermal-hydraulic channels assigned to different loops. The work presented in the paper is a part of validation of the 3D coupled code R5QC in the analysis of asymmetric transients.(author)

  10. Aging assessment of PWR [Pressurized Water Reactor] Auxiliary Feedwater Systems

    International Nuclear Information System (INIS)

    Casada, D.A.

    1988-01-01

    In support of the Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, Oak Ridge National Laboratory is conducting a review of Pressurized Water Reactor Auxiliary Feedwater Systems. Two of the objectives of the NPAR Program are to identify failure modes and causes and identify methods to detect and track degradation. In Phase I of the Auxiliary Feedwater System study, a detailed review of system design and operating and surveillance practices at a reference plant is being conducted to determine failure modes and to provide an indication of the ability of current monitoring methods to detect system degradation. The extent to which current practices are contributing to aging and service wear related degradation is also being assessed. This paper provides a description of the study approach, examples of results, and some interim observations and conclusions. 1 fig., 1 tab

  11. Identification of BWR feedwater control system using autoregressive integrated model

    International Nuclear Information System (INIS)

    Kanemoto, Shigeru; Andoh, Yasumasa; Yamamoto, Fumiaki; Idesawa, Masato; Itoh, Kazuo.

    1983-01-01

    With the view of contributing toward more reliable interpretation of noise behavior under normal operating conditions, which is essential for correct detection and/or diagnosis of incipient anomalies in nuclear power plants by noise analysis technique, studies has been undertaken of the noise behavior in a BWR feedwater control system, with use made of a multivariate autoregressive modeling technique. Noise propagation mechanisms as well as open- and closed-loop responses in the system are identified from noise data by a method in which an autoregressive integrated model is introduced. The closed-loop responses obtained with this method are compared with transient data from an actual test, and confirmed to be reliable in estimating semi-quantitative features. Other analyses performed with this model also yield results that appear most reasonable in their physical characteristics. These results have demonstrated the effectiveness of the noise analyses technique based on the autoregressive integrated model for evaluating and diagnosing the performance of feedwater control systems. (author)

  12. A probabilistic evaluation of the Shearon Harris Nuclear Power Plant auxiliary feedwater isolation system

    International Nuclear Information System (INIS)

    Anoba, R.C.

    1989-01-01

    This paper reports on a fault tree approach that was used to evaluate the safety significance of modifying the Shearon Harris Auxiliary Feedwater Isolation System. The design modification was a result of on-site reviews which identified a single failure in the Auxiliary Feedwater Isolation circuitry

  13. Reactor feedwater facility

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Tadashi; Kinoshita, Shoichiro; Akatsu, Jun-ichi

    1996-04-30

    In a reactor feedwater facility in which one stand-by system and at least three ordinary systems are disposed in parallel, each of the feedwater pumps is driven by an electromotor, and has substantially the same capacity. At least two systems among the ordinary systems have a pump rotation number variable means. Since the volume of each of the feedwater pump of each system is determined substantially equal, standardization is enabled to facilitate the production. While the number of electromotors is increased, since they are driven by electromotors, turbines, steam pipelines and valves for driving feed water pumps can be eliminated. Therefore, the feedwater pumps can be disposed to a region of low radiation dose being separated from a main turbine and a main condensator, to improve the degree of freedom in view of the installation. In addition, accessibility to equipments during operation is improved to improve the maintenance of feed water facilities. The number of parts for equipments can be reduced compared with that in a turbine-driving system thereby capable of reducing the operation amount for the maintenance and inspection. (N.H.)

  14. A reliability centered maintenance model applied to the auxiliary feedwater system of a nuclear power plant

    International Nuclear Information System (INIS)

    Araujo, Jefferson Borges

    1998-01-01

    The main objective of maintenance in a nuclear power plant is to assure that structures, systems and components will perform their design functions with reliability and availability in order to obtain a safety and economic electric power generation. Reliability Centered Maintenance (RCM) is a method of systematic review to develop or optimize Preventive Maintenance Programs. This study presents the objectives, concepts, organization and methods used in the development of RCM application to nuclear power plants. Some examples of this application are included, considering the Auxiliary Feedwater System of a generic two loops PWR nuclear power plant of Westinghouse design. (author)

  15. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

    Lu, S.C.; Tsai, N.C.

    1983-08-01

    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14

  16. Analysis of limit cycling on a boiler feedwater control system

    International Nuclear Information System (INIS)

    Thomas, P.J.; Harrison, T.A.; Hollywell, P.D.

    1986-01-01

    During operation of the UKAEA Prototype Fast Reactor, it was found that oscillations sometimes occurred in the boiler feedwater systems. These were normally of relatively low amplitude, but led to the adoption of low controller gains so that control was rather slack. While control performance proved generally adequate for steady running, the lack of tight control of steam drum levels sometimes led to difficulties during periods when plant conditions were undergoing major change. The paper discusses the methods used to gain a full understanding of the phenomena occurring, and describes how that knowledge is being used to improve the control system so as to eliminate the limit cycling modes and ensure good control of steam drum levels. A noteworthy feature of the study was the use of two independent representations of plant behaviour: (i) a frequency response model, FWRFREQ, and (ii) a time-domain simulation model, PFRTDM. The simplified analysis of FWRFREQ proved to be of enormous value in identifying modes of system behaviour; PFRTDM was used as a detailed check on the accuracy and validity of the results obtained. (author)

  17. Assessment of RELAP5/MOD2 against a main feedwater turbopump trip transient in the Vandellos II Nuclear Power Plant

    International Nuclear Information System (INIS)

    Llopis, C.; Casals, A.; Perez, J.; Mendizabal, R.

    1993-12-01

    The Consejo de Seguridad Nuclear (CSN) and the Asociacion Nuclear Vandellos (ANV) have developed a model of Vandellos II Nuclear Power Plant. The ANV collaboration consisted in the supply of design and actual data, the cooperation in the simulation of the control systems and other model components, as well as in the results analysis. The obtained model has been assessed against the following transients occurred in plant: A trip from the 100% power level (CSN); a load rejection from 100% to 50% (CSN); a load rejection from 75% to 65% (ANV); and, a feedwater turbopump trip (ANV). This copy is a report of the feedwater turbopump trip transient simulation. This transient actually occurred in the plant on June 19, 1989

  18. Getting the most out of your new plant with a chordal ultrasonic feedwater flow measurement system

    International Nuclear Information System (INIS)

    Estrada, Herb; Hauser, Ernie

    2007-01-01

    The economic advantages of a chordal ultrasonic feedwater flow measurement system over conventional (flow nozzle-based) feedwater instrumentation are analyzed for new plants having ratings ranging from 1100 MWe to 1600 MWe. Specifically, each of the following topics is considered: The value of a 1.7% increase in the rating of the new plant, made possible by the reduced uncertainty in the determination of thermal power. The value of reduced startup time owing to enhanced steam supply water level control. The value of the reduced feedwater pumping power brought about by the elimination of flow nozzles. The value of the reduced calibration burden owing to the elimination of the feedwater flow differential pressure transmitters and resistance thermometers. The net difference in the acquisition costs of the ultrasonic system versus conventional feedwater flow instrumentation. The net savings in installation costs of the ultrasonic system vis-a-vis conventional feedwater flow instrumentation. The potential savings in outage time due to the reduced frequency of low steam supply water level trips (scrams) of the reactor. (author)

  19. A novel feedwater system for the RETRAN model of the Palo Verde nuclear generating station

    International Nuclear Information System (INIS)

    Secker, P.A.; Webb, J.R.

    1988-01-01

    This paper presents a feedwater system model which supplies realistic boundary conditions to the RETRAN model of a Palo Verde Nuclear Generating Station reactor plant. The RETRAN thermal hydraulic code is used to analyze nuclear reactor system transients through a generalized thermal hydraulic volume/junction network. The feedwater system model is implemented using the control block modeling option available in the RETRAN code. The output of the control block model is coupled to the thermal hydraulic network by a fill junction. A forward Euler integration scheme is used by RETRAN for control block variables. The feedwater system model is formulated to allow implicit integration within the existing code framework. The potential need for small integration time steps is, therefore, alleviated. The model results are compared with test data

  20. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Papp, L. [Inst. of Material Engineering, Ostrava (Switzerland)

    1995-12-31

    After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed.

  1. Dynamic analysis of the condensate feedwater system in boiling water reactor plants

    International Nuclear Information System (INIS)

    Tanji, J.; Omori, T.

    1982-01-01

    The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents

  2. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    International Nuclear Information System (INIS)

    Pappx, L.

    1994-01-01

    After modification of Dukovany NPP steam generator feedwater system, the increased concentration of minerals was measured in the cold leg of modified steam generator. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators, has focused this attention on the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of flow distribution in the secondary side of SG was developed. (Author)

  3. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Papp, L [Inst. of Material Engineering, Ostrava (Switzerland)

    1996-12-31

    After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed.

  4. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    International Nuclear Information System (INIS)

    Papp, L.

    1995-01-01

    After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed

  5. Feedwater processing method in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izumitani, M; Tanno, K

    1976-09-06

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

  6. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    International Nuclear Information System (INIS)

    Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

    2014-01-01

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection

  7. 77 FR 15812 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

    Science.gov (United States)

    2012-03-16

    ... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... Feedwater Systems for Light- Water Reactors.'' DG-1265 is proposed revision 2 of Regulatory Guide 1.68.1... Plants,'' dated January 1977. This regulatory guide is being revised to: (1) expand the scope of the...

  8. Reliability analysis of the auxiliary feedwater system; Analiza zanesljivosti sistema pomozne napajalne vode

    Energy Technology Data Exchange (ETDEWEB)

    Susnik, J; Dusic, M [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1984-07-01

    The reliability of a NPP auxiliary feedwater system is evaluated using the fault tree analysis. The system is analyzed during the time interval 0 to 6 hours with the computer package program PREP/KITT which is described in more detail. (author)

  9. Mobile polishing system of feedwater at start-up feedback from the implementation and future prospects

    International Nuclear Information System (INIS)

    Faure, Celine; Eade, Kevin; Fontan, Guillaume

    2012-09-01

    The reduction of the quantity of Steam Generator (SG) metallic oxides deposits, and maintaining a good chemical composition of the secondary side of SG tubes are some of the main objectives being looked at, in order to reduce the risk of SG corrosion, regardless of the alloy used, right from the start-up phase. For all types of outage, obtaining and maintaining sufficient chemical cleanliness at the start-up requires treatment of the water. The treatments are notably: - Water movements using the purge / make-up water method until the chemical criteria have been met. This method can be long and generate large volumes of discharge. - Using suitable resins to remove pollutants from the water. The advantage of this method is that it is selective. - Filtration, allowing for the removal of any insoluble agent. In order to optimise the start-up process, Gravelines and Blayais Nuclear Power Plants (NPPs) put trials in place towards the end of the 1980s. These trials lead to a water supply treatment installation (mobile polishing system- in French Systeme Mobile d'Epuration, SME) being put in place for the start-up phase, made up of an up-stream filter, a mixed-bed resin pollutant trap and a down-stream filter to prevent losing the fines into the feedwater. At the same time, the manifestation of cracking on the secondary side of the steam generator tubes lead EDF to roll out a water treatment for the feedwater dedicated to the start-up. The choice was made not to install a condensate polishing plant, in order to limit notably the pollution risks (resin leaks or waste from the regeneration in the backwater) following difficulties during regeneration. The positive results from the first trials validated for EDF the choice to give priority to the roll-out of the SME to the NPPs judged to be most critical due to the SG material. The SME, installed on a mobile base, can be used on different units at the same station; this reduced the investment and maintenance costs, and

  10. Reliability analysis of the auxiliary feedwater system of Angra-1 including common cause failures using the multiple greek letter model

    International Nuclear Information System (INIS)

    Lapa, Celso Marcelo Franklin.

    1996-05-01

    The use of redundancy to increase the reliability of industrial systems make them subject to the occurrence of common cause events. The industrial experience and the results of safety analysis studies have indicated that common cause failures are the main contributors to the unreliability of plants that have redundant systems, specially in nuclear power plants. In this Thesis procedures are developed in order to include the impact of common cause failures in the calculation of the top event occurrence probability of the Auxiliary Feedwater System in a typical two-loop Nuclear Power Plant (PWR). For this purpose the Multiple Greek Letter Model is used. (author). 14 refs., 10 figs., 11 tabs

  11. Implementation of a digital feedwater control system at Dresden Nuclear Power Plant, Units 2 and 3: Final report

    International Nuclear Information System (INIS)

    Zapotocky, A.; Popovic, J.R.; Fournier, R.D.

    1988-12-01

    This report describes the Digital Feedwater Control System Implementation at the Dresden 2 or 3 Units of the BWR Nuclear Power Plant owned by the Commonwealth Edison Company. The digital system has been operational in Unit 3 since August 1986, and in Unit 2 since April 1987. The Bailey Control's Network 90 based digital control system replaced the obsolete GE/MAC 5000 analog control system in the reactor feedwater control loop as a ''like-for-like'' replacement. Operational experience from the Digital Feedwater Control installations has been good and the system demonstrated better performance than the old analog systems. 14 refs., 15 figs., 17 tabs

  12. Adaptation of computer code ALMOD 3.4 for safety analyses of Westighouse type NPPs and calculation of main feedwater loss

    International Nuclear Information System (INIS)

    Kordis, I.; Jerele, A.; Brajak, F.

    1986-01-01

    The paper presents theoretical foundations of ALMOD 3.4 code and modification done in order to adjust the model to westinghouse type NPP. test cases for verification of added modules functioning were done and loss of main feedwater (FW) transient at nominal power was analysed. (author)

  13. Instrument failure detection of flow measurement in the feedwater system of the Paks Nuclear Power Plant, Hungary

    International Nuclear Information System (INIS)

    Racz, A.

    1990-12-01

    The applicability of two different methods for early detection of instrument failures of the flow measurement in feedwater systems are investigated. Both methods are based on Kalman filtering technique of stochastic processes. The reliability of the model for description of a feedwater system is checked by comparing calculated values with measured data. Possible instrument failures are simulated in order to show the capability of the proposed procedures. A practical measurement system arrangement is suggested. (author) 10 refs.; 16 figs.; 4 tabs

  14. Common-cause failure analysis of McGuire Unit 2 auxiliary feedwater system

    International Nuclear Information System (INIS)

    Rasmuson, D.M.; Shepherd, J.C.; Fowler, R.D.; Summitt, R.L.; Logan, B.W.

    1982-01-01

    A powerful method for qualitative common cause failure analysis (CCFA) of nuclear power plant systems was developed by EG and G Idaho at the Idaho National Engineering Laboratory. As a cooperative project to demonstrate and evaluate the usefulness of the method, the Duke Power Company agreed to allow a CCFA of the auxiliary feedwater system (AFWS) in their McGuire Nuclear Station Unit 2. The results of the CCFA are the subject of this discussion

  15. PSA effect analysis of a design modification of the auxiliary feedwater system for a Westinghouse type plant

    International Nuclear Information System (INIS)

    Bae, Yeon Kyoung; Lee, Eun Chan

    2012-01-01

    The auxiliary feedwater system is an important system used to mitigate most accidents considered in probabilistic safety assessment (PSA). The reference plant has produced electric power for about thirty years. Due to age related deterioration and lack of parts, a turbine driven auxiliary feedwater pump (TD AFWP), some valves, and piping of the auxiliary feedwater system should be replaced. This change includes relocation of some valves, installation of valves for maintenance of the steam generator, and a new cross tie line. According to the design change, the Final Safety Analysis Report (FSAR) has been revised. Therefore, this design modification affects the PSA. It is thus necessary to assess the improvement of plant safety. In this paper, the impact of the design change of the auxiliary feedwater system on the PSA is assessed. The results demonstrate that this modification considering the plant safety decreased the total CDF

  16. Using risk-informed asset management for feedwater system preventative maintenance optimization

    International Nuclear Information System (INIS)

    Kee, Ernest; Sun, Alice; Richards, Andrew; Grantom, Rick; Liming, James; Salter, James

    2004-01-01

    The initial development of a South Texas Project Nuclear Operating Company process for supporting preventative maintenance optimization by applying the Balance-Of-Plant model and Risk-Informed Asset Management alpha-level software applications is presented. Preventative maintenance activities are evaluated in the South Texas Project Risk-Informed Asset Management software while the plant maintains or improves upon high levels of nuclear safety. In the Balance-Of-Plant availability application, the level of detail in the feedwater system is enhanced to support plant decision-making at the component failure mode and human error mode level of indenture by elaborating on the current model at the super-component level of indenture. The enhanced model and modeling techniques are presented. Results of case studies in feedwater system preventative maintenance optimization sing plant-specific data are also presented. (author)

  17. The effects of parameter variation on MSET models of the Crystal River-3 feedwater flow system

    International Nuclear Information System (INIS)

    Miron, A.

    1998-01-01

    In this paper we develop further the results reported in Reference 1 to include a systematic study of the effects of varying MSET models and model parameters for the Crystal River-3 (CR) feedwater flow system The study used archived CR process computer files from November 1-December 15, 1993 that were provided by Florida Power Corporation engineers Fairman Bockhorst and Brook Julias. The results support the conclusion that an optimal MSET model, properly trained and deriving its inputs in real-time from no more than 25 of the sensor signals normally provided to a PWR plant process computer, should be able to reliably detect anomalous variations in the feedwater flow venturis of less than 0.1% and in the absence of a venturi sensor signal should be able to generate a virtual signal that will be within 0.1% of the correct value of the missing signal

  18. Reactor feedwater device

    International Nuclear Information System (INIS)

    Igarashi, Noboru.

    1986-01-01

    Purpose: To suppress soluble radioactive corrosion products in a feedwater device. Method: In a light water cooled nuclear reactor, an iron injection system is connected to feedwater pipeways and the iron concentration in the feedwater or reactor coolant is adjusted between twice and ten times of the nickel concentration. When the nickel/iron ratio in the reactor coolant or feedwater goes nearer to 1/2, iron ions are injected together with iron particles to the reactor coolant to suppress the leaching of stainless steels, decrease the nickel in water and increase the iron concentration. As a result, it is possible to suppress the intrusion of nickel as one of parent nuclide of radioactive nuclides. Further, since the iron particles intruded into the reactor constitute nuclei for capturing the radioactive nuclides to reduce the soluble radioactive corrosion products, the radioactive nuclides deposited uniformly to the inside of the pipeways in each of the coolant circuits can be reduced. (Kawakami, Y.)

  19. Loss of main and auxiliary feedwater event at the Davis-Besse Plant on June 9, 1985

    International Nuclear Information System (INIS)

    1985-07-01

    On June 9, 1985, Toledo Edison Company's Davis-Besse Nuclear Power Plant, located in Ottawa County, Ohio, experienced a partial loss of feedwater while the plant was operating at 90% power. Following a reactor trip, a loss of all feedwater occurred. The event involved a number of equipment malfunctions and extensive operator actions, including operator actions outside the control room. Several operator errors also occurred during the event. This report documents the findings of an NRC Team sent to Davis-Besse by the NRC Executive Director for Operations in conformance with the staff-proposed Incident Investigation Program

  20. Implementation of an advanced digital feedwater control system at the Prairie Island nuclear generating station

    International Nuclear Information System (INIS)

    Paris, R.E.; Gaydos, K.A.; Hill, J.O.; Whitson, S.G.; Wirkkala, R.

    1990-05-01

    EPRI Project RP2126-4 was a cooperative effort between TVA, EPRI, and Westinghouse which resulted in the demonstration of a prototype of a full range, fully automatic feedwater control system, using fault tolerant digital technology, at the TVA Sequoyah simulator site. That prototype system also included advanced signal validation algorithms and an advanced man-machine interface that used CRT-based soft-control technology. The Westinghouse Advanced Digital Feedwater Control System (ADFCS) upgrade, which contains elements that were part of that prototype system, has since been installed at Northern States Power's Prairie Island Unit 2. This upgrade was very successful due to the use of an advanced control system design and the execution of a well coordinated joint effort between the utility and the supplier. The project experience is documented in this report to help utilities evaluate the technical implications of such a project. The design basis of the Prairie Island ADFCS signal validation for input signal failure fault tolerance is outlined first. Features of the industry-proven system control algorithms are then described. Pre-shipment hardware-in-loop and factory acceptance testing of the Prairie Island system are summarized. Post-shipment site testing, including preoperational and plant startup testing, is also summarized. Plant data from the initial system startup is included. The installation of the Prairie Island ADFCS is described, including both the feedwater control instrumentation and the control board interface. Modification of the plant simulator and operator and I ampersand C personnel training are also discussed. 6 refs., 14 figs., 3 tabs

  1. Boiler feedwater quality improvement by replacing conventional pre-treatment with advanced membrane systems

    Energy Technology Data Exchange (ETDEWEB)

    Doll, Bernhard [Process Systems Pall GmbH, Dreieich (Germany). Marketing; Venkatadri, Ramraj [Pall Corporation, Port Washington, NY (United States). Global Marketing Energy

    2013-09-01

    Two case studies in different application fields highlight significant economical and operational improvements that were achieved by replacing conventional water treatment technologies by highly-sophisticated membrane systems. The first case study deals with boiler feedwater in a power plant, focusing on the challenges faced as well as the direct and indirect benefits gained by the new system within a utility station. The second case study deals with the conventional water treatment scheme for groundwater from 13 wells at a major oil sands facility. Operational performance as well as the cost improvements gained in both cases will be presented. (orig.)

  2. Probabilistic analysis of reactor safety - The auxiliary feedwater system of Angra I

    International Nuclear Information System (INIS)

    Oliveira, L.C.R. da L.C. de.

    1981-09-01

    The unavailability of the auxiliary feedwater system (AFWS) of Angra-1, was calculated. The fault tree analysis technique was used, considering two diferent types of contribution to system unavailability: The one due to hard-ware failure and the contribution due to test and maintenance which was separately analysed. The COMBO-and SAMPLE computer codes were used. The results have shown that the AFWS of Angra-1 contains enough redundancy to guarantee a safe operation under the conditions analysed, best values having been obtained for the unavailability of AFWS of Angra 1 with those codes than with the WASH-1400. (E.G.) [pt

  3. Assessment of a potential rapid condensation induced water hammer in a passive auxiliary feedwater system

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Shin, Byung Soo; Do, Kyu Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Moody, Frederick J. [General Electric (Retired), CA (United States)

    2012-10-15

    A passive auxiliary feedwater system (PAFS) which is incorporated in the APR+ system is a kind of closed natural circulation loop. The PAFS has no operating functions during normal plant operation, but it has a dedicated safety function of the residual heat removal following initiating events, including the unlikely event of the most limiting single failure occurring coincident with a loss of offsite power, when the feedwater system becomes inoperable or unavailable. Even in the unlikely event of a station blackout, the isolation valves can be opened either by DC power or manual operation and then the PAFS can also provide adequate condensate to the steam generator (SG). The PAFS piping in the vicinity of each of the two SGs is designed to minimize the potential for destructive water hammer during start up operation by setting the stroke time for full close or full open of the condensate isolation valves upon receipt of a passive auxiliary feedwater actuation signal. The temperature of the stagnant condensate water and its surrounding tubes and piping during the reactor normal operation modes may fall to the ambient temperature. A possible concern is the introduction of saturated steam into the PAFS recirculation pipe downstream of the PCHX in the beginning of the PAFS operation. Although the steam introduction rate is expected to be slow, a rapid condensation rate is expected due to the initial cold surrounding temperature in the pipe, which could result in a localized pressure reduction and the propagation of decompression and velocity disturbances into the condensate water leg, which might cause the sudden closure of check valves and associated water hammer. Thus, it is requisite for the licensing review of the PAFS design to confirm if destructive water hammers will not be produced due to such rapid condensation induced decompressions in the system. This paper addresses an assessment of the potential local decompressions which could result from the steam

  4. Reactor feedwater pump control device

    International Nuclear Information System (INIS)

    Nishiyama, Hiroyuki.

    1990-01-01

    An amount of feedwater necessary for ensuring reactor inventory after scram is ensured automatically based on the reactor output before scram of a BWR type reactor. That is, if scram should occur, a feedwater flow rate just before the scram is stored by reactor output signals. Further, the amount of feedwater required after the scram is determined based on the output of the memory. The reactor power after the scram based on a feedwater flow rate and a main steam flow rate is inputted to an integrator, to calculate and output the amount of the feedwater flow rate (1) injected after the scram for the inventory. A coast down flowrate (2) in a case of pump trip is forecast by the output signals. Automatic trip is outputted to all turbine driving feedwater pumps when the sum of (1) and (2) exceeds a necessary and sufficient amount of feedwater required for ensuring inventory. For motor driving feedwater pumps, only a portion, for example, one of the pumps is automatically started while other pumps are stopped their operation, only in this case, to prevent excess water feeding. (I.S.)

  5. Modernization of the feedwater heaters control level of the Almaraz I Nuclear Power Plant by OVATION system

    International Nuclear Information System (INIS)

    Madronal Rodriguez, E.; Cabrero Munoz, J. E.

    2010-01-01

    As a result of the process of technological renovation of the heaters system and the power increase project, Almaraz Nuclear Power Plant has made several design changes in the feedwater heaters system. Within these changes, the old heaters control loops are replaced because the new power will increase the heaters drainage caudal. This modernization is carried out using the OVATION control system.

  6. Feedwater control method and device therefor

    International Nuclear Information System (INIS)

    Nakahara, Mitsugu; Ichikawa, Yoshiaki; Ishii, Yoshikazu; Suzuki, Katsuyuki; Tanikawa, Naoshi; Mizuki, Fumio.

    1997-01-01

    The present invention provides a method of and a device for easily changing the constitution of feedwater systems without causing change in the water level of a reactor even when a plurality of feedwater systems have imbalance points. Namely, a feedwater control device comprises at least two feedwater systems capable of feeding water to tanks independently respectively and a controller capable of controlling water level in the tanks by controlling these feedwater systems. There is disposed a means for outputting gradually increasing driving signals to other feedwater systems, when the water level controller automatically controls one of the feedwater systems. There is also disposed a means for switching from automatic control for one of the feedwater systems to automatic control for the other feedwater system by a water level controller when the other feedwater system is in a stable operation region. As a result, entire feedwater flow rate is not temporarily changed and the water level in the tanks can be maintained constant. (N.H.)

  7. Steady state flow evaluations for passive auxiliary feedwater system of APR

    International Nuclear Information System (INIS)

    Park, Jongha; Kim, Jaeyul; Seong, Hoje; Kang, Kyoungho

    2012-01-01

    This paper briefly introduces a methodology to evaluate steady state flow of APR+ Passive Auxiliary Feedwater System (PAFS). The PAFS is being developed as a safety grade passive system to completely replace the existing active Auxiliary Feedwater System (AFWS). Natural circulation cooling can be generally classified into the single-phase, two-phase, and boiling-condensation modes. The PAF is designed to be operated in a boiling-condensation natural circulation mode. The steady-state flow rate should be equal to the steady-state boiling/condensation rate determined by the steady-state energy and momentum balances in the PAFS. The determined steady-state flow rate can be used in the design optimization for the natural circulation loop of the PAFS through the steady-state momentum balance. Since the retarding force, which is to be balanced by the driving force in the natural circulation system design depends on the reliable evaluation of the success of a natural circulation system design depends on the reliable evaluation of the pressure loss coefficients. In PAFS, the core decay heat is released by natural circulation flow between the S G secondary side and the Passive Condensation Heat Exchanger (PCHX) that is immersed in the Passive Condensation Cooling Tank (PCCT). The PCCT is located on the top of Auxiliary building The driving force is determined by the difference between the S/G (heat Source) secondary water level and condensation liquid (heat sink) level. It will overcome retarding force at flowrate in the system, which is determined by vaporization and condensation of the steam which is generated at the S/G by the latent heat in system. In this study, the theoretical method to estimate the steady state flow rate in boiling-condensation natural circulation system is developed and compared with test results

  8. San Onofre/Zion auxiliary feedwater system seismic fault tree modeling

    International Nuclear Information System (INIS)

    Najafi, B.; Eide, S.

    1982-02-01

    As part of the study for the seismic evaluation of the San Onofre Unit 1 Auxiliary Feedwater System (AFWS), a fault tree model was developed capable of handling the effect of structural failure of the plant (in the event of an earthquake) on the availability of the AFWS. A compatible fault tree model was developed for the Zion Unit 1 AFWS in order to compare the results of the two systems. It was concluded that if a single failure of the San Onofre Unit 1 AFWS is to be prevented, some weight existing, locally operated locked open manual valves have to be used for isolation of a rupture in specific parts of the AFWS pipings

  9. Feedwater heater

    International Nuclear Information System (INIS)

    Murata, Shigeto; Minato, Akihiko; Yokomizo, Osamu; Masuhara, Yasuhiro.

    1991-01-01

    The present invention concerns a feedwater heater for a BWR type reactor. A cylinder is fit into the lower portion of a drain inlet pipe, to which drain water inflows from a turbine, and a disk is disposed to the lower end of the cylinder vertically to the axis of the cylinder, to constitute a drain water dispersing mechanism. Drain water inflown from the drain inlet pipe is fallen in the cylinder and collides against the disk. The collided drain water is splashed horizontally by its kinetic energy to reach the heat transfer pipe and conducts heat exchange. In this case, the drain water is converted into fine droplets by the collision against the disk and scattered in a wide range in the heater. As a result, sensible heat in the drain water can be transferred to feedwater effectively. Then, even the heat energy of the drain water can be utilized effectively for heat exchange, to improve the heat exchange efficiency. (I.N.)

  10. Single-tube condensation experiment in Passive Auxiliary Feedwater System of APR1400+

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang Wook; No, Hee Cheon; Yun, Bong Yo; Jeon, Byong Guk [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2012-05-15

    Conventional Korean nuclear power plants, Advanced Power Reactors (APR), are characterized by an active cooling system. However, Active cooling system may not prevent significant damage without any AC power source available for its operation as vividly illustrated through the recent Fukushima incident. In the APR1400+ to be designed, an independent passive cooling system was added in order to overcome the aforementioned shortcomings. In the Passive Auxiliary Feedwater System (PAFS), gravity force and density difference between steam and water are used. The system comprises of 240 condensation tubes to efficiently remove decay heat. Before applying the PAFS to APR1400+, the system's safety and heat removal performance must be verified. The present study experimentally evaluates the heat removal performance of a single tube in the PAFS. The objectives of SCOP (Single-tube Condensation experiment facility of PAFS) are the evaluation of the heat removal performance in the tube of the PAFS and database construction under various tube designs and test conditions. Reaching these objectives, we developed advanced measurement techniques for the amount of moisture, heat flux, and water film thickness.

  11. Reliability analysis of 2 types of auxiliary feedwater system for PWR

    International Nuclear Information System (INIS)

    Ekariansyah, Andi Sofrany

    2002-01-01

    This paper will explain the application of Fault Three Method for analyzing the system reliability of Auxiliary Feedwater System with 2 different configurations taken from PWR type nuclear power plant (NPP) in the USA. The first configuration of Braidwood NPP (design A) basically consists of 1 motor driven pump and 1 diesel driven pump. The second configuration of Haddam Neck NPP (Design B) consists of 2 turbine driven pumps. Based on the P and ID and success criteria the fault trees are constructed to estimate the system failure probabilities quantified from software code PIRAS 1.0. The result shows the second configuration (Design B) with 2 turbine driven pumps have the higher failure probability of 1,06 x 10 - 2 compared with design A of 1,09 x 10 - 3 . The modification of both systems are also tried to analyze its effect to the end result. Qualitatively, the common cause failures of 2 turbine driven pumps contribute to the highest risk of system failure probability. Combination with 1 turbine driven pump and 1 motor driven pump or 1 diesel driven pump will increase the system reliability about 80% and 50% without considering if this configuration is possible to realize in a real plant

  12. Application of a power plant simplification methodology: The example of the condensate feedwater system

    International Nuclear Information System (INIS)

    Seong, P.H.; Manno, V.P.; Golay, M.W.

    1988-01-01

    A novel framework for the systematic simplification of power plant design is described with a focus on the application for the optimization of condensate feedwater system (CFWS) design. The evolution of design complexity of CFWS is reviewed with emphasis upon the underlying optimization process. A new evaluation methodology which includes explicit accounting of human as well as mechanical effects upon system availability is described. The unifying figure of merit for an operating system is taken to be net electricity production cost. The evaluation methodology is applied to the comparative analysis of three designs. In the illustrative examples, the results illustrate how inclusion in the evaluation of explicit availability related costs leads to optimal configurations. These are different from those of current system design practices in that thermodynamic efficiency and capital cost optimization are not overemphasized. Rather a more complete set of design-dependent variables is taken into account, and other important variables which remain neglected in current practices are identified. A critique of the new optimization approach and a discussion of future work areas including improved human performance modeling and different optimization constraints are provided. (orig.)

  13. Probabilistic common cause failure modeling for auxiliary feedwater system after the introduction of flood barriers

    International Nuclear Information System (INIS)

    Zheng, Xiaoyu; Yamaguchi, Akira; Takata, Takashi

    2013-01-01

    Causal inference is capable of assessing common cause failure (CCF) events from the viewpoint of causes' risk significance. Authors proposed the alpha decomposition method for probabilistic CCF analysis, in which the classical alpha factor model and causal inference are integrated to conduct a quantitative assessment of causes' CCF risk significance. The alpha decomposition method includes a hybrid Bayesian network for revealing the relationship between component failures and potential causes, and a regression model in which CCF parameters (global alpha factors) are expressed by explanatory variables (causes' occurrence frequencies) and parameters (decomposed alpha factors). This article applies this method and associated databases needed to predict CCF parameters of auxiliary feedwater (AFW) system when defense barriers against internal flood are introduced. There is scarce operation data for functionally modified safety systems and the utilization of generic CCF databases is of unknown uncertainty. The alpha decomposition method has the potential of analyzing the CCF risk of modified AFW system reasonably based on generic CCF databases. Moreover, the sources of uncertainty in parameter estimation can be studied. An example is presented to demonstrate the process of applying Bayesian inference in the alpha decomposition process. The results show that the system-specific posterior distributions for CCF parameters can be predicted. (author)

  14. Parallel island genetic algorithm applied to a nuclear power plant auxiliary feedwater system surveillance tests policy optimization

    International Nuclear Information System (INIS)

    Pereira, Claudio M.N.A.; Lapa, Celso M.F.

    2003-01-01

    In this work, we focus the application of an Island Genetic Algorithm (IGA), a coarse-grained parallel genetic algorithm (PGA) model, to a Nuclear Power Plant (NPP) Auxiliary Feedwater System (AFWS) surveillance tests policy optimization. Here, the main objective is to outline, by means of comparisons, the advantages of the IGA over the simple (non-parallel) genetic algorithm (GA), which has been successfully applied in the solution of such kind of problem. The goal of the optimization is to maximize the system's average availability for a given period of time, considering realistic features such as: i) aging effects on standby components during the tests; ii) revealing failures in the tests implies on corrective maintenance, increasing outage times; iii) components have distinct test parameters (outage time, aging factors, etc.) and iv) tests are not necessarily periodic. In our experiments, which were made in a cluster comprised by 8 1-GHz personal computers, we could clearly observe gains not only in the computational time, which reduced linearly with the number of computers, but in the optimization outcome

  15. Auxiliary feedwater system risk-based inspection guide for the North Anna nuclear power plants

    International Nuclear Information System (INIS)

    Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

    1992-10-01

    In a study sponsored by the US Nuclear regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. North Anna was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the North Anna plant

  16. Auxiliary feedwater system risk-based inspection guide for the Palo Verde Nuclear Power Plant

    International Nuclear Information System (INIS)

    Bumgardner, J.D.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Sloan, J.A.

    1993-02-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Palo Verde was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Palo Verde plants

  17. Auxiliary feedwater system risk-based inspection guide for the McGuire nuclear power plant

    International Nuclear Information System (INIS)

    Bumgardner, J.D.; Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

    1994-05-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. McGuire was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the McGuire plant

  18. Auxiliary feedwater system risk-based inspection guide for the South Texas Project nuclear power plant

    International Nuclear Information System (INIS)

    Bumgardner, J.D.; Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

    1993-12-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. South Texas Project was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the South Texas Project plant

  19. Auxiliary feedwater system risk-based inspection guide for the Byron and Braidwood nuclear power plants

    International Nuclear Information System (INIS)

    Moffitt, N.E.; Gore, B.F.; Vo, T.V.

    1991-07-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Byron and Braidwood were selected for the fourth study in this program. The produce of this effort is a prioritized listing of AFW failures which have occurred at the plants and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Byron/Braidwood plants. 23 refs., 1 fig., 1 tab

  20. Auxiliary feedwater system risk-based inspection guide for the H. B. Robinson nuclear power plant

    International Nuclear Information System (INIS)

    Moffitt, N.E.; Lloyd, R.C.; Gore, B.F.; Vo, T.V.; Garner, L.W.

    1993-08-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. H. B. Robinson was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the H. B. Robinson plant

  1. Auxiliary feedwater system risk-based inspection guide for the Point Beach nuclear power plant

    International Nuclear Information System (INIS)

    Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Vehec, T.A.

    1993-02-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Point Beach was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRS. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Point Beach plant

  2. Auxiliary feedwater system risk-based inspection guide for the Ginna Nuclear Power Plant

    International Nuclear Information System (INIS)

    Pugh, R.; Gore, B.F.; Vo, T.V.; Moffitt, N.E.

    1991-09-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Ginna was selected as the eighth plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Ginna plant. 23 refs., 1 fig., 1 tab

  3. The impact of feedwater and condensate return excursions on boiler system component failures

    Energy Technology Data Exchange (ETDEWEB)

    Esmacher, Mel J. [GE Water and Process Technologies, The Woodlands, TX (United States); Rossi, Anthony [GE Water and Process Technologies, Trevose, PA (United States)

    2010-02-15

    During boiler operation, the transport of contaminants in boiler feedwater or condensate return via hardness excursions or transport of metal oxides due to corrosion can cause fouling and subsequent tube failure due to under-deposit corrosion or overheating. Case histories are reviewed and suitable corrective actions discussed. (orig.)

  4. Heat structure coupling of CUPID and MARS for the multi-scale simulation of the passive auxiliary feedwater system

    International Nuclear Information System (INIS)

    Kyu Cho, Hyoung; Cho, Yun Je; Yoon, Han Young

    2014-01-01

    Graphical abstract: - Highlights: • PAFS is designed to replace a conventional active auxiliary feedwater system. • Multi-D T/H analysis code, CUPID was coupled with the 1-D system analysis code MARS. • The coupled CUPID and MARS was applied for the multi-scale analysis of the PAFS test facility. • The simulation result showed that the coupled code can reproduce important phenomena in PAFS. - Abstract: For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been developed. In the present study, the CUPID code was coupled with a system analysis code MARS in order to apply it for the multi-scale thermal-hydraulic analysis of the passive auxiliary feedwater system (PAFS). The PAFS is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+), which is intended to completely replace the conventional active auxiliary feedwater system. For verification of the coupling and validation of the coupled code, the PASCAL test facility was simulated, which was constructed with an aim of validating the cooling and operational performance of the PAFS. The two-phase flow phenomena of the steam supply system including the condensation inside the heat exchanger tube were calculated by MARS while the natural circulation and the boil-off in the large water pool that contains the heat exchanger tube were simulated by CUPID. This paper presents the description of the PASCAL facility, the coupling method and the simulation results using the coupled code

  5. Pilot RCM application to the Diablo Canyon main stream system

    International Nuclear Information System (INIS)

    Groff, C.R.; Beckham, P.E.; Bych, K.H.

    1988-01-01

    In 1986 Pacific Gas ampersand Electric Company (PG ampersand E) became extremely interested in reliability-centered maintenance (RCM) after the initial review of two successful Electric Power Research Institute sponsored projects. RCM was visualized as a methodology to common sensitize the burgeoning preventive maintenance (PM) program at the Diablo Canyon plant. RCM could further the uses of predictive and condition-monitoring techniques, as well as eliminate maintenance on components whose failures were noncritical. An extensive review of maintenance and operation experience data, in conjunction with plant staff recommendations and a prioritization according to maintenance expenditures and operational/safety significance, produced the selected system: the turbine main steam supply system (main steam). The pilot project segmented the main steam system into eight subsystems to aid in analysis: (a) main steam isolation valves, (b) auxiliary feedwater pump turbine, (c) overpressure protection (steam dump), (d) main feedwater pump turbines, (e) main steam, (f) main turbine, (g) steam blowdown, and (h) moisture separator reheaters. System analysis activities, including the preparation of functional failure analyses, failure modes and effects analyses, and logic model analyses, were conducted in parallel with corrective and preventive maintenance data-gathering activities to maximize project team personnel participation during the project. Results and lessons learned are summarized

  6. Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air.

  7. Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

    International Nuclear Information System (INIS)

    Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae

    2016-01-01

    The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air

  8. Modeling and simulation of the feedwater system, associated controller and interface with the user for the SUN-RAH nucleo electric plants university student simulator

    International Nuclear Information System (INIS)

    Sanchez B, A.

    2003-01-01

    The simulation process of the component systems of the feedwater of a nucleo electric plant is presented, using several models of reduced order that represent the diverse elements that compose the systems like: the heaters of feedwater, the condenser, the feedwater pump, etc. The integration of the same ones in one simulative structure, and the development of a platform that to give the appearance of to be executed in continuous time, it is the objective of the feedwater simulator, as well as of the SUN-RAH simulator, of which is part. The simulator uses models of reduced order that respond to the observed behavior of a nuclear plant of BWR type. Likewise, it is presented a model of a flow controller of feedwater that will be the one in charge of regulating the demand of the system according to the characteristics and criticize restrictions of safety and controllability, assigned according to those wanted parameters of performance of this system inside the nucleo electric plant. The integration of these models, the adaptation of the variables and parameters, are presented in a way that the integration with the other ones models of the remaining systems of the plant (reactor, steam lines, turbine, etc.), be direct and coherent with the principles of thermodynamic cycles relative to this type of generation plants. The design of those graphic interfaces and the environment where the simulator works its are part of those developments of this work. The reaches and objectives of the simulator complement the description of the simulator. (Author)

  9. Control systems for the dissolved oxygen concentration in condensate- and feed-water systems in nuclear power plants

    International Nuclear Information System (INIS)

    Mikajiri, Motohiko; Hosaka, Seiichi.

    1981-01-01

    Purpose: To surely prevent the generation of corrosion products and contaminations in the systems thereby decreasing the exposure dose to operators in BWR type nuclear power plants. Constitution: Dissolved oxygen concentration in condensates is measured by a dissolved oxygen concentration meter disposed to the pipeway down stream of the condensator and the measured value is sent to an injection amount control mechanism for heater drain water. The control mechanism controls the injection amount from the injection mechanism that injection heater drain water from a feed-water heater to the liquid phase in the hot wall of the condensator. Thus, heater drawin water at high dissolved oxygen is injected to the condensates in the condensator which is de-airated and reduced with dissolved oxygen concentration, to maintain the dissolved oxygen concentration at a predetermined level, whereby stable oxide films are formed to the inner surface of the pipeways to prevent the generation of corrosion products such as rusts. (Furukawa, Y.)

  10. Analogue to digital upgrade project-boiler feedwater control system for Bruce Power nuclear units 1 & 2

    International Nuclear Information System (INIS)

    Long, R.

    2012-01-01

    Bruce Power Nuclear Generating Station A, “Bruce A” is in the final stages of its Restart Project. This capital project will see a large scale rehabilitation of Units 1 and 2 resulting in addition of 1500MW of safe, reliable, clean electricity to the Ontario grid. Restart Project Scope 375, Boiler Feedwater Controls Upgrade was sanctioned to replace obsolete analog devices with a modern digital control system. This project replaced the existing Foxboro H Line analog controls which comprised of 81 individual control modules and support instrumentation. The replacement system was a Triconex Triple Modular Redundant PLC which interfaces with two redundant touch screen monitors. The upgraded digital system incorporates the following controls: 1. Boiler Level Control Loops 2. Dearator Level Control Loops 3. Dearator Pressure Control Loops 4. Boiler Feedwater Recirculation Flow Control Loops A number of technical challenges were addressed when installing a new digital system within the existing plant configuration. Interfaces to new, old and refurbished field devices must be understood as well as implications of connecting to the plant’s Digital Control Computers (DCC’s) and newly installed Steam Generators. The overall project involved many stakeholders to address various requirements from conceptual / design stage through procurement, construction, commissioning and return to service. In addition, the project highlighted the unique requirements found in Nuclear Industry with respect to Human Factors and Software Quality Assurance. (author)

  11. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  12. Study by the disco method of critical components of a P.W.R. normal feedwater system

    International Nuclear Information System (INIS)

    Duchemin, B.; Villeneuve, M.J. de; Vallette, F.; Bruna, J.G.

    1983-03-01

    The DISCO (Determination of Importance Sensitivity of COmponents) method objectif is to rank the components of a system in order to obtain the most important ones versus availability. This method uses the fault tree description of the system and the cut set technique. It ranks the components by ordering the importances attributed to each one. The DISCO method was applied to the study of the 900 MWe P.W.R. normal feedwater system with insufficient flow in steam generator. In order to take account of operating experience several data banks were used and the results compared. This study allowed to determine the most critical component (the turbo-pumps) and to propose and quantify modifications of the system in order to improve its availability

  13. A niching genetic algorithm applied to a nuclear power plant auxiliary feedwater system surveillance tests policy optimization

    International Nuclear Information System (INIS)

    Sacco, W.F.; Lapa, Celso M.F.; Pereira, C.M.N.A.; Oliveira, C.R.E. de

    2006-01-01

    This article extends previous efforts on genetic algorithms (GAs) applied to a nuclear power plant (NPP) auxiliary feedwater system (AFWS) surveillance tests policy optimization. We introduce the application of a niching genetic algorithm (NGA) to this problem and compare its performance to previous results. The NGA maintains a populational diversity during the search process, thus promoting a greater exploration of the search space. The optimization problem consists in maximizing the system's average availability for a given period of time, considering realistic features such as: (i) aging effects on standby components during the tests; (ii) revealing failures in the tests implies on corrective maintenance, increasing outage times; (iii) components have distinct test parameters (outage time, aging factors, etc.) and (iv) tests are not necessarily periodic. We find that the NGA performs better than the conventional GA and the island GA due to a greater exploration of the search space

  14. Resolution of concerns in auxiliary feedwater piping

    International Nuclear Information System (INIS)

    Bain, R.A.; Testa, M.F.

    1994-01-01

    Auxiliary feedwater piping systems at pressurized water reactor (PWR) nuclear power plants have experienced unanticipated operating conditions during plant operation. These unanticipated conditions have included plant events involving backleakage through check valves, temperatures in portions of the auxiliary feedwater piping system that exceed design conditions, and the occurrence of unanticipated severe fluid transients. The impact of these events has had an adverse effect at some nuclear stations on plant operation, installed plant components and hardware, and design basis calculations. Beaver Valley Unit 2, a three loop pressurized water reactor nuclear plant, has observed anomalies with the auxiliary feedwater system since the unit went operational in 1987. The consequences of these anomalies and plant events have been addressed and resolved for Beaver Valley Unit 2 by performing engineering and construction activities. These activities included pipe stress, pipe support and pipe rupture analysis, the monitoring of auxiliary feedwater system temperature and pressure, and the modification to plant piping, supports, valves, structures and operating procedures

  15. A reliability centered maintenance model applied to the auxiliary feedwater system of a nuclear power plant; Um modelo de manutencao centrada em confiabilidade aplicada ao sistema de agua de alimentacaco auxiliar de uma usina nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Jefferson Borges

    1998-01-15

    The main objective of maintenance in a nuclear power plant is to assure that structures, systems and components will perform their design functions with reliability and availability in order to obtain a safety and economic electric power generation. Reliability Centered Maintenance (RCM) is a method of systematic review to develop or optimize Preventive Maintenance Programs. This study presents the objectives, concepts, organization and methods used in the development of RCM application to nuclear power plants. Some examples of this application are included, considering the Auxiliary Feedwater System of a generic two loops PWR nuclear power plant of Westinghouse design. (author)

  16. LQG/LTR [linear quadratic Gaussian with loop transfer recovery] robust control system design for a low-pressure feedwater heater train

    International Nuclear Information System (INIS)

    Murphy, G.V.; Bailey, J.M.

    1990-01-01

    This paper uses the linear quadratic Gaussian with loop transfer recovery (LQG/LTR) control system design method to obtain a level control system for a low-pressure feedwater heater train. The control system performance and stability robustness are evaluated for a given set of system design specifications. The tools for analysis are the return ratio, return difference, and inverse return difference singular-valve plots for a loop break at the plant output. 3 refs., 7 figs., 2 tabs

  17. Vulnerability of steam generator super-emergency feeding. Super-emergency feedwater system for the Mochovce NPP steam-generators

    International Nuclear Information System (INIS)

    Hlasova, M.; Jary, A.

    1997-01-01

    The following major requirements and criteria fulfillment concerned the super-emergency feedwater system (SEFW) system were proposed: to provide sufficient water amount for accident conditions, inclusive seismicity, even during required SEFW system operation for the time period of 72 hours; to analyse ensuring of residual heat removal in case of a station black-out; to state criteria for water supply by the SEFW system into the steam generators (SGs); to simplify the existing connection scheme inclusive decreasing the number of valves, which are in series; to analyse and provide the system protection against a common cause failure, which the SEFW system did not provide in some parts (possibilities of three systems failure due to flooding; vulnerability of all tanks by the operation building fall in case of a seismic event; vulnerability of all tanks due to extreme climatic conditions; vulnerability of all tanks during new seismic loading and consequent mutual endangering; the possibility of three systems failure due to common routing in the vicinity of high; energy media on the +14,7 m floor in the intermediate machinery building and due to inconsistent electrical valves secured power supply systems); to analyse temperature increase impact on the number of uses and lifetime of SGs; to perform a change of SEFW system pipelines routing layout outside the dangerous area of the +14,7 m floor in the intermediate machinery building with high energy media; checking the thanks autonomy. There were performed analyses of selected transient operation modes. The analyses had the following objectives: necessary flowrate of the SEFW in case of the primary side stabilised temperature of 140 C till 72 hours of the process duration; sufficient capacity of one subsystem for the supply of sufficient water amount; sufficient water reserve in the tanks at given conditions; and other. Accident situations were evaluated using an analysis and three characteristic operation modes were

  18. Reliability study of the auxiliary feed-water system of a pressurized water reactor by faults tree and Bayesian Network

    International Nuclear Information System (INIS)

    Lava, Deise Diana; Borges, Diogo da Silva; Guimarães, Antonio Cesar Ferreira; Moreira, Maria de Lourdes

    2017-01-01

    This paper aims to present a study of the reliability of the Auxiliary Feed-water System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10 -3 . (author)

  19. Reliability study of the auxiliary feed-water system of a pressurized water reactor by faults tree and Bayesian Network

    Energy Technology Data Exchange (ETDEWEB)

    Lava, Deise Diana; Borges, Diogo da Silva; Guimarães, Antonio Cesar Ferreira; Moreira, Maria de Lourdes, E-mail: deise_dy@hotmail.com, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    This paper aims to present a study of the reliability of the Auxiliary Feed-water System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10{sup -3}. (author)

  20. Digital feedwater and recirculation flow control for GPUN Oyster Creek

    International Nuclear Information System (INIS)

    Burjorjee, D.; Gan, B.

    1992-01-01

    This paper describes the digital system for feedwater and recirculation control that GPU Nuclear will be installing at Oyster Creek during its next outage - expected circa December 1992. The replacement was motivated by considerations of reliability and obsolescence - the analog equipment was aging and reaching the end of its useful life. The new system uses Atomic Energy of Canada Ltd.'s software platform running on dual, redundant, industrial-grade 386 computers with opto-isolated field input/output (I/O) accessed through a parallel bus. The feedwater controller controls three main feed regulating valves, two low flow regulating valves, and two block valves. The recirculation controller drives the five scoop positioners of the hydraulic couplers. The system also drives contacts that lock up the actuators on detecting an open circuit in their current loops

  1. Device for achieving pressure balance in the steam generator of a power plant in case of a main-steam pipe or a feedwater pipe break

    International Nuclear Information System (INIS)

    Wietelmann, F.

    1978-01-01

    In order to increase the safety in the steam generator of a power plant in case of a pipe break, the possibility of a pressure balance between the feedwater inlet and the initial steam outlet chambers is allowed for. According to the invention, the partition wall separating these two chambers will exhibit several overflow openings, each of which will be provided with a closure and half of which may be opened to one side only, care having been taken that in case of an accident on occurrence of a certain differential pressure they will always be opened to the low-pressure side. As closures caps, which may be swing out of the way, or rupture diaphragms are mentioned. (UWI) 891 HP [de

  2. A Smart Soft Sensor Predicting Feedwater Flow Rate

    International Nuclear Information System (INIS)

    Yang, Heon Young; Na, Man Gyun

    2009-01-01

    Since we evaluate thermal nuclear reactor power with secondary system calorimetric calculations based on feedwater flow rate measurements, we need to measure the feedwater flow rate accurately. The Venturi flow meters that are being used to measure the feedwater flow rate in most pressurized water reactors (PWRs) measure the flow rate by developing a differential pressure across a physical flow restriction. The differential pressure is then multiplied by a calibration factor that depends on various flow conditions in order to calculate the feedwater flow rate. The calibration factor is determined by the feedwater temperature and pressure. However, Venturi meters cause a buildup of corrosion products near the orifice of the meter. This fouling increases the measured pressure drop across the meter, thereby causing an overestimation of the feedwater flow rate

  3. The analysis of the functional role of man and machine in the control of a notional auxiliary feedwater system

    International Nuclear Information System (INIS)

    Cacciabue, P.C.; Codazzi, A.; Decortis, F.

    1991-01-01

    We will describe here the simulation of a moderately complex plant, i.e. the Auxiliary Feedwater System (AFWS) of a nuclear power plant, which has been developed for interacting with a cognitive model of operator in a simulation framework of man-machine system studies as well as with an external operator for verifying and validating the hypotheses of the theoretical model by experimental studies. In order to develop such simulation, which must be very flexible for satisfying the needs of interaction with an operator as well as with a cognitive model, a number of special conditions have been respected: the model of functional behaviour of the system has been extended to include the logic of control mechanisms, i.e. components, indicators and actuators; the control tasks for a number of sequences has been developed; the robustness of physical model has been tested in whole possible configuration of the plant; and finally, the interface of the simulation with the model for dynamic failures of components has also been granted. In this paper, these aspects of the deterministic model of the AFWS will be firstly presented in detail. Then, the interface of the plant simulation with an external user or with the cognitive model of the operator will be described focusing on the analysis of the control task. Finally, we will attempt to integrate our approach in an overall framework of taxonomy for studying human actions in complex work context

  4. Auxiliary feedwater system risk-based inspection guide for the Beaver Valley, Units 1 and 2 nuclear power plants

    International Nuclear Information System (INIS)

    Lloyd, R.C.; Vehec, T.A.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Rossbach, L.W.; Sena, P.P. III

    1993-02-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Beaver Valley Units 1 and 2 were selected as two of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at Beaver Valley Units 1 and 2

  5. Auxiliary feedwater system risk-based inspection guide for the J.M. Farley Nuclear Power Plant

    International Nuclear Information System (INIS)

    Vo, T.V.; Pugh, R.; Gore, B.F.; Harrison, D.G.

    1990-10-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment(PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. J. M. Farley was selected as the second plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important at the J. M. Farley plant. 23 refs., 1 fig., 1 tab

  6. Feedwater device for nuclear power plant

    International Nuclear Information System (INIS)

    Ikekita, Iwao.

    1980-01-01

    Purpose: To conduct water feeding without using high pressure steam of the reactor and with no radiation exposure by the provision of each feedwater pump driven by each motor controlled from variable frequency thyristor-inverter to a feedwater pipe connecting a condensate pump and the reactor. Constitution: High pressure steams resulted from heat exchange in the reactor core are transferred by way of a main steam check valve in a main steam pipe to a high pressure turbine, drive the high pressure turbine, flow out of the turbine and then drive a low pressure turbine by way of a moisture separator. The steams thus used for the turbine driving are condensed in a condensator and then sent under pressure by way of each condensating pump to a feedwater pipe. Since each of the feedwater pumps provided in the route of the feedwater pipe is driven by each of the motors under the control of the variable frequency thyristor-inverter in starting, shut down and normal operation, water is fed to the reactor. (Horiuchi, T.)

  7. ESBWR power maneuvering via feedwater temperature control

    International Nuclear Information System (INIS)

    Saha, P.; Marquino, W.; Tucker, L. J.

    2008-01-01

    The ESBWR is a Generation III+ Boiling Water Reactor (BWR) driven by natural circulation. For a given geometry/hardware, system pressure, downcomer water level and feedwater temperature, the core flow rate in the ESBWR is only a function of reactor power, controlled through the control blade movement. In order to provide operational flexibility, another method of core-wide or global power maneuvering via feedwater temperature control has been developed. This is independent of power maneuvering via control blade movement, and it lowers the linear heat generation rate (LHGR) changes near the tip of control blades, which improves fuel reliability. All required stability, anticipated operational occurrences (AOOs), infrequent events, special events including anticipated transients without scram (ATWS), and loss-of-coolant accident (LOCA) analyses have been performed for the 4500 MWt ESBWR. Based on the results of these analyses at 'high', nominal and 'low' feedwater temperatures, a safe Power - Feedwater Temperature operating domain has been developed. This paper summarizes the results of these analyses and presents the ESBWR Power - Feedwater Temperature operating domain or map. (authors)

  8. Auxiliary feedwater system risk-based inspection guide for the Diablo Canyon Unit 1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Gore, B.F.; Vo, T.V.; Harrison, D.G.

    1990-08-01

    This document presents a compilation of auxiliary feedwater (AFW) system failure information which has been screened for risk significance in terms of failure frequency and degradation of system performance. It is a risk-prioritized listing of failure events and their causes that are significant enough to warrant consideration in inspection planning at Diablo Canyon. This information is presented to provide inspectors with increased resources for inspection planning at Diablo Canyon. The risk importance of various component failure modes was identified by analysis of the results of probabilistic risk assessments (PRAs) for many pressurized water reactors (PWRs). However, the component failure categories identified in PRAs are rather broad, because the failure data used in the PRAs is an aggregate of many individual failures having a variety of root causes. In order to help inspectors to focus on specific aspects of component operation, maintenance and design which might cause these failures, an extensive review of component failure information was performed to identify and rank the root causes of these component failures. Both Diablo Canyon and industry-wide failure information was analyzed. Failure causes were sorted on the basis of frequency of occurrence and seriousness of consequence, and categorized as common cause failures, human errors, design problems, or component failures. This information permits an inspector to concentrate on components important to the prevention of core damage. Other components which perform essential functions, but which are not included because of high reliability or redundancy, must also be addressed to ensure that degradation does not increase their failure probabilities, and hence their risk importances. 23 refs., 1 fig., 1 tab

  9. Automatic system for redistributing feedwater in a steam generator of a nuclear power plant

    International Nuclear Information System (INIS)

    Fuoto, J.S.; Crotzer, M.E.; Lang, G.E.

    1980-01-01

    A system is described for automatically redistributing a steam generator secondary tube system after a burst in the secondary tubing. This applies to a given steam generator in a system having several steam generators partially sharing a common tube system, and employs a pressure control generating an electrical signal which is compared with given values [fr

  10. Replacement of the feedwater pipe system in reactor building outside containment at the nuclear power plant Philippsburg; Austausch der Speisewasserleitung im Reaktorgebaeude ausserhalb SHB im Kernkraftwerk Philippsburg I

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, A. [Energie-Versorgung Schwaben AG, Stuttgart (Germany); Labes, M. [Siemens AG Unternehmensbereich KWU, Offenbach am Main (Germany); Schwenk, B. [Kernkraftwerk Philippsburg GmbH (Germany)

    1998-11-01

    After full replacement of the feedwater pipe system during the inspection period in 1997, combined with a modern materials, manufacturing and analysis concept, the entire pipe system of the water/steam cycle in the reactor building of KKP 1 now consists of high-toughness materials. The safety level of the entire plant has been increased by leaving aside postulation of F2 breaks in the reactor building and providing for protection against 0.1 leaks. Based on fluid-dynamic calculations for the cases of pump failure and pipe break, as well as pipe system calculations in 5 extensive calculation cycles, about 130 documents were filed for inspection and approval (excluding preliminary test documents on restraints). Points of main interest for safety analysis in this context were the optimised closing performance of the 3rd check valves and the integrity of the nozzle region at the RPV. (oirg./CB) [Deutsch] Durch den Restaustausch der Speisewasserleitungen in der Revision 1997, verbunden mit einem modernen Werkstoff-, Fertigungs- und Nachweiskonzept, sind im Reaktorgebaeude von KKP 1 in den Hauptleitungen des Wasser-Dampf-Kreislaufes nur noch hochzaehe Werkstoffe eingesetzt. Durch den Verzicht auf das Postulat von 2F-Bruechen im Reaktorgebaeude und durch die Auslegung gegen 0,1F-Lecks wird das Sicherheitsniveau der Anlage insgesamt gesteigert. Ausgehend von fluiddynamischen Berechnungen fuer Pumpenausfall und Rohrbruch sowie Rohrsystem-Berechnungen in 5 umfangreichen Berechnungskreisen wurden fuer die Genehmigung und Begutachtung ca. 130 Unterlagen (ohne Halterungs-Vorpruefunterlagen) eingereicht und vom Gutachter geprueft. Schwerpunkte der Nachweisfuehrung waren die Optimierung des Schliessverhaltens der 3. Rueckschlagarmaturen sowie der Integritaetsnachweis des RDB-Anschlusses. (orig./MM)

  11. Application of fuzzy logic control system for reactor feed-water control

    International Nuclear Information System (INIS)

    Iijima, T.; Nakajima, Y.

    1994-01-01

    The successful actual application of a fuzzy logic control system to the a nuclear Fugen nuclear power reactor is described. Fugen is a heavy-water moderated, light-water cooled reactor. The introduction of fuzzy logic control system has enabled operators to control the steam drum water level more effectively in comparison to a conventional proportional-integral (PI) control system

  12. New Main Ring control system

    International Nuclear Information System (INIS)

    Seino, K.; Anderson, L.; Ducar, R.; Franck, A.; Gomilar, J.; Hendricks, B.; Smedinghoff, J.

    1990-03-01

    The Fermilab Main Ring control system has been operational for over sixteen years. Aging and obsolescence of the equipment make the maintenance difficult. Since the advent of the Tevatron, considerable upgrades have been made to the controls of all the Fermilab accelerators except the Main Ring. Modernization of the equipment and standardization of the hardware and software have thus become inevitable. The Tevatron CAMAC serial system has been chosen as a basic foundation in order to make the Main Ring control system compatible with the rest of the accelerator complex. New hardware pieces including intelligent CAMAC modules have been designed to satisfy unique requirements. Fiber optic cable and repeaters have been installed in order to accommodate new channel requirements onto the already saturated communication medium system. 8 refs., 2 figs

  13. Cause analysis for elbow thinning of the secondary loop feedwater system in PWR NPP

    International Nuclear Information System (INIS)

    Yu Tao; Bian Chunhua; Zhang Wei; Luo Kunjie; Wang Li; Li Yan

    2013-01-01

    Wall thickness of some secondary system pipelines were measured on site during the refueling outages in March 2012. Wall thinning happened in some components of APA system. This paper focused on the cause analysis of an elbow with that phenomenon. Wall thickness was carefully measured in laboratory using ultrasonic thickness meter and found that wall thinning happened nearly all elbows including two abnormal thinning regions. Analytical research was conducted using ICP, stereo microscope, SEM, XRD, ANSYS. The result shows that the cause of wall thinning is flow accelerated corrosion. Based on the analysis result and international research progress, this paper makes some suggestions to avoid and alleviate FAC in the secondary system. (authors)

  14. Fluid transient analysis and design considerations in TVA PWR feedwater systems and steam generators

    International Nuclear Information System (INIS)

    Kelley, B.T.

    1979-01-01

    TVA has evaluated a number of fluid transients in an effort to discover areas of potential problems and to improve overall unit operation. The transients recently or currently being evaluated fall into four major areas - accident analyses, fast valving, heater drain systems, and steam generators. A discussion of each area follows

  15. Assessment of Flow Instability in Passive Auxiliary Feedwater System (PAFS) Using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech., Yongin (Korea, Republic of); Cheon, Jong; Kim, Han-Gon [KHNP, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the occurrence possibility of both instabilities in PAFS is assessed with the best-estimate thermal hydraulic code, RELAP5. From the RELAP5 code analysis, the Ledinegg instability might not occur in PAFS. The DWO might occur in PAFS but the effect of the oscillation on the heat removal capacity of PAFS was not large. Therefore, it is concluded that PAFS is safe in terms of flow instabilities. Since PAFS is two-phase flow system, flow instabilities may occur. Flow instabilities may cause the severe deterioration of heat removal capability of PAFS due to the reduction of the condensate flow. For the reliable operation of PAFS, it is required to assess the flow instabilities in PAFS. The Ledinegg-type instability and the Density Wave Oscillation (DWO) are the representative static flow instability and the dynamic flow instability, respectively.

  16. Reactor feedwater system

    International Nuclear Information System (INIS)

    Hikabe, Katsumi.

    1978-01-01

    Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

  17. Modernization of the feedwater heaters control level of the Almaraz I Nuclear Power Plant by OVATION system; Modernizacion del control de nivel de los calentadores de agua de alimentacion de C.N. Almaraz I mediante el sistema OVATION

    Energy Technology Data Exchange (ETDEWEB)

    Madronal Rodriguez, E.; Cabrero Munoz, J. E.

    2010-07-01

    As a result of the process of technological renovation of the heaters system and the power increase project, Almaraz Nuclear Power Plant has made several design changes in the feedwater heaters system. Within these changes, the old heaters control loops are replaced because the new power will increase the heaters drainage caudal. This modernization is carried out using the OVATION control system.

  18. Collector feedwater supply and stability of the power distribution in a pressurized-water reactor

    International Nuclear Information System (INIS)

    Budnikov, V.I.; Kosolapov, S.V.; Kramerov, A.Ya.

    1980-01-01

    It is necessary to determine how the collector feedwater supply affects the disposition of the stability limits and the instability period for the power distribution in such a reactor. The main reason for the fluctuations in feedwater flow rate were shown by additional calculations with the general power regulator switched out to be due to instability on the fundamental in the neutron distribution. The power-level fluctuations are due to oscillation of the feed valve in the level regulator, and consequently to oscillations in the feedwater flow rate. If collector feed is to be employed, it is desirable to improve the response of the pressure control system for the separator drum, because under certain emergency conditions there will be a considerable fall in pressure in the separator drum. The deviation from saturation for the water in the separator drum tube is less in the second method than it is in the first, so the cavitation margin in the principal pumps may be reduced somewhat. Calculations show that this reduction will not occur if the time constant of the turbine synchronizer is about 10 sec. Also, the dynamic characteristics of the nuclear power station in these modes of feedwater supply are appreciably influenced by the parameters of the pressure-control system and the water-level control for the separator drum

  19. Aiding operator performance at low power feedwater control

    International Nuclear Information System (INIS)

    Woods, D.D.

    1986-01-01

    Control of the feedwater system during low power operations (approximately 2% to 30% power) is a difficult task where poor performance (excessive trips) has a high cost to utilities. This paper describes several efforts in the human factors aspects of this task that are underway to improve feedwater control. A variety of knowledge acquisition techniques have been used to understand the details of what makes feedwater control at low power difficult and what knowledge and skill distinguishes expert operators at this task from less experienced ones. The results indicate that there are multiple factors that contribute to task difficulty

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

    International Nuclear Information System (INIS)

    Beroux, M.

    1985-01-01

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

  1. Main injector synchronous timing system

    International Nuclear Information System (INIS)

    Blokland, W.; Steimel, J.

    1998-01-01

    The Synchronous Timing System is designed to provide sub-nanosecond timing to instrumentation during the acceleration of particles in the Main Injector. Increased energy of the beam particles leads to a small but significant increase in speed, reducing the time it takes to complete a full turn of the ring by 61 nanoseconds (or more than 3 rf buckets). In contrast, the reference signal, used to trigger instrumentation and transmitted over a cable, has a constant group delay. This difference leads to a phase slip during the ramp and prevents instrumentation such as dampers from properly operating without additional measures. The Synchronous Timing System corrects for this phase slip as well as signal propagation time changes due to temperature variations. A module at the LLRF system uses a 1.2 Gbit/s G-Link chip to transmit the rf clock and digital data (e.g. the current frequency) over a single mode fiber around the ring. Fiber optic couplers at service buildings split off part of this signal for a local module which reconstructs a synchronous beam reference signal. This paper describes the background, design and expected performance of the Synchronous Timing System. copyright 1998 American Institute of Physics

  2. Main injector synchronous timing system

    International Nuclear Information System (INIS)

    Blokland, Willem; Steimel, James

    1998-01-01

    The Synchronous Timing System is designed to provide sub-nanosecond timing to instrumentation during the acceleration of particles in the Main Injector. Increased energy of the beam particles leads to a small but significant increase in speed, reducing the time it takes to complete a full turn of the ring by 61 nanoseconds (or more than 3 rf buckets). In contrast, the reference signal, used to trigger instrumentation and transmitted over a cable, has a constant group delay. This difference leads to a phase slip during the ramp and prevents instrumentation such as dampers from properly operating without additional measures. The Synchronous Timing System corrects for this phase slip as well as signal propagation time changes due to temperature variations. A module at the LLRF system uses a 1.2 Gbit/s G-Link chip to transmit the rf clock and digital data (e.g. the current frequency) over a single mode fiber around the ring. Fiber optic couplers at service buildings split off part of this signal for a local module which reconstructs a synchronous beam reference signal. This paper describes the background, design and expected performance of the Synchronous Timing System

  3. Steam generation: fossil-fired systems: utility boilers; industrial boilers; boiler auxillaries; nuclear systems: boiling water; pressurized water; in-core fuel management; steam-cycle systems: condensate/feedwater; circulating water; water treatment

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    A survey of development in steam generation is presented. First, fossil-fired systems are described. Progress in the design of utility and industrial boilers as well as in boiler auxiliaries is traced. Improvements in coal pulverizers, burners that cut pollution and improve efficiency, fans, air heaters and economisers are noted. Nuclear systems are then described, including the BWR and PWR reactors, in-core fuel management techniques are described. Finally, steam-cycle systems for fossil-fired and nuclear power plants are reviewed. Condensate/feedwater systems, circulating water systems, cooling towers, and water treatment systems are discussed

  4. Loss-of-feedwater transients in PWRs

    International Nuclear Information System (INIS)

    Burns, R.D. III.

    1980-01-01

    Recent severe accident sequence analysis (SASA) work in LASL's Multifault Accident Analysis Section has focused on loss-of-feedwater (LOFW) transients at a 4-loop Westinghouse nuclear power reactor. In all transients studied, the initiator was loss of main feedwater and reactor coolant pump (RCP) trip, caused by temporary loss of off-site power. Subsequent automatic actions included reactor scram, closure of the main steam isolation valves, and initiation of auxiliary feedwater (AFW) flow. TRAC-PD2 calculations were designed to study the consequences of AFW delivery rates below the minimum specified in the emergency operating procedures (EOPs) for the reference 4-loop plant. Six types of LOFW scenarios have been studied, including (1) zero AFW availability (nominal case), (2) initially zero AFW but full recovery after 2 h, (3) zero AFW with steam generator (SG) atmospheric relief valve (ARV) malfunction, (4) zero AFW with high pressure charging flow initiated after 2 h, and (5) zero AFW with delay in reactor scram. Additional cases were considered to study the effects of uncertainties in pressurizer heater/spray operation, operator manual initiation of high pressure charging flow, reactor initial conditions, and RCP and power coastdown characteristics. Nominal case results, rationale for selections of other cases, and lessons learned are summarized

  5. Minimum throttling feedwater control in VVER-1000 and PWR NPPs

    International Nuclear Information System (INIS)

    Symkin, B.E.; Thaulez, F.

    2004-01-01

    This paper presents an approach for the design and implementation of advanced digital control systems that use a minimum-throttling algorithm for the feedwater control. The minimum-throttling algorithm for the feedwater control, i.e. for the control of steam generators level and of the feedwater pumps speed, is applicable for NPPs with variable speed feedwater pumps. It operates in such a way that the feedwater control valve in the most loaded loop is wide open, steam generator level in this loop being controlled by the feedwater pumps speed, while the feedwater control valves in the other loops are slightly throttling under the action of their control system, to accommodate the slight loop imbalances. This has the advantage of minimizing the valve pressure losses hence minimizing the feedwater pumps power consumption and increasing the net MWe. The benefit has been evaluated for specific plants as being roughly 0.7 and 2.4 MW. The minimum throttling mode has the further advantages of lowering the actuator efforts with potential positive impact in actuator life and of minimizing the feedwater pipelines vibrations. The minimum throttling mode of operation has been developed by the Ukrainian company LvivORGRES. It has been applied with great deal of success on several VVER-1000 NPPs, six units of Zaporizhzha in Ukraine plus, with participation of Westinghouse, Kozloduy 5 and 6 in Bulgaria and South Ukraine 1 to 3 in Ukraine. The concept operates with both ON-OFF valves and true control valves. A study, jointly conducted by Westinghouse and LvivORGRES, is ongoing to demonstrate the applicability of the concept to PWRs having variable speed feedwater pumps and having, or installing, digital feedwater control, standalone or as part of a global digital control system. The implementation of the algorithm at VVER-1000 plants provided both safety improvement and direct commercial benefits. The minimum-throttling algorithm will similarly increase the performance of PWRs. The

  6. Lead corrosion and transport in simulated secondary feedwater

    Energy Technology Data Exchange (ETDEWEB)

    McGarvey, G.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Ross, K.J.; McDougall, T.E. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Turner, C.W. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1998-07-01

    The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of,lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is due to low levels of lead although few if any failures have been experimentally linked to lead when sub-parts per billion levels are present in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the three principal corrosion products in the secondary feedwater; magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values less than that of the feedwater (9-10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces following different treatment conditions will be used to propose a model for tile transport and probable fate of lead in the secondary feedwater system. (author)

  7. Lead corrosion and transport in simulated secondary feedwater

    Energy Technology Data Exchange (ETDEWEB)

    McGarvey, G.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Ross, K.J.; McDougall, T.E. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Turner, C.W

    1999-07-01

    The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is caused by low levels of lead although few, if any, failures have been experimentally linked to lead when it is present in sub-parts per billion in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the 3 principal corrosion products in the secondary feedwater: magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values that are lower than the pH of the feedwater (9 to 10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces after different treatment conditions will be used to propose a model for the transport and probable fate of lead in the secondary feedwater system. (author)

  8. Lead corrosion and transport in simulated secondary feedwater

    International Nuclear Information System (INIS)

    McGarvey, G.B.; Ross, K.J.; McDougall, T.E.; Turner, C.W.

    1998-01-01

    The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of,lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is due to low levels of lead although few if any failures have been experimentally linked to lead when sub-parts per billion levels are present in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the three principal corrosion products in the secondary feedwater; magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values less than that of the feedwater (9-10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces following different treatment conditions will be used to propose a model for tile transport and probable fate of lead in the secondary feedwater system. (author)

  9. An effect of downcomer feedwater fraction on steam generator performance with an axial flow economizer

    International Nuclear Information System (INIS)

    Jung, Byung Ryul; Park, Hu Shin; Chung, Duk Muk; Baik, Se Jin

    2000-01-01

    The effects of feedwater flow fraction introduced into the downcomer region have been evaluated in terms of steam generator performance based on the same steam generator thermal output for the Korea Standard Nuclear Power Plant (KSNP) steam generator. The KSNP steam generator design has an integral axial flow economizer which is designed such that most of the feedwater is introduced through the economizer region and only a portion of feedwater through the downcomer region. The feedwater flow introduced into the downcomer region is not normally controlled during the power operation. However, the actual feedwater fraction into the downcomer region may differ from the design flow depending on the as-built system and component characteristics. Investigated in this paper were the downcomer feedwater flow effects on the steam pressure, circulation ratio, internal void fraction and velocity distribution in the tube bundle region at the steady state operation using SAFE and ATHOS3 codes. The results show that the steam pressure increases and the resultant total feedwater flow increases with reducing the downcomer feedwater flow fraction for the same steam generator thermal output. The slight off-design condition of downcomer feedwater flow fraction renders no significant effect on the steam generator performance such as circulation ratios, steam qualities, void fractions and internal velocity distributions. The evaluation shows that the slight off-design downcomer feedwater flow fraction deviation up to ± 5% is acceptable for the steam generator performance

  10. Technical feasibility study of a low-cost hybrid PAC-UF system for wastewater reclamation and reuse: a focus on feedwater production for low-pressure boilers.

    Science.gov (United States)

    Amosa, Mutiu Kolade; Jami, Mohammed Saedi; Alkhatib, Ma'an Fahmi R; Majozi, Thokozani

    2016-11-01

    This study has applied the concept of the hybrid PAC-UF process in the treatment of the final effluent of the palm oil industry for reuse as feedwater for low-pressure boilers. In a bench-scale set-up, a low-cost empty fruit bunch-based powdered activated carbon (PAC) was employed for upstream adsorption of biotreated palm oil mill effluent (BPOME) with the process conditions: 60 g/L dose of PAC, 68 min of mixing time and 200 rpm of mixing speed, to reduce the feedwater strength, alleviate probable fouling of the membranes and thus improve the process flux (productivity). Three polyethersulfone ultrafiltration membranes of molecular weight cut-off (MWCO) of 1, 5 and 10 kDa were investigated in a cross-flow filtration mode, and under constant transmembrane pressures of 40, 80, and 120 kPa. The permeate qualities of the hybrid processes were evaluated, and it was found that the integrated process with the 1 kDa MWCO UF membrane yielded the best water quality that falls within the US EPA reuse standard for boiler-feed and cooling water. It was also observed that the permeate quality is fit for extended reuse as process water in the cement, petroleum and coal industries. In addition, the hybrid system's operation consumed 37.13 Wh m -3 of energy at the highest applied pressure of 120 kPa, which is far lesser than the typical energy requirement range (0.8-1.0 kWh m -3 ) for such wastewater reclamation.

  11. Smart Soft-Sensing for the Feedwater Flowrate at PWRs Using a GMDH Algorithm

    Science.gov (United States)

    Lim, Dong Hyuk; Lee, Sung Han; Na, Man Gyun

    2010-02-01

    The thermal reactor power in pressurized water reactors (PWRs) is typically assessed using secondary system calorimetric calculations based on accurate measurements of the feedwater flowrate. Therefore, precise measurements of the feedwater flowrate are essential. In most PWRs, Venturi meters are used to measure the feedwater flowrate. However, the fouling phenomena of the Venturi meter deteriorate the accuracy of the existing hardware sensors. Consequently, it is essential to resolve the inaccurate measurements of the feedwater flowrate. In this study, in order to estimate the feedwater flowrate online with high precision, a smart soft sensing model for monitoring the feedwater flowrate was developed using a group method of data handling (GMDH) algorithm combined with a sequential probability ratio test (SPRT). The uncertainty of the GMDH model was also analyzed. The proposed sensing and monitoring algorithm was verified using the acquired real plant data from Yonggwang Nuclear Power Plant Unit 3.

  12. Control of feedwater composition of BWR power plant

    International Nuclear Information System (INIS)

    Sturla, P.; D'Anna, A.; Borgese, D.

    1983-01-01

    Corrosion behaviour of fuel element cladding, cycle structural materials and dose rate increase are relevant to physico-chemical characteristics of process coolants and to adopted operational conditions. A careful control of cycle chemistry, during loading and shutdown periods, is necessary to verify material choices, the polishing system and chemistry specifications. For this purpose ENEL carried out some preliminary experimental tests employing continuous control system and samples for specific analytical determinations. The cycle points checked during about two months were: main condensate; condensate after polishing system; outlet of low pressure heathers; final feedwater; inlet and outlet of clean-up system; drains to condenser. The physico-chemical analysis were related to corrosion product levels (Cu, Fe, Ni, Co) and water chemistry (pH, conductivity, dissolved oxygen etc.). The preliminary results allow to express some considerations about sampling procedures, detection limits and reliability of analytical employed methods. The acquisition data time and some morphological oxide pictures are also showed. (author)

  13. Study of the reliability of the Auxiliary Feedwater System of a LWR nuclear power plant through the Fault Tree and Bayesian Network

    International Nuclear Information System (INIS)

    Lava, Deise Diana

    2016-01-01

    This paper aims to present a study of the reliability of the Auxiliary Feedwater System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10 -3 . (author)

  14. Manual for investigation and correction of feedwater heater failures

    International Nuclear Information System (INIS)

    Bell, R.J.; Diaz-Tous, I.A.; Bartz, J.A.

    1993-01-01

    The Electric Power Research Institute (EPRI) has sponsored the development of a recently published manual which is designed to assist utility personnel in identifying and correcting closed feedwater heater problems. The main portion of the manual describes common failure modes, probable means of identifying root causes and appropriate corrective actions. These include materials selection, fabrication practices, design, normal/abnormal operation and maintenance. The manual appendices include various data, intended to aid those involved in monitoring and condition assessment of feedwater heaters. This paper contains a detailed overview of the manual content and suggested means for its efficient use by utility engineers and operations and maintenance personnel who are charged with the responsibilities of performing investigations to identify the root cause(s) of closed feedwater problems/failures and to provide appropriate corrective actions. 4 refs., 3 figs., 2 tabs

  15. An interface redesign for the feed-water system of the advanced boiling water reactor in a nuclear power plant in Taiwan

    International Nuclear Information System (INIS)

    Hsieh Minchih; Chiu Mingchuan; Hwang Sheueling

    2014-01-01

    A well-designed human-computer interface for the visual display unit in the control room of a complex environment can enhance operator efficiency and, thus, environmental safety. In fact, a cognitive gap often exists between an interface designer and an interface user. Therefore, the issue of the cognitive gap of interface design needs more improvement and investigation. This is an empirical study that presents the application of an ecological interface design (EID) using three cases and demonstrates that an EID framework can support operators in various complex situations. Specifically, it analyzes different levels of automation and emergency condition response at the Lungmen Nuclear Power Plant in Taiwan. A simulated feed-water system was developed involving two interface styles. This study uses the NASA Task Load Index to objectively evaluate the mental workload of the human operators and the Situation Awareness Rating Technique to subjectively assess operator understanding and response, and is a pilot study investigating EID display format use at nuclear power plants in Taiwan. Results suggest the EID-based interface has a remarkable advantage over the original interface in supporting operator performance in the areas of response time and accuracy rate under both normal and emergency situations and provide supporting evidence that an EID-based interface can effectively enhance monitoring tasks in a complex environment. (author)

  16. Analysis of Total Loss of Feedwater for APR1400 using SPACE

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Min; Park, Seok Jeong; Park, Chan Eok; Choi, Jong Ho; Lee, Gyu Cheon [KEPCO Engineering and Construction, Deajeon (Korea, Republic of)

    2016-10-15

    The Total Loss of FeedWater (TLOFW) event is an accident that main feedwater and auxiliary feedwater of secondary side are not supplied to steam generators. APR1400 uses the Safety Depressurization and Vent System (SDVS) for the F and B operation and SDVS is designed to perform the rapid depressurization function of Reactor Coolant System (RCS) through the remote manual operation when TLOFW is occurred. If RCS pressure falls below a Safety Injection Pump (SIP) working pressure, it can be possible to start the F and B operation which injects SIP flow to RCS and releases the RCS vapor and two-phase flow through Pilot Operated Safety Relief Valves (POSRVs) by opening the POSRVs, and then it can be possible to remove the decay heat. The design requirement of SDVS is that the core water level should be maintained at higher than 2 feet from the top of active core during the F and B operation. The TLOFW analysis was carried out to evaluate the capability of decay heat removal for APR1400 using newly developed SPACE code. The analysis results show that the F and B operation with 2 POSRVs and 2 SIPs and the F and B operation with 4 POSRVs and 4 SIPs meet the SDVS design requirement for the fuel cladding temperature. The comparison with RELAP5 shows good agreement and it validates the applicability of SPACE code for this type of accident analysis.

  17. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    Arthur, D.

    1978-03-01

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  18. Reactor feedwater control device

    International Nuclear Information System (INIS)

    Koshi, Yuji.

    1993-01-01

    In the device of the present invention, an excess response is not caused in a reactor feed water system even when voids are fluctuated by using an actual water level signal as a reactor water level signal. That is, a standard water level signal and a reactor water level signal are inputted to a comparator. An adder adds water level difference signal outputted from the comparator and mismatch flow rate signal prepared by multiplying the difference between a main steam flow rate signal and a feed water flow rate signal by a mismatch gain. A feed water controller integrates the added signal and outputs flow rate demand signal. A feed water system receives the flow rate demand signal as input. A water level calculation means is disposed to such a device for calculating an actual water level based on the change of coolant possessing amount of the reactor, and the output thereof is defined as a reactor water level signal. With such procedures, excessive elevation of water level of the reactor can be prevented even upon occurrence of void fluctuation phenomenon or the like in the reactor such as upon sole scram operation. Accordingly, plant shut down caused thereby can be avoided safely. (I.S.)

  19. Loss of feedwater heater analysis for the South Texas Project

    International Nuclear Information System (INIS)

    Joyce, K.C.; Johnson, M.R.; Albury, C.R.

    1987-01-01

    The results of the steady state and transient analyses of the low pressure feedwater heater train for the South Texas Nuclear Project are presented. The South Texas Project consists of two 1250 MW Westinghouse PWR units. This analysis was performed using the Modular Modeling System (MMS) simulation code. The model presented will be incorporated into the secondary side model in support of the plant training simulator and the analysis of secondary side transients. Results of this analysis are considered preliminary until benchmarked against actual plant data. A model description of the feedwater heater train from the condensate pumps to the deaerator is presented. The methodology used to develop the model is also discussed. Results of the steady state run are presented, and a transient, the loss of extraction steam to feedwater heater 15A, is examined

  20. Isolators Including Main Spring Linear Guide Systems

    Science.gov (United States)

    Goold, Ryan (Inventor); Buchele, Paul (Inventor); Hindle, Timothy (Inventor); Ruebsamen, Dale Thomas (Inventor)

    2017-01-01

    Embodiments of isolators, such as three parameter isolators, including a main spring linear guide system are provided. In one embodiment, the isolator includes first and second opposing end portions, a main spring mechanically coupled between the first and second end portions, and a linear guide system extending from the first end portion, across the main spring, and toward the second end portion. The linear guide system expands and contracts in conjunction with deflection of the main spring along the working axis, while restricting displacement and rotation of the main spring along first and second axes orthogonal to the working axis.

  1. Attachment of iron corrosion products on steam generator tube and feed-water pump in PWRs secondary system

    International Nuclear Information System (INIS)

    Shoda, Y.; Ishihara, N.; Miyata, H.; Ohira, T.; Watanabe, Y.; Nonaka, Y.

    2010-01-01

    Operating experience of the secondary systems in PWRs indicates that scale attachment distinctly have an effect on the performance of water-steam cycle. Attached scale on outer surface of steam generator (SG) tube could induce many problems such as decrease heat efficiency of plant, corrosion of tube by intergranular attack (IGA), and choke of flow channel. Scale attached on rotor blade of feed water pump increases the driving steam consumption to keep the constant flow rate, and results in the thermal efficiency decrease of the plant. In this study, two types of test about scale deposition on equipment were executed in the conditions simulating the secondary system of PWR. One is SG model test, which simulated the circulating boiler composed of single SG tube and blow down line. The deposition rate under AVT condition was equivalent to plants revealed with extended period. High-AVT test provided useful reference, because the deposition rate of power plant is too small to measure in a short period after the beginning of High-AVT operation in Japan. The other is feed water pump model test. The mock-up pump is composed of a rotating stainless steel disk. As a result, it is confirmed that the deposition rate depends mostly on iron concentration in water and the exfoliation rate depends mainly on pH. Applying this information, the scale deposition-growth behavior on the equipment is quantitatively expressed by the model combined of scale deposition behavior and exfoliation behavior couples with the former. These results bring effective estimation for suppressing deposition-growth by the selection of water chemistry management and/or equipment improvement in the PWR secondary system. (author)

  2. Feedwater heater performance evaluation using the heat exchanger workstation

    International Nuclear Information System (INIS)

    Ranganathan, K.M.; Singh, G.P.; Tsou, J.L.

    1995-01-01

    A Heat Exchanger Workstation (HEW) has been developed to monitor the condition of heat exchanging equipment power plants. HEW enables engineers to analyze thermal performance and failure events for power plant feedwater heaters. The software provides tools for heat balance calculation and performance analysis. It also contains an expert system that enables performance enhancement. The Operation and Maintenance (O ampersand M) reference module on CD-ROM for HEW will be available by the end of 1995. Future developments of HEW would result in Condenser Expert System (CONES) and Balance of Plant Expert System (BOPES). HEW consists of five tightly integrated applications: A Database system for heat exchanger data storage, a Diagrammer system for creating plant heat exchanger schematics and data display, a Performance Analyst system for analyzing and predicting heat exchanger performance, a Performance Advisor expert system for expertise on improving heat exchanger performance and a Water Calculator system for computing properties of steam and water. In this paper an analysis of a feedwater heater which has been off-line is used to demonstrate how HEW can analyze the performance of the feedwater heater train and provide an economic justification for either replacing or repairing the feedwater heater

  3. Ferromagnetic material inspection for feedwater heater and condenser tubes

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    In recent years, special ferritic stainless steels, such as AL29-4C/sup TM/, Sea-Cure/sup TM/, E-Brite/sup TM/, 439, and similar alloys have been introduced as tube material in condensers, feedwater heaters, moisture separator/reheaters, and other heat exchangers. In addition, carbon steel tubes are widely used in feedwater heaters and heat exchangers in chemical plants. The main problem with the in-service inspection of these ferritic alloys and carbon steel tubes lies in their highly ferromagnetic properties. These properties severely limit the application of the standard eddy current techniques. The effort was undertaken under EPRI sponsorship to develop a reliable technique for in-service inspection of ferromagnetic tubes. The new method combines the measurement of magnetic flux leakage generated around the defects with measurement of total flux in the tube wall. The heart of the inspection system is a special ID probe that magnetizes the tube and generates signals for any tube defect. A permanent record of inspection is provided with a strip-chart or magnetic tape recorder. The laboratory and field evaluation of this new system demonstrated its very good sensitivity to small defects, its reliability, and its ruggedness. Defects as small as 10% external wall loss in heavy wall carbon steel tube were detected. Tubes in the power plant were inspected at a rate of 300-500 tubes per eight-hour shift. The other advantages of this newly developed technique are its simplicity, low cost of instrumentation, easy data interpretation, and full portability

  4. Study of thermohydraulic characteristics of upgraded feedwater collector in PGV-440 steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Tarankov, G.A.; Trunov, N.B.; Titov, V.F. [OKB Gidropress (Russian Federation); Urbansky, V.V. [Rovno NPP (Ukraine); Lenkei, I.; Notarosh, M. [Paks NPP (Hungary)

    1995-12-31

    Reconstruction of feedwater distribution collector was performed at unit 1 of Rowno NPP. Main results of measurements of temperatures in water volume, reparation characteristics and impurities distribution are presented. Analysis of tests results and design criteria is given. (orig.).

  5. Study of thermohydraulic characteristics of upgraded feedwater collector in PGV-440 steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Tarankov, G A; Trunov, N B; Titov, V F [OKB Gidropress (Russian Federation); Urbansky, V V [Rovno NPP (Ukraine); Lenkei, I; Notarosh, M [Paks NPP (Hungary)

    1996-12-31

    Reconstruction of feedwater distribution collector was performed at unit 1 of Rowno NPP. Main results of measurements of temperatures in water volume, reparation characteristics and impurities distribution are presented. Analysis of tests results and design criteria is given. (orig.).

  6. Simulating the transient regime for main condensate system at Cernavoda NPP

    International Nuclear Information System (INIS)

    Nita, Iulian; Gheorghiu, Mihai; Prisecaru, Ilie; Dupleac, Daniel

    2005-01-01

    The purpose of this project is to make a Thermal Hydraulic Analysis of Main Condensate System for getting real-time answer of installation during regimes occurring during normal and abnormal operation. To obtain the analyses the MMS code was used. The boundaries of the systems analysis are extended to Main Feedwater System in order to get a realistic response of Deaerator equipment which are situated between those two systems and have entrances from both systems. In this way we made a complex analysis with main condenser and steam generators as boundaries. We obtained a model for the entire chain of condensate and feedwater preheater with interface just turbine bleed steam. From that we could reduce the number of assumptions necessary to make the analysis. The analyses consist in hydraulics and thermal hydraulics analyses, respectively. For the first case analysed are: - the nominal operation regime with main condensate pumps; - start-up regime with total circulate of condensate to condenser; - 25% MCR (Maximum Continuous Rate) regime (this regime was used in designing the condensate regulating valves at low flow; - 40% MCR regime (with circulate of some condensate flow to condenser); - operating regime of 60% MCR with one main condensate pump operating; - operating regime with auxiliary condensate pump; - operating regime with discharging a condensate flow to condensate storage tank. The thermal hydraulic analyses deal with normal and abnormal operating regimes, respectively. In the first case analysed are the following regimes: - nominal operating regime with main condensate pump operating 100% MCR; - transient regime, 100-80% MCR; - transient regime, 100-80-60% MCR with two pumps in operation and 60 % MCR with one main condensate pump in operation; - transient regime, 100-80-60-60-40 % MCR; - shut-down regime; - start-up regime from Hot zero power to rated power regime. Finally, for the abnormal operating regimes the analyses concerned: - transient regime 100

  7. Analysis of a postulated pipe rupture and subsequent check valve slam of a PWR feedwater line

    International Nuclear Information System (INIS)

    Chang, K.C.; Adams, T.M.

    1983-01-01

    System designs criteria employed in the design of pressurized water reactors (PWR) requires that, for a postulated instantaneous guillotine rupture anywhere in the steam generator feedwater system, no more than one steam generator can be allowed to blowdown. Feedwater systems in many PWR's consist of pipe lines running from the feedwater pumps into a common feedwater header then branching into each steam generator from the header. The feedwater piping to each steam generator contains swing check valves to prevent reverse flow from the steam generator. This activation of some or all of these check valves significantly complicates the system structural analysis in that not only the blowdown forces resulting from the postulated pipe rupture, but also the water hammer loads resulting from closure of the check valve at high reverse flow velocities must be considered. The loads resulting from system blowdown and check valve closure are axial in nature. Peak loads ranging from 130000 lbs. to 180000 lbs. are not uncommon and are layout dependent. The analysis and design to withstand this transient loading deviates from the usual feedwater line design in that supports are required along the piping axis in the direction normal to the usual seismic supports. A brief and general discussion of the methods employed in the generation of the thermal-hydraulic loadings is presented. However, the discussion emphasizes the piping and piping support structural design and analysis method and approaches used in evaluating a selected portion of such a feedwater system. (orig./RW)

  8. Modeling and simulation of the feedwater system, associated controller and interface with the user for the SUN-RAH nucleo electric plants university student simulator; Modelado y simulacion del sistema de agua de alimentacion, controlador asociado e interfaz con el usuario para el simulador universitario de nucleoelectricas SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez B, A. [Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: alitet@eresmas.com

    2003-07-01

    The simulation process of the component systems of the feedwater of a nucleo electric plant is presented, using several models of reduced order that represent the diverse elements that compose the systems like: the heaters of feedwater, the condenser, the feedwater pump, etc. The integration of the same ones in one simulative structure, and the development of a platform that to give the appearance of to be executed in continuous time, it is the objective of the feedwater simulator, as well as of the SUN-RAH simulator, of which is part. The simulator uses models of reduced order that respond to the observed behavior of a nuclear plant of BWR type. Likewise, it is presented a model of a flow controller of feedwater that will be the one in charge of regulating the demand of the system according to the characteristics and criticize restrictions of safety and controllability, assigned according to those wanted parameters of performance of this system inside the nucleo electric plant. The integration of these models, the adaptation of the variables and parameters, are presented in a way that the integration with the other ones models of the remaining systems of the plant (reactor, steam lines, turbine, etc.), be direct and coherent with the principles of thermodynamic cycles relative to this type of generation plants. The design of those graphic interfaces and the environment where the simulator works its are part of those developments of this work. The reaches and objectives of the simulator complement the description of the simulator. (Author)

  9. Application of the methodology of safety probabilistic analysis to the modelling the emergency feedwater system of Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Troncoso, M.; Oliva, G.

    1993-01-01

    The application of the methodology developed in the framework of the national plan of safety probabilistic analysis (APS) to the emergency feed water system for the failures of small LOCAS and external electrical supply loss in the nuclear power plant is illustrated in this work. The facilities created by the ARCON code to model the systems and its documentation are also expounded

  10. Computer-generated direct perception displays for supporting PWR feedwater system start-up and fault management: a proof-of-principle in design

    International Nuclear Information System (INIS)

    Reising, D.V.C.; Jones, B.G.; Shaheen, S.; Moray, N.; Sanderson, P.M.; Rasmussen, J.

    1998-01-01

    difficult problems which have not yet been investigated in extending the proposed approach to fault management. In the present research Rasmussen et al's framework was used for designing computer-generated graphical displays that support pressurized water reactor (PWR) start-up. Specifically, a suite of displays was developed to support a PWR's feedwater (FW) system start-up as a proof-of-principle. The suite of displays demonstrate the theoretical design approach and are not meant to represent a fully implementable interface for FW system control. (author)

  11. Alarm system for ABWR main control panels

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Yuji; Saito, Koji [Toshiba Corp., Yokohoma (Japan)

    1997-09-01

    TOSHIBA has developed integrated digital control and instrumentation system for ABWR, which is the third-generation man machine interface system for main control room that we call A-PODIA (Advanced PODIA). A-Podia has been introduced the first actual ABWR plant in Japan. in A-PODIA, TOSHIBA has realized improvement of alarm system that all operator crews in the control room can recognize plant anomalies easily. The alarm system can recognize essential alarms for plant safety easily and understand annunciators with each integrated annunciators and their prioritized color easily by classifying alarms into plant-level essential annunciators, system-level integrated annunciators and equipment level individual annunciators with hierarchical structure. This paper describes conventional alarm system and the design philosophy, alarm system design and operation of ``Alarm System for ABWR Main Control Panels``. (author). 5 refs, 8 figs, 1 tab.

  12. Alarm system for ABWR main control panels

    International Nuclear Information System (INIS)

    Kobayashi, Yuji; Saito, Koji

    1997-01-01

    TOSHIBA has developed integrated digital control and instrumentation system for ABWR, which is the third-generation man machine interface system for main control room that we call A-PODIA (Advanced PODIA). A-Podia has been introduced the first actual ABWR plant in Japan. in A-PODIA, TOSHIBA has realized improvement of alarm system that all operator crews in the control room can recognize plant anomalies easily. The alarm system can recognize essential alarms for plant safety easily and understand annunciators with each integrated annunciators and their prioritized color easily by classifying alarms into plant-level essential annunciators, system-level integrated annunciators and equipment level individual annunciators with hierarchical structure. This paper describes conventional alarm system and the design philosophy, alarm system design and operation of ''Alarm System for ABWR Main Control Panels''. (author). 5 refs, 8 figs, 1 tab

  13. Numerical simulation of a 374 tons/h water-tube steam boiler following a feedwater line break

    International Nuclear Information System (INIS)

    Deghal Cheridi, Amina Lyria; Chaker, Abla; Loubar, Ahcène

    2016-01-01

    Highlights: • We simulate the behavior of a steam boiler during feed-water line break accident. • To perform accident analysis of the steam boiler, Relap5/Mod3.2 system code is used. • A Relap5 model of the boiler is developed and qualified at the steady state level. • A good agreement between Relap5 results and available experimental data. • The Relap5 model predicts well the main transient features of the boiler. - Abstract: To ensure the operational safety of an industrial water-tube steam boiler it is very important to assess various accident scenarios in real plant working conditions. One of the most challenging scenarios is the loss of feedwater to the steam boiler. In this paper, a simulation of the behavior of an industrial water-tube radiant steam boiler during feedwater line break accident is discussed. The simulation is carried out using the RELAP5 system code. The steam boiler is installed in an Algerian natural gas liquefaction complex. The simulation shows the capabilities of RELAP5 system code in predicting the behavior of the steam boiler at both steady state and transient working conditions. From another side, the behavior of the steam boiler following the accident shows how the control system can successfully mitigate the effects and consequences of such accident and how the evaporator tubes can undergo a severe damage due to an uncontrolled increase of the wall temperature in case of failure of this system.

  14. Trace analysis of auxiliary feedwater capacity for Maanshan PWR loss-of-normal-feedwater transient

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Che-Hao; Shih, Chunkuan [National Tsing Hua Univ., Taiwan (China). Inst. of Nuclear Engineering and Science; Wang, Jong-Rong; Lin, Hao-Tzu [Atomic Energy Council, Taiwan (China). Inst. of Nuclear Energy Research

    2013-07-01

    Maanshan nuclear power plant is a Westinghouse PWR of Taiwan Power Company (Taipower, TPC). A few years ago, TPC has made many assessments in order to uprate the power of Maanshan NPP. The assessments include NSSS (Nuclear Steam Supply System) parameters calculation, uncertainty acceptance, integrity of pressure vessel, reliability of auxiliary systems, and transient analyses, etc. Since the Fukushima Daiichi accident happened, it is necessary to consider transients with multiple-failure. Base on the analysis, we further study the auxiliary feedwater capability for Loss-of-Normal-Feedwater (LONF) transient. LONF is the limiting transient of non-turbine trip initiated event for ATWS (Anticipated Transient Without Scram) which results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If the turbine fails to trip immediately, the secondary water inventory will decrease significantly before the actuation of auxiliary feedwater (AFW) system. The heat removal from the primary side decreases, and this leads to increases of primary coolant temperature and pressure. The water level of pressurizer also increases subsequently. The heat removal through the relief valves and the auxiliary feedwater is not sufficient to fully cope with the heat generation from primary side. The pressurizer will be filled with water finally, and the RCS pressure might rise above the set point of relief valves for water discharge. RCS pressure depends on steam generator inventory, primary coolant temperature, negative reactivity feedback, and core power, etc. The RCS pressure may reach its peak after core power reduction. According to ASME Code Level C service limit criteria, the Reactor Coolant System (RCS) pressure must be under 22.06 MPa. The USNRC is developing an advanced thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development of TRACE is based on TRAC and integrating with RELAP5 and other programs. SNAP

  15. Trace analysis of auxiliary feedwater capacity for Maanshan PWR loss-of-normal-feedwater transient

    International Nuclear Information System (INIS)

    Chen, Che-Hao; Shih, Chunkuan; Wang, Jong-Rong; Lin, Hao-Tzu

    2013-01-01

    Maanshan nuclear power plant is a Westinghouse PWR of Taiwan Power Company (Taipower, TPC). A few years ago, TPC has made many assessments in order to uprate the power of Maanshan NPP. The assessments include NSSS (Nuclear Steam Supply System) parameters calculation, uncertainty acceptance, integrity of pressure vessel, reliability of auxiliary systems, and transient analyses, etc. Since the Fukushima Daiichi accident happened, it is necessary to consider transients with multiple-failure. Base on the analysis, we further study the auxiliary feedwater capability for Loss-of-Normal-Feedwater (LONF) transient. LONF is the limiting transient of non-turbine trip initiated event for ATWS (Anticipated Transient Without Scram) which results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If the turbine fails to trip immediately, the secondary water inventory will decrease significantly before the actuation of auxiliary feedwater (AFW) system. The heat removal from the primary side decreases, and this leads to increases of primary coolant temperature and pressure. The water level of pressurizer also increases subsequently. The heat removal through the relief valves and the auxiliary feedwater is not sufficient to fully cope with the heat generation from primary side. The pressurizer will be filled with water finally, and the RCS pressure might rise above the set point of relief valves for water discharge. RCS pressure depends on steam generator inventory, primary coolant temperature, negative reactivity feedback, and core power, etc. The RCS pressure may reach its peak after core power reduction. According to ASME Code Level C service limit criteria, the Reactor Coolant System (RCS) pressure must be under 22.06 MPa. The USNRC is developing an advanced thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development of TRACE is based on TRAC and integrating with RELAP5 and other programs. SNAP

  16. Post-main-sequence planetary system evolution

    Science.gov (United States)

    Veras, Dimitri

    2016-01-01

    The fates of planetary systems provide unassailable insights into their formation and represent rich cross-disciplinary dynamical laboratories. Mounting observations of post-main-sequence planetary systems necessitate a complementary level of theoretical scrutiny. Here, I review the diverse dynamical processes which affect planets, asteroids, comets and pebbles as their parent stars evolve into giant branch, white dwarf and neutron stars. This reference provides a foundation for the interpretation and modelling of currently known systems and upcoming discoveries. PMID:26998326

  17. 49 CFR 230.57 - Injectors and feedwater pumps.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Injectors and feedwater pumps. 230.57 Section 230... Appurtenances Injectors, Feedwater Pumps, and Flue Plugs § 230.57 Injectors and feedwater pumps. (a) Water.... Injectors and feedwater pumps must be kept in good condition, free from scale, and must be tested at the...

  18. Isolated systems with wind power. Main report

    DEFF Research Database (Denmark)

    Lundsager, P.; Bindner, Henrik W.; Clausen, Niels-Erik

    2001-01-01

    The overall objective of this research project is to study the development of methods and guidelines rather than "universal solutions" for the use of wind energy in isolated communities. The main specific objective of the project is to develop and present amore unified and generally applicable...... approach for assessing the technical and economical feasibility of isolated power supply systems with wind energy. As a part of the project the following tasks were carried out: Review of literature, fieldmeasurements in Egypt, development of an inventory of small isolated systems, overview of end...... for Isolated Systems with Wind Power, applicable for international organisations such as donoragencies and development banks....

  19. Simulation of the behaviour of a servo actuated check valve upon rupture of the feedwater pipe

    International Nuclear Information System (INIS)

    Lucas, A.M. de; Perezagua, R.L.; Rosa, B. de la; Sanz, J.

    1995-01-01

    The steam generator replacement programme at Almaraz NPP, provides for the installation of a replacement damped non-return valve for the feedwater system. the function of this valve is to protect the steam generator in the event of a rupture in the feedwater pipe. Sudden closure of the check valve, against the flow and following rupture of the feedwater pipe, causes overpressure in the valve which is transmitted to the steam generator nozzle. It is therefore necessary to know this when designing the internal systems of the steam generator. Using the RELAP5/MODE3 code, it has been possible to simulate the dynamic behaviour of a check valve upon rupture of a feedwater pipe postulated outside the containment. The calculation model has been applied to different types of check valve. (Author)

  20. Isolated systems with wind power. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Lundsager, P.; Bindner, H.; Clausen, N.E.; Frandsen, S.; Hansen, L.H.; Hansen, J.C.

    2001-06-01

    The overall objective of this research project is to study the development of methods and guidelines rather than 'universal solutions' for the use of wind energy in isolated communities. The main specific objective of the project is to develop and present a more unified and generally applicable approach for assessing the technical and economical feasibility of isolated power supply systems with wind energy. As a part of the project the following tasks were carried out: Review of literature, field measurements in Egypt, development of an inventory of small isolated systems, overview of end-user demands, analysis of findings and development of proposed guidelines. The project is reported in one main report and four topical reports, all of them issued as Risoe reports. This is the Main Report Risoe-R-1256, summing up the activities and findings of the project and outlining an Implementation Strategy for Isolated Systems with Wind Power, applicable for international organisations such as donor agencies and development banks. (au)

  1. Feedwater connection repair and modification at GKN

    Energy Technology Data Exchange (ETDEWEB)

    Witteman, C; Klees, J E

    1985-03-01

    From January to March 1983 the feedwater connection of GKN was repaired using a boring lathe, spark machining and semi-automatic welding. Nondestructive examination was performed by ultrasonic and eddy-current testing.

  2. Feedwater connection repair and modification at GKN

    International Nuclear Information System (INIS)

    Witteman, C.; Klees, J.E.

    1985-01-01

    From Jan. to March 1983 the feedwater connection of GKN was repaired using a boring lathe, spark machining and semi-automatic welding. Nondestructive examination was performed by ultrasonic and eddy-current testing

  3. Boiler feedwater treatment using reverse osmosis at Suncor OSG

    International Nuclear Information System (INIS)

    Brown, T.

    1997-01-01

    The installation of a new 1000 cu m/hr reverse osmosis water treatment system for boiler feedwater at a Suncor plant was discussed. The selection process began in 1993 when Suncor identified a need to increase its boiler feedwater capacity. The company reviewed many options available to increase the treated water capacity. These included: contracting the supply of treated water, adding additional capacity, replacing the entire plant, reverse osmosis, and demineralization. The eventual decision was to build a new 1000 cu m/hr reverse osmosis water treatment plant with the following key components: a Degremont Infilco Ultra Pulsator Clarifier and a Glegg Water Conditioning multimedia filter, Amberpack softeners and reverse osmosis arrays. The reverse osmosis plant was environmentally favourable over an equivalent demineralization plant. A technical comparison was provided between demineralization and reverse osmosis. The system has proven to be successful and economical in meeting the plants needs. 5 figs

  4. Preliminary design of RDE feedwater pump impeller

    International Nuclear Information System (INIS)

    Sri Sudadiyo

    2018-01-01

    Nowadays, pumps are being widely used in the thermal power generation including nuclear power plants. Reaktor Daya Experimental (RDE) is a proposed nuclear reactor concept for the type of nuclear power plant in Indonesia. This RDE has thermal power 10 MW th , and uses a feedwater pump within its steam cycle. The performance of feedwater pump depends on size and geometry of impeller model, such as the number of blades and the blade angle. The purpose of this study is to perform a preliminary design on an impeller of feedwater pump for RDE and to simulate its performance characteristics. The Fortran code is used as an aid in data calculation in order to rapidly compute the blade shape of feedwater pump impeller, particularly for a RDE case. The calculations analyses is solved by utilizing empirical correlations, which are related to size and geometry of a pump impeller model, while performance characteristics analysis is done based on velocity triangle diagram. The effect of leakage, pass through the impeller due to the required clearances between the feedwater pump impeller and the volute channel, is also considered. Comparison between the feedwater pump of HTR-10 and of RDE shows similarity in the trend line of curve shape. These characteristics curves will be very useful for the values prediction of performance of a RDE feedwater pump. Preliminary design of feedwater pump provides the size and geometry of impeller blade model with 5-blades, inlet angle 14.5 degrees, exit angle 25 degrees, inside diameter 81.3 mm, exit diameter 275.2 mm, thickness 4.7 mm, and height 14.1 mm. In addition, the optimal values of performance characteristics were obtained when flow capacity was 4.8 kg/s, fluid head was 29.1 m, shaft mechanical power was 2.64 kW, and efficiency was 52 % at rotational speed 1750 rpm. (author)

  5. PLEIADES SYSTEM ARCHITECTURE AND MAIN PERFORMANCES

    Directory of Open Access Journals (Sweden)

    M. A. Gleyzes

    2012-07-01

    Full Text Available France, under the leadership of the French Space Agency (CNES, has set up a cooperative program with Austria, Belgium, Spain, Sweden, in order to develop a space Earth Observation system called PLEIADES. PLEIADES is a dual system, this means that it is intended to fulfill an extended panel of both civilian and Defense user’s needs.. This paper reports the status of the satellite after its launch and the in orbit commissioning, the PLEIADES satellite first model has been launched at the end of year 2011, the second model will be launched about 12 months later. It describes the main mission characteristics and performances status. It exposes how the system, satellite and ground segment have been designed in order to be compliant with a dual exploitation between civilian and defense partners. The system is based on the use of a set of newly European developed technologies to feature the satellite. In order to maximize the agility of the satellite, weight and inertia have been reduced using a compact hexagonal shape for the satellite bus. The optical mission consists in Earth optical observation composed of 0.7 m nadir resolution for the panchromatic band and 2.8 m nadir resolution for the four multi-spectral bands. The image swath is about 20 km. PLEIADES delivers optical high resolution products consisting in a Panchromatic image, into which is merged a four multispectral bands image, orthorectified on a Digital Terrain Model (DTM. Thanks to the huge satellite agility obtained with control momentum gyros as actuators, the optical system delivers as well instantaneous stereo images, under different stereoscopic conditions and mosaic images, issued from along the track thus enlarging the field of view. The ground segment is composed of a dual ground center located in CNES Toulouse premises in charge of preparing the dual mission command plan and of the real time contacts with the satellite through a control center. The dual ground center

  6. Trace analysis of loss of feedwater flow event in Lungmen ABWR

    International Nuclear Information System (INIS)

    Wang Jongrong; Lin Haotzu; Wang Weichen; Yang Shuming; Shih Chunkuan

    2009-01-01

    TRACE (TRAC/RELAP Advanced Computational Engine) model of Lungmen Nuclear Power Plant was used to analyze the Loss of Feedwater Flow transient as defined in Lungmen FSAR Chapter 15. The results were compared with those from FSAR and RETRAN02. Lungmen TRACE model will have two models: In model A, vessel is divided into 11 axial levels, 4 radial rings and 1 azimuthal sectors; In model B, vessel is divided into 11 axial levels, 4 radial rings, and 6 azimuthal sectors. The above models include feedwater control system, narrow range water level control system, and wide range water level control system. The loss of feedwater flow (LOFW) transient began with the trip of two operating feedwater pumps either from the pump mechanical/electric failure, or the operator human error, or high water level signal. Feedwater flow was assumed to descend to 0 in 5 seconds and led to the decrease of reactor water level. At L3 low water level setpoint, the system actuated reactor scram signal and RIP trip signal for RIPs not connected to the M/G set. At L2 low-low water level setpoint, the system would trip the other six RIPs. This paper compares those important thermal parameters at steady state, such as the dome pressure and temperature of reactor vessel, steam flow, feedwater flow, core flow, and RIP flow, etc.. It also compares system parameters under transient conditions, such as core thermal power, core flow, steam flow, feedwater flow, Narrow Range Water Level (NRWL), Wide Range Water Level (WRWL) and RIP flow, etc.. It was concluded that the steady state and transient results of TRACE calculations are in good agreement with those from RETRAN02. In summary, our studies concluded that Lungmen TRACE model is correct and accurate enough for future safety analysis applications. (author)

  7. Feedwater flow measurements: challenges, current solutions, and 'soft' developments

    International Nuclear Information System (INIS)

    Ruan, D.; Roverso, D.; Fantoni, P.F.; Sanabrias, J.I.; Carrasco, J.A.; Fernandez, L.

    2002-07-01

    This report presents an early progress of a feasibility study of a computational intelligence approach to the enhancement of the accuracy of feedwater flow measurements in the framework of an ongoing cooperation between Tecnatom s.a. in Madrid and the OECD Halden Reactor Project (HRP) in Halden. The aim of this research project is to contribute to the development and validation of a flow sensor in a nuclear power plant (NPP). The basic idea is to combine the use of applied computational intelligence approaches (noise analysis, neural networks, fuzzy systems, wavelets etc.) with existing traditional flow measurements, and in particular with cross correlation flow meter concepts. In this report, Section 2 outlines relevant aspects of thermal power calculations on electrical power plants. Section 3 reviews from the available literature possible approaches and solutions for feedwater flow measurement, including ultrasonic flow meters, cross-correlation flow meters, and 'Virtural' flow meters with artificial neural networks. Section 4 reports typical experimental measurements at the Tecnatom's facility. Section 5 presents an integration approach and preliminary experimental tests. Section 6 discusses the role of soft computing techniques in the context of feedwater flow measurements related nuclear fields, and Section 7 highlights the future research direction. (Author)

  8. Revisiting Reuse in Main Memory Database Systems

    OpenAIRE

    Dursun, Kayhan; Binnig, Carsten; Cetintemel, Ugur; Kraska, Tim

    2016-01-01

    Reusing intermediates in databases to speed-up analytical query processing has been studied in the past. Existing solutions typically require intermediate results of individual operators to be materialized into temporary tables to be considered for reuse in subsequent queries. However, these approaches are fundamentally ill-suited for use in modern main memory databases. The reason is that modern main memory DBMSs are typically limited by the bandwidth of the memory bus, thus query execution ...

  9. Evolution of carbon steel corrosion in feedwater conditions reproduce by the Fortrand loop

    International Nuclear Information System (INIS)

    Delaunay, Sophie; Bescond, Aurelien; Mansour, Carine; Bretelle, Jean-Luc

    2012-09-01

    Fouling and tubes support plate blockage of steam generators (SG) are major problems in the secondary circuit of pressurized water reactor (PWR) plants. Corrosion products (CP) responsible of these phenomena are mainly constituted of magnetite. Limit the amount of these CP, generated in the feedwater system and transported to SG, constitutes one way to limit fouling and blockage of SGs. This work requires the understanding of CP behaviour in the feedwater system conditions. A specific experimental circulating water loop, FORTRAND, was built at EDF to follow the formation, the transport and the deposition of iron oxides in representative conditions of the secondary circuit feedwater system. The test section operating at high temperature (up to 250 deg. C) is made in carbon steel and includes three removable segments while all the other parts of the loop are made in stainless steel. First results confirm the formation of iron oxides on carbon steel and stainless steel surface in the conditions of PWR secondary circuits. The surface characterizations show that magnetite is the corrosion product formed on carbon steel and stainless steel at 220 deg. C and goethite is formed at room temperature on stainless steel. The aim of the most recent tests performed in FORTRAND loop was to follow the evolution of corrosion in the feedwater conditions. Tests were performed in one-phase flow conditions at 150 L.h -1 with a linear velocity of 0.82 m/s at 220 deg. C in morpholine/ammonia/hydrazine medium, at pH 25C equal to 9.2. To conduct this study, a removable segment constituted by ten tubes was added to the loop. Several tests were performed to follow the deposit thickness, the iron lost in solution and the oxide morphology with time from two to nine hundred sixty hours. Chemical conditions were controlled and the reproducibility of the results was confirmed by the observation of three tubes at each test. SEM pictures present kinetics with three steps: after the first hours the

  10. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, R.; Harrell, J.

    1996-12-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

  11. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    International Nuclear Information System (INIS)

    Fuller, R.; Harrell, J.

    1996-01-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves

  12. Multi-unit shutdown due to boiler feedwater chemical excursion

    International Nuclear Information System (INIS)

    Diebel, M.E.

    1991-01-01

    Ontario Hydro's Bruce Nuclear Generating Station 'B' consists of four 935 W CANDU units located on the east shore of Lake Huron in the province of Ontario, Canada. On July 25 and 26, 1989 three of the four operating units were shutdown due to boiler feedwater chemical excursions initiated by a process upset in the Water Treatment Plant that provides demineralized make-up water to all four units. The chemicals that escaped from an ion exchange vessel during a routine regeneration very quickly spread through the condensate make-up system and into the boiler feedwater systems. This resulted in boiler sulfate levels exceeding shutdown limits. A total of 260 GWH of electrical generation was unexpectedly made unavailable to the grid at a time of peak seasonal demand. This event exposed several unforeseen deficiencies and vulnerabilities in the automatic demineralized water make-up quality protection scheme, system designs, operating procedures and the ability of operating personnel to recognize and appropriately respond to such an event. The combination of these factors contributed towards turning a minor system upset into a major multi-unit shutdown. This paper provides the details of the actual event initiation in the Water Treatment Plant and describes the sequence of events that led to the eventual shutdown of three units and near shutdown of the fourth. The design inadequacies, procedural deficiencies and operating personnel responses and difficulties are described. The process of recovering from this event, the flushing out of system piping, boilers and the feedwater train is covered as well as our experiences with setting up supplemental demineralized water supplies including trucking in water and the use of rental trailer mounted demineralizing systems. System design, procedural and operational changes that have been made and that are still being worked on in response to this event are described. The latest evidence of the effect of this event on boiler tube

  13. Thermal-hydraulics of PGV-4 water volume during damage of the feedwater collector nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Logvinov, S.A.; Titov, V.F. [OKB Gidropress (Russian Federation); Notaros, U.; Lenkei, I. [NPP Paks (Hungary)

    1995-12-31

    A number of VVER-440 plants has experienced the distributing nozzles of feedwater collector being damaged due to corrosion-erosion wearing. Such phenomenon could result in feedwater redistribution within the SG inventory with undesirable consequences. The collector with damaged nozzles has to be replaced but a certain time is needed for the preparatory works. The main objective of the investigation conducted is to assess if the safe operation of SG is possible before collector replacement. It was shown that the nozzle damage as observed did not result in the dangerous disturbances of thermobydraulics as compared with the conditions existing at the initial period of operation. (orig.).

  14. Thermal-hydraulics of PGV-4 water volume during damage of the feedwater collector nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Logvinov, S A; Titov, V F [OKB Gidropress (Russian Federation); Notaros, U; Lenkei, I [NPP Paks (Hungary)

    1996-12-31

    A number of VVER-440 plants has experienced the distributing nozzles of feedwater collector being damaged due to corrosion-erosion wearing. Such phenomenon could result in feedwater redistribution within the SG inventory with undesirable consequences. The collector with damaged nozzles has to be replaced but a certain time is needed for the preparatory works. The main objective of the investigation conducted is to assess if the safe operation of SG is possible before collector replacement. It was shown that the nozzle damage as observed did not result in the dangerous disturbances of thermobydraulics as compared with the conditions existing at the initial period of operation. (orig.).

  15. Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

    International Nuclear Information System (INIS)

    Nurkkala, P.; Hoikkanen, J.

    1997-01-01

    This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. 'grounded' and 'with goose neck'). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.)

  16. Space Shuttle Main Propulsion System Anomaly Detection: A Case Study

    Data.gov (United States)

    National Aeronautics and Space Administration — The space shuttle main engine (SSME) is part of the Main Propnlsion System (MPS) which is an extremely complex system containing several sub-systems and components,...

  17. Evaluation of examination techniques for ferritic stainless steel feedwater heater tubing

    International Nuclear Information System (INIS)

    Nugent, M.J.; Catapano, M.C.

    1995-01-01

    Ferritic stainless steel has been finding increased application in utility plant feedwater heaters due to good strength and corrosion resistance and absence of potential copper contamination of feedwater system. Ferritic stainless steel is highly magnetic and is generally not inspectable using conventional eddy current testing techniques. A variety of techniques have been developed for inspection of this tubing material used in typical heat exchanger applications. Through a project funded by the Empire State Electric Energy Research Corporation (ESEERCO), the evaluation of data generated by four present state of the art NDE testing techniques were evaluated on a controlled mock-up of the heater tubing with service related defects. The primary objective was to determine the strengths and limitations of each method. The testing of two in service feedwater heaters at the Consolidated Edison Company of New York, Inc. (Con Edison's) Arthur Kill Generating Station also allowed further evaluations based on actual field conditions

  18. RIMACS, Reactor Inspection Main Control System

    International Nuclear Information System (INIS)

    2008-01-01

    1 - Description of program or function: RIMACS prepares for automatic inspection files on each inspection item for the reactor. These automatic inspection files provide the data to move RIROB (Reactor Inspection Robot) with laser by interpreting the coordinates of LASPO (Laser Positioner) and the laser detecting device of RIROB in three dimensional space. In addition, when RIROB arrives at the inspecting location, the files provide all values of the manipulator's motions to acquire the ultrasonic data. RIMACS provides various modules in order to perform these complex functions, and the functions are programmed on graphic user interface for the convenience of the user. RIMACS provides various functions, such as insertion of reactor production data, selection of the reactor for inspection, the creation of automatic inspection file, the selection of the inspection item, inspection simulation, and automatic inspection procedures. It also provides all other functions, which are necessary for the inspection, such as operating program download and manual control of LASPO and RIROB, the inspection simulation and the inspection status display by means of the graphic screen, and SODAS (ultra-Sonic Data Acquisition System) drive verification. 2 - Methods: Moving path and operation procedures for inspection robot are generated automatically with Kinematics algorithm. 3 - Restrictions on the complexity of the problem: A graphics display with MS-Window capability is required

  19. Analysis of steam-generator tube-rupture events combined with auxiliary-feedwater control-system failure for Three Mile Island-Unit 1 and Zion-Unit 1 pressurized water reactors

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1986-01-01

    A steam-generator tube-rupture (SGTR) event combined with loss of all offsite alternating-current power and failure of the auxiliary-feedwater (AFW) control system has been investigated for the Three Mile Island-Unit 1 (TMI-1) and Zion-Unit 1 (Zion-1) pressurized water reactors. The Transient Reactor Analysis Code was used to simulate the accident sequence for each plant. The objectives of the study were to predict the plant transient response with respect to tube-rupture flow termination, extent of steam generator overfill, and thermal-hydraulic conditions in the steam lines. Two transient cases were calculated: (1) a TMI-1 SGTR and runaway-AFW transient, and (2) a Zion-1 SGTR and runaway-AFW transient. Operator actions terminated the tube-rupture flow by 1342 s (22.4 min) and 1440 s (24.0 min) for TMI-1 and Zion-1, respectively, but AFW injection was continued. The damaged steam generator (DSG) overfilled by 1273 s (21.2 min) for the TMI-1 calculation and by 1604 s (26.7 min) for the Zion-1 calculation. The DSG steam lines were completely filled by 1500 s (25 min) and 2000 s (33.3 min) for TMI-1 and Zion-1, respectively. The maximum subcooling in the steam lines was approx.63 K (approx.113 0 F) for TMI-1 and approx.44 K (approx.80 0 F) for Zion-1

  20. Aging and low-flow degradation of auxilary feedwater pumps

    International Nuclear Information System (INIS)

    Adams, M.L.

    1992-01-01

    This paper documents the results of research done under the auspices of the Nuclear Regulatory Commission Nuclear Plant Aging Research Program. It examines the degradation imparted to safety related Auxiliary Feedwater System pumps at nuclear plants due to the low flow operation. The Auxiliary Feedwater (AFW) System is normally a stand-by system. As such it is operated most often in the test mode. Since few plants are equipped with full flow test loops, most testing is accomplished at minimum flow conditions in pump by-pass lines. It is the vibration and hydraulic forces generated at low flow conditions that have been shown to be the major causes of AFW pump aging and degradation. The wear can be manifested in a number of ways, such as impeller or diffuser breakage, thrust bearing and/or balance device failure due to excessive loading, cavitation damage on such stage impellers, increase seal leakage or failure, sear injection piping failure, shaft or coupling breakage, and rotating element seizure

  1. Aging and low-flow degradation of auxiliary feedwater pumps

    International Nuclear Information System (INIS)

    Adams, M.L.

    1991-01-01

    This paper documents the results of research done under the auspices of the Nuclear Regulatory Commission Nuclear Plant Aging Research Program. It examines the degradation imparted to safety Auxiliary Feedwater System pumps at nuclear plants due to the low flow operation. The Auxiliary Feedwater (AFW) System is normally a stand-by system. As such it is operated most often in the test mode. Since few plants are equipped with full flow test loops, most testing is accomplished at minimum flow conditions in pump by-pass lines. It is the vibration and hydraulic forces generated at low flow conditions that have been shown to be the major causes of AFW pump aging and degradation. The wear can be manifested in a number of ways, such as impeller or diffuser breakage, thrust bearing and/or balance device failure due to excessive loading, cavitation damage on such stage impellers, increase seal leakage or failure, sear injection piping failure, shaft or coupling breakage, and rotating element seizure

  2. Investigation of accident management procedures related to loss of feedwater and station blackout in PSB-VVER integral test facility

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, A. [EC JRC, (JRC F.5) PO Box 2, 1755 ZG Petten (Netherlands); Del Nevo, A., E-mail: alessandro.delnevo@enea.it [ENEA, C.R. Brasimone, 40032 Camugnano (Italy); Moretti, F.; D' Auria, F. [GRNSPG, Universita di Pisa, via Diotisalvi 2, 56100 Pisa (Italy); Elkin, I.V.; Melikhov, O.I. [Electrogorsk Research and Engineering Centre, Electrogorsk, Moscow Region (Russian Federation)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Four integral test facility experiments related to VVER-1000 reactor. Black-Right-Pointing-Pointer TH response of the VVER-1000 primary system following total loss of feedwater and station blackout scenarios. Black-Right-Pointing-Pointer Accident management procedures in case of total loss of feedwater and station blackout. Black-Right-Pointing-Pointer Experimental data represent an improvement of existing database for TH code validation. - Abstract: VVER 1000 reactors have some unique and specific features (e.g. large primary and secondary side fluid inventory, horizontal steam generators, core design) that require dedicated experimental and analytical analyses in order to assess the performance of safety systems and the effectiveness of possible accident management strategies. The European Commission funded project 'TACIS 2.03/97', Part A, provided valuable experimental data from the large-scale (1:300) PSB-VVER test facility, investigating accident management procedures in VVER-1000 reactor. A test matrix was developed at University of Pisa (responsible of the project) with the objective of obtaining the experimental data not covered by the OECD VVER validation matrix and with main focus on accident management procedures. Scenarios related to total loss of feed water and station blackout are investigated by means of four experiments accounting for different countermeasures, based on secondary cooling strategies and primary feed and bleed procedures. The transients are analyzed thoroughly focusing on the identification of phenomena that will challenge the code models during the simulations.

  3. Operational challenges to feedwater/steam generator water level control

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, V.M.; Whaley, S.D.; Federico, P.A. [Westinghouse Electric Company, Cranberry Township, Pennsylvania (United States)

    2012-07-01

    Feedwater control and turbine control have historically been at the top of the list of contributors to unplanned outages and forced curtailments in the nuclear industry, and they remain so according to recent industry data. Much has been done and is available by way of measures to improve this area and, in spite of much progress, opportunities remain to extend implementation. Toward this end, this paper aims to focus upon feedwater control and provide background on associated characteristics and attributes as a context for identifying the issues which are key challenges that lie at the root of this concern. Primary groupings of these issues will be discussed in order to better define their nature and to establish a basis for a presentation of the range of solutions which have been implemented and remain available to address them. The need for a systems engineering approach, and the role of I&C and field-mounted equipment to application of these solutions will be discussed. (author)

  4. BWR feedwater nozzle and control-rod-drive return line nozzle cracking

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    In its 1978 Annual Report to Congress, the Nuclear Regulatory Commission identified as an unresolved safety issue the appearance of cracks in feedwater nozzles at boiling-water reactors (BWRs). Later similar cracking, detected in return water lines for control-rod-drive systems at BWRs, was designated Part II of the issue. This article outlines the resolution of these cracking problems

  5. VGB conference 'Chemistry in the power plant 1984' - VGB feedwater conditioning conference

    International Nuclear Information System (INIS)

    1984-01-01

    The conference bears various aspects of feedwater conditioning for power plant cooling systems and steam generators as well as on the analytical assessment of water quality and its translation into operational method approaches. 5 out of the total 14 papers were entered separately in the database. (RB) [de

  6. Signal validation and failure correction algorithms for PWR steam generator feedwater control

    International Nuclear Information System (INIS)

    Nasrallah, C.N.; Graham, K.F.

    1986-01-01

    A critical contributor to the reliability of a nuclear power plant is the reliability of the control systems which maintain plant operating parameters within desired limits. The most difficult system to control in a PWR nuclear power plant and the one which causes the most reactor trips is the control of the feedwater flow to the steam generators. The level in the steam generator must be held within relatively narrow limits, with reactor trips set for both too high and too low a level. The steam generator level is inherently unstable in that it is an open integrator of feedwater flow steam flow mismatch. The steam generator feedwater control system relies on sensed variables in order to generate the appropriate feedwater valve control signal. In current systems, each of these sensed variables comes from a single sensor which may be a separate control sensor or one of the redundant protection sensors that is manually selected by the operator. In case this single signal is false, either due to sensor malfunction or due to a test signal being substituted during periodic test and maintenance, the control system will generate a wrong control signal to the feedwater control valve. This will initiate a steam generator level upset. The solution to this problem is for the control system to sense a given variable with more than one redundant sensor. Normally there are three or four sensors for each variable monitored by the reactor protection system. The techniques discussed allow the control system to compare these redundant sensor signals and generate a validated signal for each measured variable that is insensitive to false signals

  7. The role of main olfactory and vomeronasal systems in animal ...

    African Journals Online (AJOL)

    In many terrestrial tetrapod, olfactory sensory communication is mediated by two anatomically and functionally distinct sensory systems; the main olfactory system and vomeronasal system (accessory olfactory system). Recent anatomical studies of the central pathways of the olfactory and vomeronasal systems showed that ...

  8. Analysis of KNU1 loss of normal feedwater

    International Nuclear Information System (INIS)

    Kim, Hho-Jung; Chung, Bub-Dong; Lee, Young-Jin; Kim, Jin-Soo

    1986-01-01

    Simulation of the system thermal-hydraulic parameters was carried out following the KNU1 (Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on November 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS (Reactor Coolant system) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018. (author)

  9. Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

    Energy Technology Data Exchange (ETDEWEB)

    Nurkkala, P.; Hoikkanen, J. [Imatran Voima Oy, Vantaa (Finland)

    1997-12-31

    This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. `grounded` and `with goose neck`). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.). 3 refs.

  10. Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

    Energy Technology Data Exchange (ETDEWEB)

    Nurkkala, P; Hoikkanen, J [Imatran Voima Oy, Vantaa (Finland)

    1998-12-31

    This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. `grounded` and `with goose neck`). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.). 3 refs.

  11. 'Better feedwater quality through heat exchange equipment renovation'

    International Nuclear Information System (INIS)

    Pouzenc, C.

    2002-01-01

    In a fossil-fired or nuclear steam power plant, the water secondary circuit is a critical part of its thermodynamic cycle, as it achieves conditioning, pressurizing and heating of the condensate to match the conditions required at the steam generator inlet. Furthermore, the power plant electrical output and efficiency depend on availability and performances of each component of this secondary circuit from the condenser to the steam generator. Erosion and corrosion phenomena are at the origin of most significant failures in these components and related interconnecting systems. Feedwater chemistry is, together with the selection of materials and optimization of fluid velocities, one of the key levers to protect, as efficiently as possible, the components of the water secondary. (authors)

  12. Targeting the orexinergic system: Mainly but not only for sleep ...

    African Journals Online (AJOL)

    Orexin receptors belong to the big family of G protein coupled receptors (GPCRs) that constitute the main targets in the modern pharmacological approaches. Although the orexinergic system is involved in a variety of processes, treating sleep-wakefulness disorders such as narcolepsy and insomnia, remains the main ...

  13. Regional Ocean Modeling System (ROMS): Main Hawaiian Islands: Data Assimilating

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Regional Ocean Modeling System (ROMS) 3-day, 3-hourly data assimilating hindcast for the region surrounding the main Hawaiian islands at approximately 4-km...

  14. Regional Ocean Modeling System (ROMS): Main Hawaiian Islands

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Regional Ocean Modeling System (ROMS) 7-day, 3-hourly forecast for the region surrounding the main Hawaiian islands at approximately 4-km resolution. While...

  15. Feedwater heater tube-to-tubesheet connections

    International Nuclear Information System (INIS)

    Yokell, S.

    1993-01-01

    This paper discusses some practical aspects of expanded, welded, and welded-and-expanded feedwater heater tube-to-tubesheet joints. It outlines elastic-plastic tube expanding theory. It examines uniform-pressure-expanded tube joint strength and correlating roller-expanded joint strength with wall reduction and rolling torque. For materials subject to stress-corrosion cracking (SCC), it recommends heat treating tube ends before expanding. For materials subject to fatigue and tube-end cracking, it advocates two-stage expanding: (1) expanding enough to create firm tube-hole contact over the full tubesheet thickness; and (2) re-expanding at full pressure or torque. The paper emphasizes the desirability of segregating heats of tubing, mapping the tube-heat locations and making the heat map a permanent part of the heater maintenance file. It recommends when to provide TEMA/HEI Power Plant Standard annular grooves for roller-expanding and provides an equation for determining optimum groove width for uniform-pressure expanding. The paper also reviews welding requirements for welds of tubes to tubesheets. The review covers front-face welding before and after expanding and the reasons for welding first. It outlines current thinking about definitions of strength- and seal-welds of front-face welded joint in terms of their functions and load-carrying abilities. It presents a proposal for determining the required size of strength welds for use in Section VIII of the ASME Boiler and Pressure Vessel Code (the Code). It shows why welded-and-expanded feedwater heater tube-to-tubesheet joints should be full-strength and full-depth expanded. It makes recommendations for pressure- and leak-testing. This work also proposes the industry consider butt welding the tubes to the steam-side face of the tubesheet as a regular method of tube joining. The results of a survey of manufacturers practices are appended. 30 refs., 14 figs

  16. Development of main steam safety valve set pressure evaluating system

    International Nuclear Information System (INIS)

    Oketani, Koichiro; Manabe, Yoshihisa.

    1991-01-01

    A main steam safety valve set pressure test is conducted for all valves during every refueling outage in Japan's PWRs. Almost all operations of the test are manually conducted by a skilled worker. In order to obtain further reliability and reduce the test time, an automatic test system using a personnel computer has been developed in accordance with system concept. Quality assurance was investigated to fix system specifications. The prototype of the system was manufactured to confirm the system reliability. The results revealed that this system had high accuracy measurement and no adverse influence on the safety valve. This system was concluded to be applicable for actual use. (author)

  17. The main aspects of methodology of quality management system

    Directory of Open Access Journals (Sweden)

    Smirnova E.K.

    2017-03-01

    Full Text Available this article describes the formation and development of quality management as an integrated system. The author considers the theory and methodology of quality management since the early XXth century to the present day and describes the main problems encountered in the process of quality management system, as well as the ways to overcome them.

  18. 49 CFR 229.49 - Main reservoir system.

    Science.gov (United States)

    2010-10-01

    ... automatic air brake system shall be adjusted so that the compressor will start when the main reservoir..., DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Safety Requirements Brake System § 229.49... least one safety valve that shall prevent an accumulation of pressure of more than 15 pounds per square...

  19. Design of SMART alarm system using main memory database

    International Nuclear Information System (INIS)

    Jang, Kue Sook; Seo, Yong Seok; Park, Keun Oak; Lee, Jong Bok; Kim, Dong Hoon

    2001-01-01

    To achieve design goal of SMART alarm system, first of all we have to decide on how to handle and manage alarm information and how to use database. So this paper analyses concepts and deficiencies of main memory database applied in real time system. And this paper sets up structure and processing principles of main memory database using nonvolatile memory such as flash memory and develops recovery strategy and process board structures using these. Therefore this paper shows design of SMART alarm system is suited functions and requirements

  20. Integrated Main Propulsion System Performance Reconstruction Process/Models

    Science.gov (United States)

    Lopez, Eduardo; Elliott, Katie; Snell, Steven; Evans, Michael

    2013-01-01

    The Integrated Main Propulsion System (MPS) Performance Reconstruction process provides the MPS post-flight data files needed for postflight reporting to the project integration management and key customers to verify flight performance. This process/model was used as the baseline for the currently ongoing Space Launch System (SLS) work. The process utilizes several methodologies, including multiple software programs, to model integrated propulsion system performance through space shuttle ascent. It is used to evaluate integrated propulsion systems, including propellant tanks, feed systems, rocket engine, and pressurization systems performance throughout ascent based on flight pressure and temperature data. The latest revision incorporates new methods based on main engine power balance model updates to model higher mixture ratio operation at lower engine power levels.

  1. Open channel steam generator feedwater system

    International Nuclear Information System (INIS)

    Kim, R.F.; Min-Hsiung Hu.

    1985-01-01

    A steam generator which utilizes a primary fluid to vaporize a secondary fluid is provided with an open flow channel and elevated discharge nozzle for the introduction of secondary fluid. The discharge nozzle is positioned above a portion of the inlet line such that the secondary fluid passes through a vertical section of inlet line prior to its discharge into the open channel. (author)

  2. Feedwater control system in BWR type reactor

    International Nuclear Information System (INIS)

    Tanji, Jun-ichi; Oomori, Takashi.

    1980-01-01

    Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

  3. Results and analysis of a loss-of-feedwater induced ATWS experiment in the LOFT Facility

    International Nuclear Information System (INIS)

    Grush, W.H.; Koizumi, Y.; Woerth, S.C.

    1983-01-01

    An anticipated transient without scram (ATWS), initiated by a loss of feedwater, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a two-position actuator relief valve to simulate a scaled power-operated relief valve (PORV) and safety relief valve (SRV) representative of those in a commercial PWR. Auxiliary feedwater injection was delayed during the experiment until the plant recovery phase where long-term shutdown was achieved by an operator-controlled plant recovery procedure without inserting the control rods. The system transient response predicted by the RELAP5/MOD1 computer code showed good agreement with the experimental data

  4. Integrated TRAC/MELPROG analysis of core damage from a severe feedwater transient in the Oconee-1 PWR

    International Nuclear Information System (INIS)

    Henninger, R.J.; Boyack, B.E.

    1986-01-01

    A postulated complete loss-of-feedwater event in the Oconee-1 pressurized water reactor has been analyzed. With an initial version of the lonked TRAC and MELPROG codes, we have modeled the loss-of-feedwater event from initiation to the time of complete disruption of the core, which was calculated to occur by 6800 s. The highest structure temperatures otuside the vessel are on the flow path from the vessel to the pressurizer relief valve. Temperatures in excess of 1200 K could result in failure and depressurization of the primary system before vessel failure

  5. Automatic regulation of the feedwater turbo-pump capacity for the single-turbine 1000 MW NPP unit

    International Nuclear Information System (INIS)

    Pavlysh, O.N.; Garbuzov, I.P.; Reukov, Yu.N.

    1985-01-01

    A schematic of the flow regulators (FR) of feedwater turbo-pumps (FTP) for the single-turbine 1000 MW NPP unit is presented. The FR operate in response to feedwoter signals from FTP or in response to FTP rotor rotational speed and control automatic speed governars. The FR automatic regulation ensures limitation of FTP rotor maximum rotational speed at a feedwater flow rate excess equal to 3600 T/h. The transients in the automatic regulation system are considered. Production tests of FTP FR confirmed the FR operation reliability and the right choice of the regulator concept and structure

  6. Transient simulation of feedwater vaporization during a DBA LOP/LOCA using RELAP5/MOD3.1

    International Nuclear Information System (INIS)

    Harrell, J.R.; Fuller, R.W.

    1996-01-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station (GGNS) are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. The original design and testing requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. Given this condition, the appropriate testing criteria would be based on air with a relatively tight allowable limit. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leakage flow exists from the reactor vessel to the condenser through the feedwater piping during the reactor vessel blowdown phase. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves

  7. Loss-of-normal-feedwater sensitivity studies for AP600 behavior characterization

    International Nuclear Information System (INIS)

    Saiu, G.

    1996-01-01

    Activity concerning the development of a RELAP5/MOD3 model to simulate the Westinghouse Electric Corporation AP600 is summarized. The aim is to gain initial insight into the capability of RELAP5 to simulate the behavior of AP600 safety features. A-loss-of-normal-feedwater event is studied. Of the transients that must be investigated, this transient has been chosen to be one of the most relevant because the response of the AP600 to a loss-of-normal-feedwater event differs significantly from that of current pressurized water reactors in the extensive use of passive safety features peculiar to the AP600. Also, strong interactions among the AP600 safety systems, which should be further analyzed to permit full optimization of the system actuation logic and operation, are shown. Finally, a loss of normal feedwater without reactor scram, performed to investigate short-term plant behavior, shows that the pressure peak is affected by critical discharge flow coefficients applied to the pressurizer safety valves, while a relatively small reduction of the pressure peak is observed when both heat exchangers of the passive heat removal system are operating as opposed to the case in which only one is available. The data used for this study are derived from the Standard Safety Analysis Report configuration of the Westinghouse AP600 as of 1992

  8. Study on applicability of evaluation model of manpower needs for dismantling of equipments in FUGEN-1. Dismantling process in 3rd/4th feedwater heater room

    International Nuclear Information System (INIS)

    Shibahara, Yuji; Izumi, Masanori; Nanko, Takashi; Tachibana, Mitsuo; Ishigami, Tsutomu

    2010-10-01

    Manpower needs for the dismantling process on the dismantling of equipments in FUGEN 3rd/4th feedwater heater room was calculated with the management data evaluation system (PRODIA Code), and it was inspected whether the conventional evaluation model had applicability for FUGEN or not. It was confirmed that the conventional evaluation model for feedwater heater had no applicability. In comparison of the calculated value with the actual data, we found two difference: 1) the calculated value were significantly larger than the actual data, 2) the actual data for the dismantling of 3rd feedwater heater was twice larger than that of 4th feedwater heater, though these equipments were almost same weight. It was found that these were brought 1) by the difference in the work descriptions of dismantling between JPDR and FUGEN, and 2) by that in the cutting number between 3rd feedwater heater and 4th one. The manpower needs for the dismantling of both feedwater heaters were calculated with a new calculation equation reflecting the descriptions of dismantling, and it was found that these results showed the good agreement with the actual data. (author)

  9. Probabilistic safety analysis of the Kozloduy NPP units 1-4 (WWER-440/230) using independent emergency feedwater system; Veroyatnostnyj analiz bezopasnosti I-IV blokov AEhS `Kozloduy` s reaktorami tipa WWER-440 (V 230) pri vklyuchenii nezavisimoj sistemy avarijnoj podpitki PG

    Energy Technology Data Exchange (ETDEWEB)

    Kalchev, B; Marinov, M; Dimitrov, B; Avdzhiev, K [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    The safety of the Kozloduy NPP is being promoted by backfitting and improved operational practice. Special measures mitigating potential severe accidents consequences are needed because of some deficiencies in the original design of the four WWER-440 units. In conditions of a total LOCA (Loss Of Coolant Accident) it is impossible to ensure decay heat removal using the existing safety system. In such cases an extra emergency feedwater system independent of the plant`s other systems has been introduced which offers a new alternative means of removing the residual heat from the reactor. A probabilistic safety analysis is carried out using the method of event trees. A comparison between the existing safety system and the newly proposed is made. The simulation results of the unit behaviour prove that the damage frequency of the active zone is lower with the new system. 3 refs., 3 tabs., 2 figs.

  10. Experience with the TRIUMF Main Tank Vacuum Control System

    International Nuclear Information System (INIS)

    Sarkar, S.; Yandon, J.C.; Sievers, W.; Bennett, P.; Gurd, D.P.; Harmer, P.; Nelson, J.

    1993-01-01

    The TRIUMF Main Tank Vacuum Control System was upgraded in 1984. The earlier system, which consisted of a collection of hardwired relay logic boxes housed in three standard instrumentation racks, was replaced with a compact and flexible microprocessor-based control system. The user interface, previously distributed over the three racks, was consolidated into a single hardwired control and mimic panel. Since 1984, the Main Tank Vacuum System has undergone a series of changes in configuration and vacuum pumping hardware with necessary changes being implemented in the control system logic. Corresponding changes to the user interface were sometimes difficult to implement and in time exhausted the spare input/output capacity which had been built into the panel. The availability of inexpensive personal computers with adequate graphics capability and the ease of modifying, or adding to a programmable user interface precipitated the retirement of the hardwired panel and its replacement by a PC-based graphics user interface. System configuration, safety considerations, the hardware and the software implementation using the open-quote C close-quote programming language are described. The evolution of the control system and its performance, both over the years and in adapting to the vacuum system changes, are discussed

  11. Aging assessment of auxiliary feedwater pumps

    International Nuclear Information System (INIS)

    Greenstreet, W.L.

    1987-01-01

    ORNL is conducting aging assessments of auxiliary feedwater pumps to provide recommendations for monitoring and assessing the severity of time-dependent degradation as well as to recommend maintenance and replacement practices. Cornerstones of these activities are the identification of failure modes and causes and ranking of causes. Failure modes and causes of interest are those due to aging and service wear. Design details, functional requirements, and operating experience data were used to identify failure modes and causes and to rank the latter. Based on this input, potentially useful inspection, surveillance, and condition monitoring methods that are currently available for use or in the developmental stage were examined and recommendations made. The methods selected are listed and discussed in terms of use and information to be obtained. Relationships between inspection, surveillance, and monitoring and maintenance practices entered prominently into maintenance recommendations. These recommendations, therefore, embrace predictive as well as corrective and preventative maintenance practices. The recommendations are described, inspection details are discussed, and periodic inspection and maintenance interval guidelines are given. Surveillance testing at low-flow conditions is also discussed. It is shown that this type of testing can lead to accelerated aging

  12. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  13. Description of leakage monitoring system at Angra 2 nuclear power plant primary circuit

    International Nuclear Information System (INIS)

    Costa, Lilian Rose Sobral da; Mendes, Jorge Eduardo de Souza

    1999-01-01

    This paper describes the Leakage Monitoring System installed in Angra 2 NPP. This system has the task of detecting, localizing and quantifying leaks in systems for which rupture preclusion is cited. These systems include the reactor coolant pressure boundary, the main steam and feedwater lines within the containment, and the main steam safety and relief valve station in the valve annex. (author)

  14. Equipment reliability and life cycle optimization of a nuclear plant feedwater heater

    International Nuclear Information System (INIS)

    Thomas, Daniel; Coakley, Michael; Catapano, Michael; Svensson, Eric

    2006-01-01

    Many papers published over the last 25 years have strongly emphasized the need for an ongoing program of inspection and testing with subsequent failure cause analysis of feedwater heaters. Plants must be run more competitively; therefore, Utilities must lower operation and maintenance costs, while optimizing overall plant efficiency and capacity factor. One recognized area that needs to be addressed in accomplishing this goal is the heat cycle. This paper specifically deals with the feedwater heating system. Utility engineers must monitor feedwater heater performance in order to recognize degradation, identify and mitigate failure mechanisms, and prevent in-service failures thereby optimizing availability. Periodic tube plugging without complete analysis of the degraded/failed areas resolves the immediate need for return to service; however, heater life will not be optimized. This paper illustrates a complete life cycle management inspection, testing, and maintenance program implemented at Peach Bottom Atomic Power Station (PBAPS). Concerns that tubes may have been too conservatively plugged due to insufficient data and lack of root cause analysis, justified a program that included: - Removal of previously installed plugs; - Video-probe inspection of failed areas; - Extraction of tube samples for further analysis; - Eddy current testing of selected tubes; - Evaluation of the condition of 'insurance' plugged tubes for return to service; - Hydrostatic testing of selected individual tubes; - Final repair plan based on the results of the above program. This paper concludes that no single method of inspection or testing should solely be relied upon in establishing: - The extent of actual degraded conditions; - The mechanism(s) of failure; - The details of repair to be implemented. Evaluating all data affords the best chance in arresting problems and optimizing feedwater heater life. Problem heaters should be continuously monitored and inspected over time until the facts

  15. Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

    International Nuclear Information System (INIS)

    Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

    1984-01-01

    This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

  16. Monitoring the performance of Aux. Feedwater Pump using Smart Sensing Model

    Energy Technology Data Exchange (ETDEWEB)

    No, Young Gyu; Seong, Poong Hyun [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    Many artificial intelligence (AI) techniques equipped with learning systems have recently been proposed to monitor sensors and components in NPPs. Therefore, the objective of this study is the development of an integrity evaluation method for safety critical components such as Aux. feedwater pump, high pressure safety injection (HPSI) pump, etc. using smart sensing models based on AI techniques. In this work, the smart sensing model is developed at first to predict the performance of Aux. feedwater pump by estimating flowrate using group method of data handing (GMDH) method. If the performance prediction is achieved by this feasibility study, the smart sensing model will be applied to development of the integrity evaluation method for safety critical components. Also, the proposed algorithm for the performance prediction is verified by comparison with the simulation data of the MARS code for station blackout (SBO) events. In this study, the smart sensing model for the prediction performance of Aux. feedwater pump has been developed. In order to develop the smart sensing model, the GMDH algorithm is employed. The GMDH algorithm is the way to find a function that can well express a dependent variable from independent variables. This method uses a data structure similar to that of multiple regression models. The proposed GMDH model can accurately predict the performance of Aux.

  17. Monitoring the performance of Aux. Feedwater Pump using Smart Sensing Model

    International Nuclear Information System (INIS)

    No, Young Gyu; Seong, Poong Hyun

    2015-01-01

    Many artificial intelligence (AI) techniques equipped with learning systems have recently been proposed to monitor sensors and components in NPPs. Therefore, the objective of this study is the development of an integrity evaluation method for safety critical components such as Aux. feedwater pump, high pressure safety injection (HPSI) pump, etc. using smart sensing models based on AI techniques. In this work, the smart sensing model is developed at first to predict the performance of Aux. feedwater pump by estimating flowrate using group method of data handing (GMDH) method. If the performance prediction is achieved by this feasibility study, the smart sensing model will be applied to development of the integrity evaluation method for safety critical components. Also, the proposed algorithm for the performance prediction is verified by comparison with the simulation data of the MARS code for station blackout (SBO) events. In this study, the smart sensing model for the prediction performance of Aux. feedwater pump has been developed. In order to develop the smart sensing model, the GMDH algorithm is employed. The GMDH algorithm is the way to find a function that can well express a dependent variable from independent variables. This method uses a data structure similar to that of multiple regression models. The proposed GMDH model can accurately predict the performance of Aux

  18. Iron concentration controller in feedwater in nuclear plant

    International Nuclear Information System (INIS)

    Aizawa, Motohiro; Isaka, Yoshitaka

    1990-01-01

    The purpose of the present invention is to prevent chlorine ions from flowing into a reactor when sea water leakage accident should occur in a condenser upon control of Fe concentration in feedwater. That is, a sensor is disposed for detecting the leakage of the sea water at the exit of the condenser. The controller receives a detection signal as the input and delivers a control signal as the output. A control system receives the control signal and actuates valves in bypass systems. In view of the above, the electroconductivity or chlorine ion concentration of the condensate, which varies upon occurrence of sea water leakages in the condenser, is detected by the sensor, and then the controller closes a valve dispposed in the bypass systems in a processing device for filtering and desalting the condensates. Accordingly, the chlorine ions mixed into the condensates are removed by a desalting device without flowing into the reactor. In view of the above, an effect capable of keeping integrity of the plant is obtainable. (I.S.)

  19. Right main bronchial fracture resolution by digital thoracic drainage system.

    Science.gov (United States)

    Cortés Julián, Gildardo; Mier, José M; Iñiguez, Marco A; Guzmán de Alba, Enrique

    2016-03-01

    Tracheobronchial stenosis is common in the thoracic surgery service, and iatrogenic injury of the airway after manipulation is not infrequent. When a digital thoracic drainage system came onto the market, many advantages were evident. A 24-year-old woman with critical right main bronchial stenosis underwent airway dilation that was complicated by a tear with a massive air leak, resulting in a total right pneumothorax. We employed a pleural drain connected to a digital thoracic drainage system. The drain was removed 2 days after successful resolution of the air leak. © The Author(s) 2015.

  20. Operation safety of complex industrial systems. Main concepts

    International Nuclear Information System (INIS)

    Zwingelstein, G.

    2009-01-01

    Operation safety consists in knowing, evaluating, foreseeing, measuring and mastering the technological system and human failures in order to avoid their impacts on health and people's safety, on productivity, and on the environment, and to preserve the Earth's resources. This article recalls the main concepts of operation safety: 1 - evolutions in the domain; 2 - failures, missions and functions of a system and of its components: functional failure, missions and functions, industrial processes, notions of probability; 3 - basic concepts and operation safety: reliability, unreliability, failure density, failure rate, relations between them, availability, maintainability, safety. (J.S.)

  1. Nuclear plant power up-rate study: feedwater heater evaluations

    International Nuclear Information System (INIS)

    Svensson, Eric; Catapano, Michael; Coakley, Michael; Thomas, Dan

    2014-01-01

    Given today's nuclear industry business climate, it has become common for Utility companies to consider increasing unit capacities through turbine replacement and power up-rates. An integral part of the studies conducted by many towards this end, involve the generation of a set of turbine cycle heat balances with predicted performance parameters for the up-rated condition. Once these tentative operating values are established, it becomes necessary to evaluate the suitability of the existing components within each system to ensure they are capable of continued safe and reliable operation. The ultimate cost for the up-rate, including the cost for any major required modifications or significant replacements is weighed against increased revenue generated from the up-rate over time. Exelon's Peach Bottom Atomic Power Station (PBAPS) is currently planning for an Extended Power up-rate (EPU) for both units. To ensure the existing Feedwater Heaters (FWH) could maintain the operating and transient response margins at the EPU condition, an engineering study was conducted. Powerfect Inc. in conjunction with SPX Heat Transfer LLC were contracted to provide engineering services to analyze the design, thermal performance, reliability and operating conditions at projected EPU conditions. Specifically, to address the following with regard to the station's Feedwater Heaters (FWHs): 1. Evaluate Drain Cooler (DC) Velocities - including zone inlet velocity, cross and window velocities and outlet velocities. 2. Evaluate Drain Cooler Zone Pressure Drop for effect on drain cooler margins to flashing. 3. Evaluate differential pressure allowable across the pass partition plate. 4. Evaluate Drain Cooler Tube Vibration Potential. 5. Perform detailed steam dome velocity calculations. The goal of the study was to identify the most susceptible areas within the heaters for problems and potential failures when operating at the higher duty of the EPU condition for the remaining life

  2. Advanced Health Management System for the Space Shuttle Main Engine

    Science.gov (United States)

    Davidson, Matt; Stephens, John; Rodela, Chris

    2006-01-01

    Pratt & Whitney Rocketdyne, Inc., in cooperation with NASA-Marshall Space Flight Center (MSFC), has developed a new Advanced Health Management System (AHMS) controller for the Space Shuttle Main Engine (SSME) that will increase the probability of successfully placing the shuttle into the intended orbit and increase the safety of the Space Transportation System (STS) launches. The AHMS is an upgrade o the current Block II engine controller whose primary component is an improved vibration monitoring system called the Real-Time Vibration Monitoring System (RTVMS) that can effectively and reliably monitor the state of the high pressure turbomachinery and provide engine protection through a new synchronous vibration redline which enables engine shutdown if the vibration exceeds predetermined thresholds. The introduction of this system required improvements and modification to the Block II controller such as redesigning the Digital Computer Unit (DCU) memory and the Flight Accelerometer Safety Cut-Off System (FASCOS) circuitry, eliminating the existing memory retention batteries, installation of the Digital Signal Processor (DSP) technology, and installation of a High Speed Serial Interface (HSSI) with accompanying outside world connectors. Test stand hot-fire testing along with lab testing have verified successful implementation and is expected to reduce the probability of catastrophic engine failures during the shuttle ascent phase and improve safely by about 23% according to the Quantitative Risk Assessment System (QRAS), leading to a safer and more reliable SSME.

  3. Beam position measurement system at the Fermilab main accelerator

    International Nuclear Information System (INIS)

    Kerns, Q.A.

    1975-01-01

    The beam position system of the Fermilab Main Ring contains one horizontal and one vertical Electrostatic Beam Pickup in each of the 96 cells of the machine. A pair of 75 ohm cables transmits the induced signal from the machine tunnel to the nearest service building. In each of the 24 service buildings, there is a solid-state multiplexer and a beam position detector which processes the A-B signal pairs to produce an intensity-normalized voltage proportional to beam displacement. This voltage is digitized, read into buffer of the Lockheed MAC A, and in turn transferred to the Xerox 530. Horizontal or vertical orbits can be obtained in 50 millisec. Orbits are obtained at injection and at a Main Ring Sample time, if requested, anywhere on the acceleration cycle. Injection orbits can be flattened automatically by a program that sets dipole trim magnets. (auth)

  4. Investigation of modeling and simulation on a PWR power conversion system with RELAP5

    International Nuclear Information System (INIS)

    Rui Gao; Yang Yanhua; Lin Meng; Yuan Minghao; Xie Zhengrui

    2007-01-01

    Based on the power conversion system of nuclear and conventional islands of Dayabay nuclear power station, this paper models the thermal-hydraulic systems for PWR by using the best-estimate program, RELAP5. To simulate the full-scope power conversion system, not only the reactor coolant system (RCP) of nuclear island, but also the main steam system (VVP), turbine steam and drain system (GPV), bypass system (GCT), feedwater system (FW), condensate extraction system (CEX), moisture separator reheater system (GSS), turbine-driven feedwater pump (APP), low-pressure and high-pressure feedwater heater systems (ABP and AHP) of conventional island are considered and modeled. A comparison between the simulated results and the actual data of reactor under full-power demonstrates a fine match for Dayabay, and also manifests the feasibility in simulating full-scope power conversion system of PWR with RELAP5. (author)

  5. Dependence of steam generator vibrations on feedwater pressure

    International Nuclear Information System (INIS)

    Sadilek, J.

    1989-01-01

    Vibration sensors are attached to the bottom of the steam generator jacket between the input and output primary circuit collectors. The effective vibration value is recorded daily. Several times higher vibrations were observed at irregular intervals; their causes were sought, and the relation between the steam generator vibrations measured at the bottom of its vessel and the feedwater pressure was established. The source of the vibrations was found to be in the feedwater tract of the steam generator. The feedwater tract is described and its hydraulic characteristics are given. Vibrations were measured on the S02 valve. It is concluded that vibrations can be eliminated by reducing the water pressure before the control valves and by replacing the control valves with ones with more suitable control characteristics. (E.J.). 3 figs., 1 tab., 3 refs

  6. Main Aspects of the Rodos System Implementation in Croatia

    International Nuclear Information System (INIS)

    Sucic, B.; Medakovic, S.; Skanata, D.; Tomisa, T.

    2002-01-01

    Analyses made in Europe after the accident in Chernobyl Nuclear Power Plant have shown that the key-element with the biggest impact on the development of consequences during and immediately after the accident was deficient notifying which was also unsystematic and in disorder. Therefore, scientific and expert organisations of the European Union have been faced with the task of system development above all meant to improve communication between European countries in terms of measured parameters exchange (the radiological and meteorological ones), that would be generally applicable in all the European countries. Development of such a system going by the name of RODOS (Real Time On Line DecisiOn Support) began in 1989 within the framework of the European Commission's research and technological development program. The main aspects associated with the implementation of the RODOS system in Croatia are described within this article. By the main aspects of implementation we mean technical prerequisites that the Republic of Croatia is due to fulfil before joining in, and they include activities like: (1) translating users' interfaces into Croatian, (2) adapting various models in consistence with local parameters, (3) collecting geographical data, (4) making specific local maps and collecting statistical data (GIS), and (5) linking the RODOS real-time data base with the net of local measure stations (radiological and meteorological). Activities of this scale and technically so demanding, request certain organisational prerequisites as well as considerable material resources. From the point of view of the Republic of Croatia, establishment of the RODOS system and inclusion into the real-time international exchange of measured radiological data would mean major improvement of the nuclear emergency response preparedness system. (author)

  7. Classification of Feedwater Heater Performance Degradation Using Residual Sign Matrix

    International Nuclear Information System (INIS)

    Ha, Gayeon; Heo, Gyunyoung; Song, Seok Yoon

    2016-01-01

    Since a performance of Feedwater Heater (FWH) is directly related to the thermodynamic efficiency of Nuclear Power Plants (NPPs), performance degradation of FWH results in loss of thermal power and ultimately business benefit. Nevertheless, it is difficult to diagnose its degradation of performance during normal operation due to its minor changes in process parameters, for instance, pressure, temperature, and flowrate. In this paper, six degradation modes have been analyzed and the performance indices for FWH such as Terminal Temperature Difference (TTD) and Drain Cooling Approach (DCA) have been used to diagnose degradation modes. PEPSE (Performance Evaluation of Power System Efficiencies) simulation, which is a plant simulation software simulating plant static characteristic and building energy balance model, has been used to generate the data of performance indices of FWH and actual measurements of FWH from NPPs was used to validate the classification model. In this paper, six degradation modes have been analyzed and the performance indices for FWH have been used to diagnose what degradation mode occurs. The RSM was proposed as a trend identifier of variables. Using RSM, it is possible to obtain appropriate information of the variables in noise environment since noise can be compressed while the original information is being converted to a trend. The SVC has been performed to classify the degradation mode of FWH, and then actual measurements of FWH from NPPs was used to validate the classification model. Performance indices under various leakage conditions show different patterns. In further study, tube leakage simulations for the various cases will be needed

  8. Classification of Feedwater Heater Performance Degradation Using Residual Sign Matrix

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Gayeon; Heo, Gyunyoung [Kyung Hee University, Seoul (Korea, Republic of); Song, Seok Yoon [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    Since a performance of Feedwater Heater (FWH) is directly related to the thermodynamic efficiency of Nuclear Power Plants (NPPs), performance degradation of FWH results in loss of thermal power and ultimately business benefit. Nevertheless, it is difficult to diagnose its degradation of performance during normal operation due to its minor changes in process parameters, for instance, pressure, temperature, and flowrate. In this paper, six degradation modes have been analyzed and the performance indices for FWH such as Terminal Temperature Difference (TTD) and Drain Cooling Approach (DCA) have been used to diagnose degradation modes. PEPSE (Performance Evaluation of Power System Efficiencies) simulation, which is a plant simulation software simulating plant static characteristic and building energy balance model, has been used to generate the data of performance indices of FWH and actual measurements of FWH from NPPs was used to validate the classification model. In this paper, six degradation modes have been analyzed and the performance indices for FWH have been used to diagnose what degradation mode occurs. The RSM was proposed as a trend identifier of variables. Using RSM, it is possible to obtain appropriate information of the variables in noise environment since noise can be compressed while the original information is being converted to a trend. The SVC has been performed to classify the degradation mode of FWH, and then actual measurements of FWH from NPPs was used to validate the classification model. Performance indices under various leakage conditions show different patterns. In further study, tube leakage simulations for the various cases will be needed.

  9. The Livingstone Model of a Main Propulsion System

    Science.gov (United States)

    Bajwa, Anupa; Sweet, Adam; Korsmeyer, David (Technical Monitor)

    2003-01-01

    Livingstone is a discrete, propositional logic-based inference engine that has been used for diagnosis of physical systems. We present a component-based model of a Main Propulsion System (MPS) and say how it is used with Livingstone (L2) in order to implement a diagnostic system for integrated vehicle health management (IVHM) for the Propulsion IVHM Technology Experiment (PITEX). We start by discussing the process of conceptualizing such a model. We describe graphical tools that facilitated the generation of the model. The model is composed of components (which map onto physical components), connections between components and constraints. A component is specified by variables, with a set of discrete, qualitative values for each variable in its local nominal and failure modes. For each mode, the model specifies the component's behavior and transitions. We describe the MPS components' nominal and fault modes and associated Livingstone variables and data structures. Given this model, and observed external commands and observations from the system, Livingstone tracks the state of the MPS over discrete time-steps by choosing trajectories that are consistent with observations. We briefly discuss how the compiled model fits into the overall PITEX architecture. Finally we summarize our modeling experience, discuss advantages and disadvantages of our approach, and suggest enhancements to the modeling process.

  10. Limbic system, the main focus of dementia syndrome

    International Nuclear Information System (INIS)

    Matsuzawa, Taiju

    1990-01-01

    Alzheimer disease and multi-infarct dementia are two entirely different diseases producing almost the same abnormalities as dementia syndrome. The statistical studies with MRI to locate the focus of dementia syndrome in the neocortex was an absolute failure. With MRI there is drastic atrophy and destruction of the amygdala and hippocampus suggesting the limbic system as the focus of dementia syndrome. Destruction of the limbic system in particular amygdala and hippocampus produced the functional obstruction brought about by the marked reduction in the glucose utilization with PET in the bilateral temporal, parietal and occipital association cortices. Although this type constitutes only about 1/5 of all dementia patients. It is considered the fundamental type of dementia syndrome. Aside from this, there is a type wherein simultaneous and symmetrical reductions in glucose utilization of the frontal association cortex and the motor association cortex in the anterior part of the neocortex. This is referred to as type II. It constitutes about 4/5 of all dementia patients which is far more than type I. Based on these results, it is thought that limbic system is the main focus of dementia syndrome. (author)

  11. Ultrasonic pattern recognition study of feedwater nozzle inner radius indication

    International Nuclear Information System (INIS)

    Yoneyama, H.; Takama, S.; Kishigami, M.; Sasahara, T.; Ando, H.

    1983-01-01

    A study was made to distinguish defects on feed-water nozzle inner radius from noise echo caused by stainless steel cladding by using ultrasonic pattern recognition method with frequency analysis technique. Experiment has been successfully performed on flat clad plates and nozzle mock-up containing fatigue cracks and the following results which shows the high capability of frequency analysis technique are obtained

  12. The Main Asteroid Belt: The Crossroads of the Solar System

    Science.gov (United States)

    Michel, Patrick

    2015-08-01

    Orbiting the Sun between Mars and Jupiter, main belt asteroids are leftover planetary building blocks that never accreted enough material to become planets. They are therefore keys to understanding how the Solar System formed and evolved. They may also provide clues to the origin of life, as similar bodies may have delivered organics and water to the early Earth.Strong associations between asteroids and meteorites emerged thanks to multi-technique observations, modeling, in situ and sample return analyses. Spacecraft images revolutionized our knowledge of these small worlds. Asteroids are stunning in their diversity in terms of physical properties. Their gravity varies by more orders of magnitude than its variation among the terrestrial planets, including the Moon. Each rendezvous with an asteroid thus turned our geological understanding on its head as each asteroid is affected in different ways by a variety of processes such as landslides, faulting, and impact cratering. Composition also varies, from ice-rich to lunar-like to chondritic.Nearly every asteroid we see today, whether of primitive or evolved compositions, is the product of a complex history involving accretion and one or more episodes of catastrophic disruption that sometimes resulted in families of smaller asteroids that have distinct and indicative petrogenic relationships. These families provide the best data to study the impact disruption process at scales far larger than those accessible in laboratory. Tens, perhaps hundreds, of early asteroids grew large enough to thermally differentiate. Their traces are scattered pieces of their metal-rich cores and, more rarely, their mantles and crusts.Asteroids represent stages on the rocky road to planet formation. They have great stories to tell about the formation and evolution of our Solar System as well as other planetary systems: asteroid belts seem common around Sun-like stars. We will review our current knowledge on their properties, their link to

  13. Beam monitors and transverse feedback system of TRISTAN Main Ring

    International Nuclear Information System (INIS)

    Ieiri, T.; Ishii, H.; Kishiro, J.; Mizumachi, Y.; Mori, K.; Nakajima, K.; Ogata, A.; Shintake, T.; Tejima, M.

    1987-01-01

    The construction of 30 GeV TRISTAN Main Ring (MR) started in 1983 soon after the commissioning of 8 GeV Accumulation Ring (AR). The authors prepared 392 position monitors, 6 synchrotron radiation monitors, 9 screen monitors, 2 DCCT's, 3 scrapers, 12 bunch monitors, transverse feedback systems for two beams and DC separators. Since the required monitoring devices of AR and MR are almost the same, the experiences in AR were very useful in the design of MR monitors. However, machine parameters of two rings are very different and the authors had to review the performance of each item. From the monitor point of view the most important is the difference of revolution frequency; 794.6 kHz for AR and 99.33 kHz for MR. This means that average beam current of MR is 1/8 as small as AR current with the same bunch number and intensity. Therefore, the sensitivity of each monitor must be better in MR. The second difference is that MR should be used as a collider from the beginning. Therefore they must prepare for multi-beam and multi-bunch operation

  14. Removal of Iron Oxide Scale from Feed-water in Thermal Power Plant by Using Magnetic Separation

    Science.gov (United States)

    Nakanishi, Motohiro; Shibatani, Saori; Mishima, Fumihito; Akiyama, Yoko; Nishijima, Shigehiro

    2017-09-01

    One of the factors of deterioration in thermal power generation efficiency is adhesion of the scale to inner wall in feed-water system. Though thermal power plants have employed All Volatile Treatment (AVT) or Oxygen Treatment (OT) to prevent scale formation, these treatments cannot prevent it completely. In order to remove iron oxide scale, we proposed magnetic separation system using solenoidal superconducting magnet. Magnetic separation efficiency is influenced by component and morphology of scale which changes their property depending on the type of water treatment and temperature. In this study, we estimated component and morphology of iron oxide scale at each equipment in the feed-water system by analyzing simulated scale generated in the pressure vessel at 320 K to 550 K. Based on the results, we considered installation sites of the magnetic separation system.

  15. Application of a Long Term Asset Management Strategy for HP Feedwater Heaters

    International Nuclear Information System (INIS)

    Won, Se Youl; Yun, Eun Sub; Park, Young Sheop

    2008-01-01

    As the commercial operating year of nuclear power plants is increased, it becomes imperative to develop integrated cost-effective asset management and to improve plans for degraded Structures, Systems, and Components (SSCs) in terms of safety and economical consideration. A long-term asset management (LTAM) strategy can improve the condition of nuclear plants, maximize their value, and optimize their operational life by maintaining their safety. This paper presents an optimized LTAM plan for HP feedwater heaters at a specific nuclear power plant

  16. Influence of the loop design of the feedwater- and steam quality in a power plant with pressurized water reactor

    International Nuclear Information System (INIS)

    Bennert, J.; Becher, L.

    1977-01-01

    At nuclear power plants with pressurized water reactors, condensate occurs on the high pressure part of the water-steam circuit, caused by the operation with low steam parameters. The behaviour of the electrolytes which entered into the circuit (solubility, distribution in water and/or steam) shows that these electrolytes (salts) are to be found mainly in the condensate. The insinuated electrolytes are reconcentrated during the common arrangements with 'Small Circuit' - consisting of steam generator, high pressure turbine, water separator, feedwater vessel, and have a negative influence on the feedwater - boiler water - and the steam quality. Remedy is possible by modified arrangements, during which these electrolyte-containing condensates will be treated and traced back into the main circuit. Nevertheless that the efficiency decrease is insignificant and additional efforts are necessary, a change over to these arrangements is recommendable, due to the fact that the feedwater quality, the boiler water quality, the steam quality in front of the turbine, and finally also the operational safety, as well as the availability will be improved. (orig.) [de

  17. Study of the reliability of the Auxiliary Feedwater System of a LWR nuclear power plant through the Fault Tree and Bayesian Network; Estudo de confiabilidade do Sistema Auxiliar de Agua de Alimentacao de uma central nuclear a agua leve por arvore de falhas e rede Bayesiana

    Energy Technology Data Exchange (ETDEWEB)

    Lava, Deise Diana

    2016-10-01

    This paper aims to present a study of the reliability of the Auxiliary Feedwater System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10{sup -3}. (author)

  18. Application of neural networks to validation of feedwater flow rate in a nuclear power plant

    International Nuclear Information System (INIS)

    Khadem, M.; Ipakchi, A.; Alexandro, F.J.; Colley, R.W.

    1993-01-01

    Feedwater flow rate measurement in nuclear power plants requires periodic calibration. This is due to the fact that the venturi surface condition of the feedwater flow rate sensor changes because of a chemical reaction between the surface coating material and the feedwater. Fouling of the venturi surface, due to this chemical reaction and the deposits of foreign materials, has been observed shortly after a clean venturi is put in operation. A fouled venturi causes an incorrect measurement of feedwater flow rate, which in turn results in an inaccurate calculation of the generated power. This paper presents two methods for verifying incipient and continuing fouling of the venturi of the feedwater flow rate sensors. Both methods are based on the use of a set of dissimilar process variables dynamically related to the feedwater flow rate variable. The first method uses a neural network to generate estimates of the feedwater flow rate readings. Agreement, within a given tolerance, of the feedwater flow rate instrument reading, and the corresponding neural network output establishes that the feedwater flow rate instrument is operating properly. The second method is similar to the first method except that the neural network predicts the core power which is calculated from measurements on the primary loop, rather than the feedwater flow rates. This core power is referred to the primary core power in this paper. A comparison of the power calculated from the feedwater flow measurements in the secondary loop, with the calculated and neural network predicted primary core power provides information from which it can be determined whether fouling is beginning to occur. The two methods were tested using data from the feedwater flow meters in the two feedwater flow loops of the TMI-1 nuclear power plant

  19. Pressure waves transient occurred in the steam generators feedwater lines of the Atucha-1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Balino, J.L.; Carrica, P.M.; Larreteguy, A.E.

    1993-01-01

    The pressure transient occurred at Atucha I Nuclear Power Plant in March 1990 is simulated. The transient was due to the fast closure of a flow control valve at the steam generators feedwater lines. The system was modelled, including the actuation of the relief valves. The minimum closure time for no actuation of the relief valves and the evolution of the velocity and piezo metric head for different cases were calculated. (author)

  20. Ethanolamine properties and use for feedwater pH control: A pressurized water reactor case study

    International Nuclear Information System (INIS)

    Keeling, D.L.; Polidoroff, C.T.; Cortese, S.; Cushner, M.C.

    1995-01-01

    Ethanolamine (ETA) as a feedwater pH control additive has been recently used to minimize corrosion of secondary water components in the nuclear power industry pressurized water reactors (PWRs). The use of ETA is compared with ammonia. Relative volatility effects on various parts of the system are analyzed and chemistry changes are presented. Materials of construction and the use of existing plant equipment for ETA service are discussed. Properties of ETA as well as safety, storage and handling issues are compared with ammonia. Health d aquatic toxicity are reviewed. warnings, safety, handling guidelines, biodegradability an Diablo Canyon Power Plant used ammonia for pH control from 1985 until a change over to ETA in 1993/1994. Full flow condensate polishers that are required to protect the plant from saltwater cooling incursions limit the amount of pH additive. Iron levels in the secondary water systems are compared before and after changing to ETA and replacement of corrosion-susceptible piping. Iron reduction benefits are assessed along with other effects on the feedwater nozzles, low pressure turbine, polisher resin capacity and polisher regeneration system

  1. Optimization algorithms intended for self-tuning feedwater heater model

    International Nuclear Information System (INIS)

    Czop, P; Barszcz, T; Bednarz, J

    2013-01-01

    This work presents a self-tuning feedwater heater model. This work continues the work on first-principle gray-box methodology applied to diagnostics and condition assessment of power plant components. The objective of this work is to review and benchmark the optimization algorithms regarding the time required to achieve the best model fit to operational power plant data. The paper recommends the most effective algorithm to be used in the model adjustment process.

  2. European Reliability Data System: main developments and use

    International Nuclear Information System (INIS)

    Amesz, J.; Capobianchi, S.; Kalfsbeek, H.W.; Mancini, G.

    1985-01-01

    The paper revises the aims and objectives of the European Reliability Data System (ERDS), a centralized system collecting and organizing information related to the operation of light water reactors. The paper first describes the recent developments of the four data banks constituting the system: Component Event Data Bank, Abnormal Occurrences Reporting System, Operating Unit Status Report and Reliability Parameters Data Bank. Then several issues are discussed referring mostly to the status of classification schemes and their use, to the operation of the banks (data input and transcoding) and to the retrieval and utilization of the information; in this latter case particularly the analysis potential of the data collection scheme of the AORS (Abnormal Occurrences Reporting System) is demonstrated. Finally, emphasis is given to the increasing role which artificial intelligence techniques such as natural language and expert systems and fuzzy logic may play in improving the future capabilities of the system

  3. Changes in feedwater organic matter concentrations based on intake type and pretreatment processes at SWRO facilities, Red Sea, Saudi Arabia

    KAUST Repository

    Dehwah, Abdullah

    2015-03-01

    Transparent exopolymer particles (TEP), natural organic matter, and bacterial concentrations in feedwater are important factors that can lead to membrane biofouling in seawater reverse osmosis (SWRO) systems. Two methods for controlling these concentrations in the feedwater prior to pretreatment have been suggested; use of subsurface intake systems or placement of the intake at a greater depth in the sea. These proposed solutions were tested at two SWRO facilities located along the Red Sea of Saudi Arabia. A shallow well intake system was very effective in reducing the algae and bacterial concentrations and somewhat effective in reducing TEP concentrations. An intake placed at a depth of 9. m below the surface was found to have limited impact on improving water quality compared to a surface intake. The algae and bacteria concentration in the feedwater (deep) was lower compared to the surface seawater, but the overall TEP concentration was higher. Bacteria and TEP measurements made in the pretreatment process train in the plant and after the cartridge filters suggest that regrowth of bacteria is occurring within the cartridge filters.

  4. Changes in feedwater organic matter concentrations based on intake type and pretreatment processes at SWRO facilities, Red Sea, Saudi Arabia

    KAUST Repository

    Dehwah, Abdullah; Li, Sheng; Almashharawi, Samir; Winters, Harvey; Missimer, Thomas M.

    2015-01-01

    Transparent exopolymer particles (TEP), natural organic matter, and bacterial concentrations in feedwater are important factors that can lead to membrane biofouling in seawater reverse osmosis (SWRO) systems. Two methods for controlling these concentrations in the feedwater prior to pretreatment have been suggested; use of subsurface intake systems or placement of the intake at a greater depth in the sea. These proposed solutions were tested at two SWRO facilities located along the Red Sea of Saudi Arabia. A shallow well intake system was very effective in reducing the algae and bacterial concentrations and somewhat effective in reducing TEP concentrations. An intake placed at a depth of 9. m below the surface was found to have limited impact on improving water quality compared to a surface intake. The algae and bacteria concentration in the feedwater (deep) was lower compared to the surface seawater, but the overall TEP concentration was higher. Bacteria and TEP measurements made in the pretreatment process train in the plant and after the cartridge filters suggest that regrowth of bacteria is occurring within the cartridge filters.

  5. Analisis Termal High Pressure Feedwater Heater di PLTU PT. XYZ

    Directory of Open Access Journals (Sweden)

    Maria Ulfa Damayanti

    2017-01-01

    Full Text Available Abstrak- PT. XYZ mengoperasikan tiga unit Pembangkit Listrik Tenaga Uap (PLTU unit 3, 7 dan 8 berkapasitas 2.030 MegaWatt. Pada PLTU Paiton unit 7 dan 8 terdapat delapan buah feedwater heater yaitu empat buah Low Pressure Water Heater (LPWH, tiga buah High Pressure Water Heater (HPWH, dan sebuah dearator. Pada PLTU Paiton unit 7 dan 8 terdapat kerusakan pada HPWH 6 yang menyebabkan penurunan efisiensi dari siklus secara keseluruhan. Penurunan efisiensi dapat terjadi karena temperatur feedwater sebelum masuk ke boiler terlalu rendah, sehingga kalor yang dibutuhkan oleh boiler untuk memanaskan feedwater meningkat. Oleh karena itu konsumsi batubara akan meningkat dan menyebabkan terjadi kenaikan biaya operasional harian dalam sistem pembangkit. Dari data Divisi Produksi PT. XYZ Unit 7 dan 8 diperoleh spesifikasi HPWH 6, 7, dan 8 dan propertis fluida dalam HPWH 6, 7, dan 8. Data tersebut digunakan sebagai dasar analisis termal yang meliputi performa masing-masing HPH. Tahap selanjutnya dalam analisis termal adalah memvariasikan beban 25%, 50%, 75%, 100%, dan 105%. Tahap terakhir analisis adalah menghitung performa dengan variasi sumbatan (plug 5%, 10%, 15%, dan 20% sesuai dengan variasi beban. Hasil yang didapatkan dari penelitian tugas akhir ini adalah nilai effectiveness tertinggi tercapai pada pembebanan 100% serta menghasilkan pressure drop tertinggi pada pembebanan 105%, nilai effectiveness terbesar serta nilai pressure drop terkecil terjadi pada zona Condensing, serta sumbatan (plugging pada HPH akan menyebabkan penurunan nilai effectiveness dan kenaikan pressure drop sisi tube.

  6. Structure of the main control system for the FTU device

    International Nuclear Information System (INIS)

    Panaccione, L.

    1983-01-01

    The computer based control system for FTU (Frascati Tokamak Upgrade) machine is a distributed intelligence control system consisting of various subsystems, called ''specific units''. A specific unit can work as a stand-alone small system or as a part of more complex systems, together with other units, in any possible combination. In the case of combined units working mode, a dedicated specific unit performs supervisory functions. Communication between specific units is done on a dedicated bus-like network, in the form of temporary master-slave relationship. There are three groups of specific units: Plant Control Units, Supervisory Control Unit and Synoptic Handling Unit. Performances and structures of these specific units are described. In addition, a very simple hard-wired safety system works in parallel with the computer based control system, in order to save personnel and apparatus. (author)

  7. Embedded expert system for space shuttle main engine maintenance

    Science.gov (United States)

    Pooley, J.; Thompson, W.; Homsley, T.; Teoh, W.; Jones, J.; Lewallen, P.

    1987-01-01

    The SPARTA Embedded Expert System (SEES) is an intelligent health monitoring system that directs analysis by placing confidence factors on possible engine status and then recommends a course of action to an engineer or engine controller. The technique can prevent catastropic failures or costly rocket engine down time because of false alarms. Further, the SEES has potential as an on-board flight monitor for reusable rocket engine systems. The SEES methodology synergistically integrates vibration analysis, pattern recognition and communications theory techniques with an artificial intelligence technique - the Embedded Expert System (EES).

  8. ATWS analysis for total loss of feedwater sequence in UCN 3 and 4

    International Nuclear Information System (INIS)

    Park, S. H.; Song, Y. M.; Kim, D. H.; Kim, S. D.; Park, S. Y.

    1999-01-01

    ATWS is a trip-failed severe accident initiated from the transients like a turbine trip, a control bank withdrawal, and a loss of feedwater which are expected to occur comparatively often (one or two occurrences / year). In this study, an ATWS sequence in Ulchin 3 and 4 is analyzed and the effects of the important systems are studied for accident management purpose using a MIDAS/PK computer code. The MIDAS/PK code has been developed via coupling a point kinetics module with the MELCOR code. The code calculates a primary peak pressure of about 24MPa at 240 seconds for the ATWS initiated by a TLOF (Total Loss of Feedwater) transient. Along with the basic ATWS analysis, several sensitivity runs are performed. From these, the turbines and the safety depressurization system (SDS) are judged to be important. The turbine trip resulting in a loss of offsite power and a RCP trip, degrades primary heat transfer to the secondary sides, and in turn, increases primary coolant temperature which reduces the reactor power due to the negative moderator temperature coefficient. Manual operation of SDS has an effect to lower the primary peak pressure considerably via supplementary depressurization in addition to the PORVs

  9. Main features and potentialities of gas-blanket systems

    International Nuclear Information System (INIS)

    Lehnert, B.

    1977-02-01

    A review is given of the features and potentialities of cold-blanket systems, with respect to plasma equilibrium, stability, and reactor technology. The treatment is concentrated on quasi-steady magnetized plasmas confined at moderately high beta values. The cold-blanket concept has specific potentialities as a fusion reactor, e.g. in connection with the desired densities and dimensions of full-scale systems, refuelling, as well as ash and impurity removal, and stability. (author)

  10. Development of methodology for evaluating and monitoring steam generator feedwater nozzle cracking in PWRs

    International Nuclear Information System (INIS)

    Shvarts, S.; Gerber, D.A.; House, K.; Hirschberg, P.

    1994-01-01

    The objective of this paper is to describe a methodology for evaluating and monitoring steam generator feedwater nozzle cracking in PWR plants. This methodology is based in part on plant test data obtained from a recent Diablo Canyon Power Plant (DCPP) Unit 1 heatup. Temperature sensors installed near the nozzle-to-pipe weld were monitored during the heatup, along with operational parameters such as auxiliary feedwater (AFW) flow rate and steam generator temperature. A thermal stratification load definition was developed from this data. Steady state characteristics of this data were used in a finite element analysis to develop relationship between AFW flow and stratification interface level. Fluctuating characteristics of this data were used to determine transient parameters through the application of a Green's Function approach. The thermal stratification load definition from the test data was used in a three-dimensional thermal stress analysis to determine stress cycling and consequent fatigue damage or crack growth during AFW flow fluctuations. The implementation of the developed methodology in the DCPP and Sequoyah Nuclear Plant (SNP) fatigue monitoring systems is described

  11. Commissioning of the vacuum system of the KATRIN Main Spectrometer

    Czech Academy of Sciences Publication Activity Database

    Arenz, M.; Dragoun, Otokar; Kovalík, Alojz; Lebeda, Ondřej; Ryšavý, Miloš; Sentkerestiová, Jana; Slezák, Martin; Špalek, Antonín; Vénos, Drahoslav; Zbořil, Miroslav

    2016-01-01

    Roč. 11, APR (2016), P04011 ISSN 1748-0221 R&D Projects: GA ČR(CZ) GAP203/12/1896; GA MŠk LG14004 Institutional support: RVO:61389005 Keywords : gas systems and purification * neutriono detectors * spectrometers Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.220, year: 2016

  12. Validated design of the ITER main vacuum pumping systems

    International Nuclear Information System (INIS)

    Day, Chr.; Antipenkov, A.; Dremel, M.; Haas, H.; Hauer, V.; Mack, A.; Boissin, J.-C.; Class, G.; Murdoch, D.K.; Wykes, M.

    2005-01-01

    Forschungszentrum Karlsruhe is developing the ITER high vacuum cryogenic pumping systems (torus, cryostat, NBI) as well as the corresponding mechanical roughing pump trains. All force-cooled big cryopumps incorporate similar design of charcoal coated cryopanels cooled to 5 K with supercritical helium. A model of the torus exhaust cryopump was comprehensively characterised in the TIMO testbed at Forschungszentrum. This paper discusses the vacuum performance results of the model pump and outlines how these data were incorporated in a sound design of the whole ITER torus exhaust pumping system. To do this, the dedicated software package ITERVAC was developed which is able to describe gas flow in viscous, transitional and molecular flow regimes as needed for the gas coming through the divertor slots and along the pump ducts into the cryopumps. The entrance section between the divertor cassettes and each pumping duct was identified to be the bottleneck of the gas flow. The interrelation of achievable throughputs as a function of the divertor pressure and the cryopump pumping speed is discussed. The system design is completed by assessment of the NBI cryopump system and integrating performance curves for the roughing pump trains needed during the regeneration phases of the cryopumps. (author)

  13. The AGS Booster main ring power supply system

    International Nuclear Information System (INIS)

    Soukas, A.; Hughes, K.; Sandberg, J.; Toldo, F.; Zhang, S.Y.

    1989-01-01

    The AGS Booster is being designed as a very versatile particle accelerator. Its primary function is to be a high quality injector to the currently operating Alternating Gradient Synchrotron (AGS). The Booster/AGS combination will produce proton intensities greater than 5 x 10 13 protons per pulse (ppp), and accelerate heavy ions, with mass up to 200, to a maximum energy of 15 GeV per atomic mass unit (GeV/amu). The power supply for the Booster Main Ring (BMRPS) has to accommodate a wide range of cycles and a wide range of operating parameters. The cycles range from storage for several seconds to rapid cycling at 7.5 Hz. The peak output power is 18 MW. This paper will describe the AGS Booster machine powering requirements, the choice of power supply, the a.c. circuit tie-in and its associated problems and some of the details of the design of the BMRPS. 9 refs., 2 figs

  14. TASKA-M exhaust system and its main components

    International Nuclear Information System (INIS)

    Kleefeldt, K.W.; Mueller, R.A.; Schramm, K.

    1985-01-01

    TASKA-M is a study for a mirror based D-T plasma device for fusion technology tests. Mature technology was applied whereever possible. The axial confinement time is relatively short, resulting in a large gas throughput compared to the fusion power level of 6.8 MW. The technological requirements of the exhaust system will not cause undue development problems in either of the two major areas: highly loaded dumps for the conversion of the escaping particle and plasma streams to thermal gas; vacuum pumping facilities. (orig.)

  15. Space shuttle main propulsion pressurization system probabilistic risk assessment

    International Nuclear Information System (INIS)

    Plastiras, J.K.

    1989-01-01

    This paper reports that, in post-Challenger discussions with Congressional Committees and the National Research Council Risk Management Oversight Panel, criticism was levied against NASA because of the inability to prioritize the 1300+ single point failures. In the absence of a ranking it was difficult to determine where special effort was needed in failure evaluation, in design improvement, in management review of problems, and in flight readiness reviews. The belief was that the management system was overwhelmed by the quantity of critical hardware items that were on the Critical Items List (CIL) and that insufficient attention was paid to the items that required it. Congressional staff members from Congressman Markey's committee who have oversight responsibilities in the nuclear industry, and specifically over the nuclear power supplies for NASA's Galileo and Ulysses missions, felt very strongly that the addition of Probabilistic Risk Assessment (PRA) to the existing Failure Mode Effects Analysis/Hazard Analysis (FMEA/HA) methods was exceedingly important. Specifically, the Markey committee recognized that the inductive, qualitative component-oriented FMEA could be supplemented by the deductive, quantitative systems-oriented PRA. Furthermore, they felt that the PRA approach had matured to the extent that it could be used to assess risk, even with limited shuttle-specific failure data. NASA responded with arguments that the FMEA/HA had illuminated all significant failure modes satisfactorily and that no failure rate data base was available to support the PRA approach

  16. On-line validation of feedwater flow rate in nuclear power plants using neural networks

    International Nuclear Information System (INIS)

    Khadem, M.; Ipakchi, A.; Alexandro, F.J.; Colley, R.W.

    1994-01-01

    On-line calibration of feedwater flow rate measurement in nuclear power plants provides a continuous realistic value of feedwater flow rate. It also reduces the manpower required for periodic calibration needed due to the fouling and defouling of the venturi meter surface condition. This paper presents a method for on-line validation of feedwater flow rate in nuclear power plants. The method is an improvement of the previously developed method which is based on the use of a set of process variables dynamically related to the feedwater flow rate. The online measurements of this set of variables are used as inputs to a neural network to obtain an estimate of the feedwater flow rate reading. The difference between the on-line feedwater flow rate reading, and the neural network estimate establishes whether there is a need to apply a correction factor to the feedwater flow rate measurement for calculation of the actual reactor power. The method was applied to the feedwater flow meters in the two feedwater flow loops of the TMI-1 nuclear power plant. The venturi meters used for flow measurements are susceptible to frequent fouling that degrades their measurement accuracy. The fouling effects can cause an inaccuracy of up to 3% relative error in feedwater flow rate reading. A neural network, whose inputs were the readings of a set of reference instruments, was designed to predict both feedwater flow rates simultaneously. A multi-layer feedforward neural network employing the backpropagation algorithm was used. A number of neural network training tests were performed to obtain an optimum filtering technique of the input/output data of the neural networks. The result of the selection of the filtering technique was confirmed by numerous Fast Fourier Transform (FFT) tests. Training and testing were done on data from TMI-1 nuclear power plant. The results show that the neural network can predict the correct flow rates with an absolute relative error of less than 2%

  17. Main copper-porphyry systems of the Lesser Caucasus

    International Nuclear Information System (INIS)

    Melkonyan, R.L.; Tayan, P.N.; Goukassyan, R.Kh.; Hovakimyan, S.E.; Moritz, R.; Selbi, D.

    2013-01-01

    Two belts of porphyry-copper systems were identified the Late Jurassic Early Cretaceous Somkheto-Karabakh (S-K) island-arc belt within the same name terrain of the southern termination of the Eurasian Plate stretching for 230 km (the tonalitic model) and the Early Miocene Tsaghkounk-Zanghezour (Ts-Z) post-collision belt (Tz-Z) within the same name terrain of the northern margin of the Gondwana, stretching over 280 km (the monzonite-granodiorite model). The formation of the S-K and Ts-Z belts had proceeded in pulses and spanned intervals of 12 million years and 24 million years, respectively. The Rb-Sr isochrones and TIMS U-Pb estimations of the age of zircons from the Meghri pluton ( 1,500 km 2 ), the largest one in the Lesser Caucasus, it appeared possible to establish the three stages of its formation: the Late Eocene, Early Oligocene, and Early Miocene, each accompanied by development of deposits having similar ages. The PC deposits of the S-K and Ts-Z belts have distinct differences of age, geodynamic regime of formation, specificity of mineral composition, sources of water and sulfur of hydrothermal solutions, and formation models. The single, discrete Armenian-Iranian belt of PC deposits was identified; it has a Late Eocene-Middle Miocene age and a length of about 2,000 km, being related with intrusive complexes of the monzonite-granite-granodiorite series, the activity of which had been manifesting itself over 32 million years. This belt, including giant-deposits such as Kajaran and Sar-Cheshmeh, was identified as the special Armenian-Iranian PC province

  18. Prediction of main factors’ values of air transportation system safety based on system dynamics

    Science.gov (United States)

    Spiridonov, A. Yu; Rezchikov, A. F.; Kushnikov, V. A.; Ivashchenko, V. A.; Bogomolov, A. S.; Filimonyuk, L. Yu; Dolinina, O. N.; Kushnikova, E. V.; Shulga, T. E.; Tverdokhlebov, V. A.; Kushnikov, O. V.; Fominykh, D. S.

    2018-05-01

    On the basis of the system-dynamic approach [1-8], a set of models has been developed that makes it possible to analyse and predict the values of the main safety indicators for the operation of aviation transport systems.

  19. Fuzzy Logic Approach to Diagnosis of Feedwater Heater Performance Degradation

    International Nuclear Information System (INIS)

    Kang, Yeon Kwan; Kim, Hyeon Min; Heo, Gyun Young; Sang, Seok Yoon

    2014-01-01

    Since failure in, damage to, and performance degradation of power generation components in operation under harsh environment of high pressure and high temperature may cause both economic and human loss at power plants, highly reliable operation and control of these components are necessary. Therefore, a systematic method of diagnosing the condition of these components in its early stages is required. There have been many researches related to the diagnosis of these components, but our group developed an approach using a regression model and diagnosis table, specializing in diagnosis relating to thermal efficiency degradation of power plant. However, there was a difficulty in applying the method using the regression model to power plants with different operating conditions because the model was sensitive to value. In case of the method that uses diagnosis table, it was difficult to find the level at which each performance degradation factor had an effect on the components. Therefore, fuzzy logic was introduced in order to diagnose performance degradation using both qualitative and quantitative results obtained from the components' operation data. The model makes performance degradation assessment using various performance degradation variables according to the input rule constructed based on fuzzy logic. The purpose of the model is to help the operator diagnose performance degradation of components of power plants. This paper makes an analysis of power plant feedwater heater by using fuzzy logic. Feedwater heater is one of the core components that regulate life-cycle of a power plant. Performance degradation has a direct effect on power generation efficiency. It is not easy to observe performance degradation of feedwater heater. However, on the other hand, troubles such as tube leakage may bring simultaneous damage to the tube bundle and therefore it is the object of concern in economic aspect. This study explains the process of diagnosing and verifying typical

  20. Fuzzy Logic Approach to Diagnosis of Feedwater Heater Performance Degradation

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Yeon Kwan; Kim, Hyeon Min; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Sang, Seok Yoon [Engineering and Technical Center, Korea Hydro, Daejeon (Korea, Republic of)

    2014-08-15

    Since failure in, damage to, and performance degradation of power generation components in operation under harsh environment of high pressure and high temperature may cause both economic and human loss at power plants, highly reliable operation and control of these components are necessary. Therefore, a systematic method of diagnosing the condition of these components in its early stages is required. There have been many researches related to the diagnosis of these components, but our group developed an approach using a regression model and diagnosis table, specializing in diagnosis relating to thermal efficiency degradation of power plant. However, there was a difficulty in applying the method using the regression model to power plants with different operating conditions because the model was sensitive to value. In case of the method that uses diagnosis table, it was difficult to find the level at which each performance degradation factor had an effect on the components. Therefore, fuzzy logic was introduced in order to diagnose performance degradation using both qualitative and quantitative results obtained from the components' operation data. The model makes performance degradation assessment using various performance degradation variables according to the input rule constructed based on fuzzy logic. The purpose of the model is to help the operator diagnose performance degradation of components of power plants. This paper makes an analysis of power plant feedwater heater by using fuzzy logic. Feedwater heater is one of the core components that regulate life-cycle of a power plant. Performance degradation has a direct effect on power generation efficiency. It is not easy to observe performance degradation of feedwater heater. However, on the other hand, troubles such as tube leakage may bring simultaneous damage to the tube bundle and therefore it is the object of concern in economic aspect. This study explains the process of diagnosing and verifying typical

  1. An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

    International Nuclear Information System (INIS)

    Hall, D.G.; Watkins, J.C.

    1987-01-01

    This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

  2. Analysis of nuclear piping system seismic tests with conventional and energy absorbing supports

    International Nuclear Information System (INIS)

    Park, Y.; DeGrassi, G.; Hofmayer, C.; Bezler, P.; Chokshi, N.

    1997-01-01

    Large-scale models of main steam and feedwater piping systems were tested on the shaking table by the Nuclear Power Engineering Cooperation (NUPEC) of Japan, as part of the Seismic Proving Test Program. This paper describes the linear and nonlinear analyses performed by NRC/BNL and compares the results to the test data

  3. Inferential smart sensing for feedwater flowrate in PWRs

    International Nuclear Information System (INIS)

    Na, M. G.; Hwang, I. J.; Lee, Y. J.

    2006-01-01

    The feedwater flowrate that is measured by Venturi flow meters in most pressurized water reactors can be over-measured because of the fouling phenomena that make corrosion products accumulate in the Venturi meters. Therefore, in this work, two kinds of methods, a support vector regression method and a fuzzy modeling method, combined with a sequential probability ratio test, are used in order to accurately estimate online the feedwater flowrate, and also to monitor the status of the existing hardware sensors. Also, the data for training the support vector machines and the fuzzy model are selected by using a subtractive clustering scheme to use informative data from among all acquired data. The proposed inferential sensing and monitoring algorithm is verified by using the acquired real plant data of Yonggwang Nuclear Power Plant Unit 3. In the simulations, it was known that the root mean squared error and the relative maximum error are so small and the proposed method early detects the degradation of an existing hardware sensor. (authors)

  4. Analysis Of Feedwater Line Break Of APR1400 By MARS Code

    International Nuclear Information System (INIS)

    Nguyen Thi Thanh Thuy; Le Dai Dien, Hoang Minh Giang

    2011-01-01

    This paper will deal with analysis of Feed water Line Break problem (FWLB) of the APR 1400 NPP with initial conditions: operation at 100% of power, double-ended break area of 0.058 m 2 and the break location of the feedwater line between the check valve and the steam generator. The analysis was simulated by MARS code through two step: calculation for steady state and calculation for transient state with initial condition mentioned. Some output result were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as temperature, pressure, steam generator water levels as well as DNBR, etc. before and after the accident. (author)

  5. The Concept of Steam Pressure Control by Changing the Feedwater Flow during Heatup Operation for an Integral Reactor with a Once-Through Steam Generator

    International Nuclear Information System (INIS)

    Seo, Jae Kwang; Choi, Ki Yong; Kang, Han Ok; Kim, Young In; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    The design features of a once-through steam generator (OTSG) for an integral reactor are significantly different from the commercial U-tube type steam generator from several aspects such as the general arrangement, size, operation conditions, and so on. Therefore a sufficient understanding of the thermal-hydraulic characteristics of the OTSG is essential for the design of the nuclear steam supply system (NSSS) and the power conversion system (PCS). It is also necessary to develop operation procedures complying to the unique design features of the OTSG of interest. The OTSG is sized to produce a sufficiently superheated steam during a normal power operation and therefore the secondary system can be simple relative to that of the other types of steam generators. For the plant adopting the OTSG, the steam pressure in the secondary circuit (tube side of the OTSG) is controlled to be constant during a normal power operation. Constant steam pressure is realized by regulating the control valve on the main steam line dedicated for this purpose. However during a heatup operation, at which the fluid state at the exit of the OTSG is a single phase hot water or two phases, it is not proper to use the control valve on the main steam line due to a control problem at low and multi-phase flow conditions and possibly an erosion problem. For these reasons, another dedicated line called a startup cooling line is used during a heatup condition. There may be several operational conditions for the secondary fluid required to pass through during heatup operation, depending on the design of the PCS. In general, there are two conditions: One is a condition for a vacuum operation for the condenser and another is an entry condition for a steam pressure control operation for an auxiliary power system. In this study, the concept of using a simple startup cooling line with a fixed flow resistance and changing the feedwater flow for the pressure control of the PCS during a heatup period are

  6. Heat exchanger inventory cost optimization for power cycles with one feedwater heater

    International Nuclear Information System (INIS)

    Qureshi, Bilal Ahmed; Antar, Mohamed A.; Zubair, Syed M.

    2014-01-01

    Highlights: • Cost optimization of heat exchanger inventory in power cycles is investigated. • Analysis for an endoreversible power cycle with an open feedwater heater is shown. • Different constraints on the power cycle are investigated. • The constant heat addition scenario resulted in the lowest value of the cost function. - Abstract: Cost optimization of heat exchanger inventory in power cycles with one open feedwater heater is undertaken. In this regard, thermoeconomic analysis for an endoreversible power cycle with an open feedwater heater is shown. The scenarios of constant heat rejection and addition rates, power as well as rate of heat transfer in the open feedwater heater are studied. All cost functions displayed minima with respect to the high-side absolute temperature ratio (θ 1 ). In this case, the effect of the Carnot temperature ratio (Φ 1 ), absolute temperature ratio (ξ) and the phase-change absolute temperature ratio for the feedwater heater (Φ 2 ) are qualitatively the same. Furthermore, the constant heat addition scenario resulted in the lowest value of the cost function. For variation of all cost functions, the smaller the value of the phase-change absolute temperature ratio for the feedwater heater (Φ 2 ), lower the cost at the minima. As feedwater heater to hot end unit cost ratio decreases, the minimum total conductance required increases

  7. A connection of the steam generator feedwater section of WWER type nuclear power plants

    International Nuclear Information System (INIS)

    Matal, O.; Sadilek, J.

    1989-01-01

    In the feedwater piping of each steam generator, a plate for additional water pressure reduction is inserted before the first closing valve. During a steady water flow, the plate gives rise to a constant hydraulic resistance, bringing about steady reduction of the feedwater pressure; this also contributes to a stabilization of the feedwater flow rate into the steam generator. The control valve thus is stressed by minimal hydrodynamic forces. In this manner its load is decreased, its vibrations are damped, and the frequency of failures - and thereby the frequency of the nuclear power plant unit outages -is reduced. (J.P.). 1 fig

  8. Leak Detection of High Pressure Feedwater Heater Using Empirical Models

    International Nuclear Information System (INIS)

    Lee, Song Kyu; Kim, Eun Kee; Heo, Gyun Young; An, Sang Ha

    2009-01-01

    Even small leak from tube side or pass partition within the high pressure feedwater heater (HPFWH) causes a significant deficiency in its performance. Plant operation under the HPFWH leak condition for long time will result in cost increase. Tube side leak within HPFWH can produce the high velocity jet of water and it can cause neighboring tube failures. However, most of plants are being operated without any information for internal leaks of HPFWH, even though it is prone to be damaged under high temperature and high pressure operating conditions. Leaks from tubes and/or pass partition of HPFWH occurred in many nuclear power plants, for example, Mihama PS-2, Takahama PS-2 and Point Beach Nuclear Plant Unit 1. If the internal leaks of HPFWH are monitored, the cost can be reduced by inexpensive repairs relative to loss in performance and moreover plant shutdown as well as further tube damages can be prevented

  9. A decision theoretic approach to an accident sequence: when feedwater and auxiliary feedwater fail in a nuclear power plant

    International Nuclear Information System (INIS)

    Svenson, Ola

    1998-01-01

    This study applies a decision theoretic perspective on a severe accident management sequence in a processing industry. The sequence contains loss of feedwater and auxiliary feedwater in a boiling water nuclear reactor (BWR), which necessitates manual depressurization of the reactor pressure vessel to enable low pressure cooling of the core. The sequence is fast and is a major contributor to core damage in probabilistic risk analyses (PRAs) of this kind of plant. The management of the sequence also includes important, difficult and fast human decision making. The decision theoretic perspective, which is applied to a Swedish ABB-type reactor, stresses the roles played by uncertainties about plant state, consequences of different actions and goals during the management of a severe accident sequence. Based on a theoretical analysis and empirical simulator data the human error probabilities in the PRA for the plant are considered to be too small. Recommendations for how to improve safety are given and they include full automation of the sequence, improved operator training, and/or actions to assist the operators' decision making through reduction of uncertainties, for example, concerning water/steam level for sufficient cooling, time remaining before insufficient cooling level in the tank is reached and organizational cost-benefit evaluations of the events following a false alarm depressurization as well as the events following a successful depressurization at different points in time. Finally, it is pointed out that the approach exemplified in this study is applicable to any accident scenario which includes difficult human decision making with conflicting goals, uncertain information and with very serious consequences

  10. Manpower development for safe operation of nuclear power plant. China. Main steam bypass system operation and maintenance. Task: 6.1.6. Technical report

    International Nuclear Information System (INIS)

    Stubley, P.H.

    1994-01-01

    This mission concentrated on the Steam Bypass system of Qinshan Nuclear Power Plant. The system had experienced spurious opening of the bypass valves, disrupting the steam pressure control and the steam generator level control system. A series of commissioning type tests were defined which should allow the operators to revise the setpoints used in the control of the bypass system, and thus prevent spurious opening while maintaining the desired steam pressure control during power maneuvering. Training also included giving experience from other operating plants on aspects of steam and feedwater systems and components, especially as this experience affected maintenance or gave rise to problems. Steam generated maintenance experience is especially applicable, and a future mission is planned for an expert in this field. In addition other aspects of the Chinese nuclear program was assessed to guide future missions. This included assessment of operating procedures from an availability point of view

  11. Service Quality: A Main Determinant Factor for Health Information System Success in Low-resource Settings.

    Science.gov (United States)

    Tilahun, Binyam; Fritz, Fleur

    2015-01-01

    With the increasing implementation of different health information systems in developing countries, there is a growing need to measure the main determinants of their success. The results of this evaluation study on the determinants of HIS success in five low resource setting hospitals show that service quality is the main determinant factor for information system success in those kind of settings.

  12. Simulation and analysis of main steam control system based on heat transfer calculation

    Science.gov (United States)

    Huang, Zhenqun; Li, Ruyan; Feng, Zhongbao; Wang, Songhan; Li, Wenbo; Cheng, Jiwei; Jin, Yingai

    2018-05-01

    In this paper, after thermal power plant 300MW boiler was studied, mat lab was used to write calculation program about heat transfer process between the main steam and boiler flue gas and amount of water was calculated to ensure the main steam temperature keeping in target temperature. Then heat transfer calculation program was introduced into Simulink simulation platform based on control system multiple models switching and heat transfer calculation. The results show that multiple models switching control system based on heat transfer calculation not only overcome the large inertia of main stream temperature, a large hysteresis characteristic of main stream temperature, but also adapted to the boiler load changing.

  13. Main propulsion system design recommendations for an advanced Orbit Transfer Vehicle

    Science.gov (United States)

    Redd, L.

    1985-01-01

    Various main propulsion system configurations of an advanced OTV are evaluated with respect to the probability of nonindependent failures, i.e., engine failures that disable the entire main propulsion system. Analysis of the life-cycle cost (LCC) indicates that LCC is sensitive to the main propulsion system reliability, vehicle dry weight, and propellant cost; it is relatively insensitive to the number of missions/overhaul, failures per mission, and EVA and IVA cost. In conclusion, two or three engines are recommended in view of their highest reliability, minimum life-cycle cost, and fail operational/fail safe capability.

  14. Optimization of the main control room habitability system in nuclear power plant

    International Nuclear Information System (INIS)

    Zheng Guanghui; Zhao Xinyan

    2013-01-01

    This article describes the optimization of main control room habitability system in nuclear power plant. It also describes the design shortage in terms of habitability in the main control room. Through modification and optimization, habitable conditions are met for personnel staying in the emergency area of the main control room for a period of time, with an aim to take accident intervention measures smoothly and reduce the accident loss and radioactive contamination as low as possible. (authors)

  15. Experiment Operating Specification for the Semiscale MOD-2C feedwater and steam line break experiment series. Appendix S-FS-6 and 7

    International Nuclear Information System (INIS)

    Boucher, T.J.; Owca, W.A.

    1985-05-01

    This document is the Semiscale MOD-2C feedwater and steam line break experiment series Experiment Operating Specification Appendix for tests S-FS-6 and S-FS-7. Test S-FS-6 is the third test in the series and simulates a 100% break in a steam generator bottom feedwater line downstream of the check valve accompanied by compounding factors (such as check valve failure, loss-of-offsite power at SIS and SIS delayed until low steam generator pressure signal). The test is terminated after plant stabilization and recovery procedures including unaffected loop steam and feed, pressurizer heater operation, pressurizer auxiliary spray operation, and normal charging/letdown operation. Test S-FS-7 is the fourth test in the series and simulates a 14.3% break in a steam generator bottom feedwater line downstream of the check valve, accompanied by compounding factors. The test is terminated after plant stabilization procedures including unaffected loop steam and feed, pressurizer heater operation, and normal charging/letdown operation. The test was followed by an affected loop secondary refill after isolating the break. The Appendix contains information on the major fluid systems, initial experiment conditions, experiment boundary conditions, and sequence of experiment events. Also included is a discussion of the scaling criteria and philosophy used to develop the experiment initial and boundary conditions and system configuration

  16. Modern technical diagnostic system for the main components of powerful turbine generator

    International Nuclear Information System (INIS)

    Ezovit, G.P.; Uglyarenko, V.P.; Burlaka, S.I.; Goroz, N.I.; Orinin, S.E.; Komaritsa, V.N.; Zav'yalov, D.N.; Mazurenko, O.A.

    2011-01-01

    The modern diagnostic system to monitor the technical state of a powerful turbine generator is considered. This system permits the detection of defects in its main components and cooling system at the early stage of their development, prevention of damage and, as a consequence, emergency shutdown of nuclear power units

  17. Cleaning the feed-water pipeline internal surfaces

    International Nuclear Information System (INIS)

    Podkopaev, V.A.

    1984-01-01

    The procedure of cleaning the feed-water pipeline internal surfaces at the Chernobylsk-4 power unit is described. Cleaning was conducted in five stages. Pipelines were cleaned from mechanical impurities at the first stage. At the second stage the pipelines were washing by water heated up to 80 deg C. At the third stage nitric acid was added to 95-100 deg C water the acid concentration in the circuit = 60 mg/l, purification period = 14 h. At the fourth stage hydrogen peroxide was added to the circuit at 95-100 deg C (the solution concentration was equal to 5-6 mg/l, the solution stayed in the circuit for 1 h 20 min). At the fifth stage sodium nitrite concentrated to 20 mg/l was introduced to the circuit in 75 minutes; this promoted strengthening of the oxide layer in the circuit on the base of nitric acid and hydrogen peroxide. Data on the water acidity in the circuit, water electric conductivity and iron concentration after the fourth stage and on completion of the circuit cleaning are presented. The described method of cleaning enables to save scarce reagents and use cheaper ones

  18. Cleaning the feed-water pipeline internal surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Podkopaev, V.A.

    1984-12-01

    The procedure of cleaning the feed-water pipeline internal surfaces at the Chernobylsk-4 power unit is described. Cleaning was conducted in five stages. Pipelines were cleaned from mechanical impurities at the first stage. At the second stage the pipelines were washed by water heated up to 80 deg C. At the third stage nitric acid was added to 95-100 deg C water with the acid concentration in the circuit = 60 mg/l, purification period = 14 h. At the fourth stage hydrogen peroxide was added to the circuit at 95-100 deg C (the solution concentration was equal to 5-6 mg/l, the solution stayed in the circuit for 1 h 20 min). At the fifth stage sodium nitrite concentrated to 20 mg/l was introduced to the circuit in 75 minutes; this promoted strengthening of the oxide layer in the circuit on the base of nitric acid and hydrogen peroxide. Data on the water acidity in the circuit, water electric conductivity and iron concentration after the fourth stage and on completion of the circuit cleaning are presented. The described method of cleaning enables to save scarce reagents and use cheaper ones.

  19. The simulation of control system for the KEK main ring power supply, 1

    International Nuclear Information System (INIS)

    Sato, Hikaru; Shintomi, Takakazu; Kubo, Tadashi; Masuda, Masayoshi; Nakano, Michio.

    1975-09-01

    The simulation of control system for the KEK main ring power supply is performed. The inductance of the magnet is assumed to be linear against to the current. The results of the simulations are described. (auth.)

  20. Main approaches to automation of management systems in the coal industry. [Czechoslovakia

    Energy Technology Data Exchange (ETDEWEB)

    Zafouk, P; Dlabaya, I; Frous, S

    1980-01-01

    The main approaches to automation of management systems in the coal industry of Czechoslovakia are enumerated. Organizational structure of the branch and concern form of organization. Complex improvement of management system and source of continued development of the branch. Automated control systems, an integral part of the complex management system. Primary problem - automation in the area of design of the information system. Centralization of methodological management of operations in the area of control system development. Unified approach to breakdown of control system into branches. Organizational support of the development of the control system, problems solved by the department of control system development of the Ministry, main department of control system development of the Research Institute, departmental committees in the branch. The use of principles of control system development in the Ostravsko-Karvinsk mining concern is demonstrated. Preparation for development of the control system in the concern: elaboration of concepts and programs of control system development. Design of control system of the concern. Control system of an enterprise in the concern as an integral control system. Support of control system development in organizations, participants in this process, their jurisdiction and obligations. Annual plans of control system development. Centralized subsystems and enterprises. Methods of coordination of the process of improvement of control and support of the harmony of decisions made. Technical support of control system development, construction of a unified network of computer centers in enterprises with combined resources of computer technology.

  1. Ultrasonic meters in the feedwater flow to recover thermal power in the reactor of nuclear power plant of Laguna Verde U1 and U2

    International Nuclear Information System (INIS)

    Tijerina S, F.

    2008-01-01

    The engineers in nuclear power plants BWRs and PWRs based on the development of the ultrasonic technology for the measurement of the mass, volumetric flow, density and temperature in fluids, have applied this technology in two primary targets approved by the NRC: the use for the recovery of thermal power in the reactor and/or to be able to realize an increase of thermal power licensed in a 2% (MUR) by 1OCFR50 Appendix K. The present article mentions the current problem in the measurement of the feedwater flow with Venturi meters, which affects that the thermal balance of reactor BWRs or PWRs this underestimated. One in broad strokes describes the application of the ultrasonic technology for the ultrasonic measurement in the flow of the feedwater system of the reactor and power to recover thermal power of the reactor. One is to the methodology developed in CFE for a calibration of the temperature transmitters of RTD's and the methodology for a calibration of the venturi flow transmitters using ultrasonic measurement. Are show the measurements in the feedwater of reactor of the temperature with RTD's and ultrasonic measurement, as well as the flow with the venturi and the ultrasonic measurement operating the reactor to the 100% of nominal thermal power, before and after the calibration of the temperature transmitters and flow. Finally, is a plan to be able to realize a recovery of thermal power of the reactor, showing as carrying out their estimations. As a result of the application of ultrasonic technology in the feedwater of reactor BWR-5 in Laguna Verde, in the Unit 1 cycle 13 it was recover an equivalent energy to a thermal power of 25 MWt in the reactor and an exit electrical power of 6 M We in the turbogenerator. Also in the Unit 2 cycle 10 it was recover an equivalent energy to a thermal power of 40 MWt in the reactor and an exit electrical power of 16 M We in the turbogenerator. (Author)

  2. Method of controlling the operation of a feedwater system

    International Nuclear Information System (INIS)

    Takemaru, Koichi; Omori, Takashi.

    1975-01-01

    Object: At the time of pump trip at upstream, to maintain a good operating characteristic without stopping a pump at the downstream. Structure: In the event that one of pumps at the upstream (for example, a condenser pump) is tripped, interlock is activated to forcibly open a circulating valve for a condensation booster pump to partly return the flow rate of discharge to a pump suction part, thus preventing a decrease in suction pressure. In this case, in order to control suction pressure to a suitable level of pressure more than a net suction water head as required, a pressure adjusting meter for the condensation booster pump is provided on a pipe line on the side of suction so that the flow rate of circulation may be changed by the pressure adjusting valve to maintain the suction pressure in constant. (Kawakami, Y.)

  3. 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.)

  4. 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.)

  5. The upgraded data acquisition system for beam loss monitoring at the Fermilab Tevatron and Main Injector

    International Nuclear Information System (INIS)

    Baumbaugh, A.; Briegel, C.; Brown, B.C.; Capista, D.; Drennan, C.; Fellenz, B.; Knickerbocker, K.; Lewis, J.D.; Marchionni, A.; Needles, C.; Olson, M.

    2011-01-01

    A VME-based data acquisition system for beam-loss monitors has been developed and is in use in the Tevatron and Main Injector accelerators at the Fermilab complex. The need for enhanced beam-loss protection when the Tevatron is operating in collider-mode was the main driving force for the new design. Prior to the implementation of the present system, the beam-loss monitor system was disabled during collider operation and protection of the Tevatron magnets relied on the quench protection system. The new Beam-Loss Monitor system allows appropriate abort logic and thresholds to be set over the full set of collider operating conditions. The system also records a history of beam-loss data prior to a beam-abort event for post-abort analysis. Installation of the Main Injector system occurred in the fall of 2006 and the Tevatron system in the summer of 2007. Both systems were fully operation by the summer of 2008. In this paper we report on the overall system design, provide a description of its normal operation, and show a number of examples of its use in both the Main Injector and Tevatron.

  6. Main maintenance operations for Test Blanket Systems in ITER TBM port cells

    Energy Technology Data Exchange (ETDEWEB)

    Pascal, R., E-mail: romain.pascal@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Cortes, P.; Friconneau, J.-P.; Giancarli, L.M.; Gotewal, K.K.; Iseli, M.; Kim, B.Y.; Levesy, B.; Martins, J.-P.; Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Nevière, J.-C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Patisson, L. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Siarras, A. [Sogetti, Parc de la Duranne, 13857 Aix-en-Provence (France); Tesini, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: • The Test Blanket System components layout in Port Cell room is described. • The maintenance of the two Test Blanket Systems in ITER port cell is addressed. • The overall replacement/maintenance strategy is defined. • The main maintenance tasks of the systems are discussed. • The maintenance strategy and required tools are presented. -- Abstract: Each Test Blanket System in ITER is formed by an in-vessel component, the Test Blanket Module, and several associated ancillary systems (coolant and Tritium systems, instrumentation and control systems). The paper describes the overall replacement/maintenance strategy and the main maintenance tasks that have to be considered in the design of the systems. It shows that there are no critical issues.

  7. The hydraulic lifting system for the main magnets of CYCIAE-30 cyclotron

    CERN Document Server

    Zhao Zhen Lu; Chen Rong Fan; Chu Cheng Jie

    2002-01-01

    The oil-line structure, control system and their working principles of the hydraulic lifting system for the main magnets of CYCIAE-30 cyclotron are introduced. The six years practice proves that the specification of the system matches the requirements: the oil cylinder maximum stroke of 850 mm, the eight slot positioning dowels repositioning accuracy of +-0.01 mm, the two oil cylinders moving in step accuracy of 5-10 mm. The system is safe, reliable and easy to be operated

  8. Main Directions in Ensuring Business Continuity for Information and Telecommunication Systems of High Availability

    Directory of Open Access Journals (Sweden)

    Boris Mikhailovich Egorov

    2014-12-01

    Full Text Available Enterprise Continuity Program, Information and Telecommunication System of Bank of Russia The results of the analysis of the main directions to ensure business continuity for the information and telecommunication systems of high availability in relation to expansion of the problems range, their intellectualization and the modern IT implementation are given.

  9. Validation study on reliability analysis of main safety system in Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Cho, Chang Keun; Kim, Yong Hui; Kim, Tae Hyeong; Hong, Seo Kee; Park, Keon Woo; Park, Chang Jea [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Cheong, Woo Sik [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Moon Kyu [KEPRI, Taejon (Korea, Republic of)

    1993-12-15

    The scope and contents of this validation study are to review the design changes of the four main safety systems in Wolsong 2/3/4 Nuclear Power Plants, to review the consideration of the above design changes in the AECL reports, the structure of fault trees, and the data base used in the quantification of the fault trees, to quantify the unavailabilities of main safety systems and check them if they meet the requirements, and to recommend desirable design changes in the emergency core cooling system to reduce the unavailability.

  10. Flexibility analysis of main primary heat transport system : Narora Atomic Power Project

    International Nuclear Information System (INIS)

    Rastogi, S.K.

    1975-01-01

    The paper presents flexibility analysis problem of main primary heat transport system and the approximate analysis that has been made to estimate the loads coming on major equipments. The primary heat transport system for Narora Atomic Power Project is adopting vertical steam generators and pumps equally divided on either side of the reactor with inter-connecting pipes and feeders. Since the system is mainly spring supported with movement of a few points in certain direction defined but no anchorage, it represents a good problem for flexibility analysis which can only be solved in one step by developing a good computer programme. (author)

  11. Simulation of the control system for the KEK main ring power supply-II

    International Nuclear Information System (INIS)

    Kabe, Atsushi; Sato, Hikaru; Masuda, Masayoshi; Takeda, Masahiro; Nakano, Michio

    1977-05-01

    The tracking error between the magnetic field of the bending magnets and that of the quadrupole magnets is a very important problem of the KEK main ring power supply. Tacking performance is simulated using DDS (Digital Dynamics Simulator) in cases of 8 GeV and 12 GeV operation of the KEK main ring. The results of simulation coincide with the data of actual system. (auth.)

  12. Influence of Shaft Torsional Stiffness on Dynamic Response of Four-Stage Main Transmission System

    Directory of Open Access Journals (Sweden)

    Yuan Chen

    2018-01-01

    Full Text Available Dynamic response analysis has potential for increasing fatigue life of the components in the transmission of a multistage main transmission system. The calculated data can demonstrate the influence of shaft torsional stiffness on dynamic characteristics of the system. Detecting key shafts of the system and analyzing their sensitivity are important for the design of four-stage helicopter gear box. Lumped mass method is applied for dynamic modeling and Fourier method is used to solve differential equation of the system. Results of the analysis indicate that key shafts can be designed carefully to improve the performance of the transmission system.

  13. The main cooling system of the RSG-GAS: a study of dissabilities

    International Nuclear Information System (INIS)

    Sudiyono

    1999-01-01

    The RSG-GAS has been operating since August 1987 until now. During that period it has been many papers explain the performance and the safety features of the system. Those papers does not mean any think to the system, but the wickednesses of the system. Is still remain stick on it. This paper describes the wickednesses of the RSG-GAS main cooling system with an objective that their will be a corrective action of the management to improve the system so that such wickednesses can be minimized or even be avoided

  14. Welding overlay analysis of dissimilar metal weld cracking of feedwater nozzle

    International Nuclear Information System (INIS)

    Tsai, Y.L.; Wang, Li. H.; Fan, T.W.; Ranganath, Sam; Wang, C.K.; Chou, C.P.

    2010-01-01

    Inspection of the weld between the feedwater nozzle and the safe end at one Taiwan BWR showed axial indications in the Alloy 182 weld. The indication was sufficiently deep that continued operation could not be justified considering the crack growth for one cycle. A weld overlay was decided to implement for restoring the structural margin. This study reviews the cracking cases of feedwater nozzle welds in other nuclear plants, and reports the lesson learned in the engineering project of this weld overlay repair. The overlay design, the FCG calculation and the stress analysis by FEM are presented to confirm that the Code Case structural margins are met. The evaluations of the effect of weld shrinkage on the attached feedwater piping are also included. A number of challenges encountered in the engineering and analysis period are proposed for future study.

  15. Audit calculation of the limiting CESSAR feedwater-line-break transient with RELAP5/MOD1

    International Nuclear Information System (INIS)

    Chung, K.S.; Kennedy, M.F.; Guttmann, J.

    1983-01-01

    Argonne National Laboratory (ANL) performed a series of audit calculations of the limiting FLB transient presented in Appendix 15B to the CESSAR FSAR, supported by a limited number of additional calculations to investigate the sensitivity of the results (in terms of peak primary reactor system pressure) to break area and reactor trip time. The latter calculations were performed to quantify potential benefits in crediting reactor tip on low steam generator downcomer water level, which occurs earlier than the trip shown in the limiting FSAR transient, which tripped on high pressurizer pressure. These calculations were performed to verify the break spectrum results presented by C-E and to insure that C-E did indeed analyze the limiting transient. All of the ANL calculations were performed with RELAP5/MOD1 (cycle 18) using an input deck developed at ANL from CESSAR plant data provided by C-E. In this paper we compare the results and provide insight into the generic behavior of a Feedwater Line Break transient

  16. The New Control and Interlock System for the SPS Main Power Converters

    CERN Document Server

    Denis, B; Mugnier, C; Varas, J

    1999-01-01

    The Control and Interlock System (CIS) of the SPS main power converters was designed in the mid-70s and became increasingly difficult to maintain. A new system based on Programmable Logic Controllers has been developed by an external contractor in close collaboration with CERN. The system is now operational and fully integrated in the SPS/LEP control infrastructure. The CIS is the first major contracted industrial solution used to control accelerator equipment directly involved in the production of particle beams at CERN. This paper gives an overview of the SPS main power converter installation and describes both the contractual and technical solution adopted for the CIS. It first explains how the system was specified and how the contractual relationship was defined to respect CERN’s purchasing rules and the operational requirements of the SPS accelerator. The architectural design of the new system is presented with special emphasis on how the conflict between safety and availability has been addressed.

  17. Main control computer security model of closed network systems protection against cyber attacks

    Science.gov (United States)

    Seymen, Bilal

    2014-06-01

    The model that brings the data input/output under control in closed network systems, that maintains the system securely, and that controls the flow of information through the Main Control Computer which also brings the network traffic under control against cyber-attacks. The network, which can be controlled single-handedly thanks to the system designed to enable the network users to make data entry into the system or to extract data from the system securely, intends to minimize the security gaps. Moreover, data input/output record can be kept by means of the user account assigned for each user, and it is also possible to carry out retroactive tracking, if requested. Because the measures that need to be taken for each computer on the network regarding cyber security, do require high cost; it has been intended to provide a cost-effective working environment with this model, only if the Main Control Computer has the updated hardware.

  18. Three Types of Active Lubrication Systems for the Main Bearings of Reciprocating Machines

    DEFF Research Database (Denmark)

    Santos, Ilmar; Pulido, E. E.

    2010-01-01

    method (flexible components). The main equations that govern the dynamics of the injection for a hydraulic-actuated, a piezoelectric-actuated and a mechanical-actuated oil injector are presented in this study. The global system is numerically solved using as a case of study a single-cylinder combustion......In the paper the authors investigate three different schemes for the realization of the controllable oil injection system to be couple to the main engine bearings. The use of active lubrication in fluid film bearings helps to enhance the hydrodynamic fluid film by increasing the fluid film...

  19. Non-volatile main memory management methods based on a file system.

    Science.gov (United States)

    Oikawa, Shuichi

    2014-01-01

    There are upcoming non-volatile (NV) memory technologies that provide byte addressability and high performance. PCM, MRAM, and STT-RAM are such examples. Such NV memory can be used as storage because of its data persistency without power supply while it can be used as main memory because of its high performance that matches up with DRAM. There are a number of researches that investigated its uses for main memory and storage. They were, however, conducted independently. This paper presents the methods that enables the integration of the main memory and file system management for NV memory. Such integration makes NV memory simultaneously utilized as both main memory and storage. The presented methods use a file system as their basis for the NV memory management. We implemented the proposed methods in the Linux kernel, and performed the evaluation on the QEMU system emulator. The evaluation results show that 1) the proposed methods can perform comparably to the existing DRAM memory allocator and significantly better than the page swapping, 2) their performance is affected by the internal data structures of a file system, and 3) the data structures appropriate for traditional hard disk drives do not always work effectively for byte addressable NV memory. We also performed the evaluation of the effects caused by the longer access latency of NV memory by cycle-accurate full-system simulation. The results show that the effect on page allocation cost is limited if the increase of latency is moderate.

  20. Analysis of loss of normal feedwater transient using RELAP5/MOD1/NSC; KNU1 plant simulation

    International Nuclear Information System (INIS)

    Kim, Hho Jung; Chung, Bub Dong; Lee, Young Jin; Kim, Jin Soo

    1986-01-01

    Simulation of the system thermal-hydraulic parameters was carried out following the KNU1(Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on november 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS(Reactor Coolant System) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018. (Author)

  1. Development of a hardware-in- loop simulation platform for NPP main control systems

    Directory of Open Access Journals (Sweden)

    Liu Pengfei

    2017-01-01

    Full Text Available The simulation technology of the nuclear power plant are gradually applying to the nuclear power industry. However, most of the research on nuclear power plant simulation system only focus on pure computerized simulation at present, and it is difficult to fully display the characteristics of the simulating objects. In order to simulate the response characteristics of control system more really, a hardware-in-loop simulation platform of main control systems in the nuclear power plant has been developed in this paper. This simulation platform consists of thermal-hydraulic model, control and protection system model, physical DCS system and real-time interactive database. A physical industrial DCS system has been coupled to this platform to simulate the main control systems in the NPP, which makes the simulation result much closer to the actual control systems. The devoloped simulation platform has been validated by some steady and transient cases in this paper. This hardware-in-loop simulation platform can be used in the simulation and optimal design of NPP control systems. Furthermore, it can be used in the failure mode and effect analysis of the instrumentation and control systems in the nuclear power plant.

  2. Reliability model for helicopter main gearbox lubrication system using influence diagrams

    International Nuclear Information System (INIS)

    Rashid, H.S.J.; Place, C.S.; Mba, D.; Keong, R.L.C.; Healey, A.; Kleine-Beek, W.; Romano, M.

    2015-01-01

    The loss of oil from a helicopter main gearbox (MGB) leads to increased friction between components, a rise in component surface temperatures, and subsequent mechanical failure of gearbox components. A number of significant helicopter accidents have been caused due to such loss of lubrication. This paper presents a model to assess the reliability of helicopter MGB lubricating systems. Safety risk modeling was conducted for MGB oil system related accidents in order to analyse key failure mechanisms and the contributory factors. Thus, the dominant failure modes for lubrication systems and key contributing components were identified. The Influence Diagram (ID) approach was then employed to investigate reliability issues of the MGB lubrication systems at the level of primary causal factors, thus systematically investigating a complex context of events, conditions, and influences that are direct triggers of the helicopter MGB lubrication system failures. The interrelationships between MGB lubrication system failure types were thus identified, and the influence of each of these factors on the overall MGB lubrication system reliability was assessed. This paper highlights parts of the HELMGOP project, sponsored by the European Aviation Safety Agency to improve helicopter main gearbox reliability. - Highlights: • We investigated methods to optimize helicopter MGB oil system run-dry capability. • Used Influence Diagram to assess design and maintenance factors of MGB oil system. • Factors influencing overall MGB lubrication system reliability were identified. • This globally influences current and future helicopter MGB designs

  3. Main Principles of the Perspective System of SNF Management in Russia - 13333

    International Nuclear Information System (INIS)

    Baryshnikov, Mikhail

    2013-01-01

    For the last several years the System of the Spent Nuclear Fuel management in Russia was seriously changed. The paper describes the main principles of the changes and the bases of the Perspective System of SNF Management in Russia. Among such the bases there are the theses with the interesting names like 'total knowledge', 'pollutant pays' and 'pay and forget'. There is also a brief description of the modern Russian SNF Management Infrastructure. And an outline of the whole System. The System which is - in case of Russia - is quite necessary to adjust SNF accumulation and to utilize the nuclear heritage. (authors)

  4. Design optimization of single-main-amplifier KrF laser-fusion systems

    International Nuclear Information System (INIS)

    Harris, D.B.; Pendergrass, J.H.

    1985-01-01

    KrF lasers appear to be a very promising laser fusion driver for commercial applications. The Large Amplifier Module for the Aurora Laser System at Los Alamos is the largest KrF laser in the world and is currently operating at 5 kJ with 10 to 15 kJ eventually expected. The next generation system is anticipated to be a single-main-amplifier system that generates approximately 100 kJ. This paper examines the cost and efficiency tradeoffs for a complete single-main-amplifier KrF laser fusion experimental facility. It has been found that a 7% efficient $310/joule complete laser-fusion system is possible by using large amplifier modules and high optical fluences

  5. Study of an effective structural system with rational parameters based on main energy principles

    Directory of Open Access Journals (Sweden)

    Toporkova Katerina

    2017-01-01

    Full Text Available The effective architectural and construction system with rational parameters based on main energy principles applied to construction of buildings and constructions with a small-step and large-step frames, representing the metal frame of full factory production collected in spatial system on high-strength bolts and previously strained combined prestressed concrete slabs which formed without timbering is proposed in this paper. The main constructive and technological features of the proposed frame, which allows reducing construction period, increasing working efficiency, and reducing labor intensity by using factory-made materials, quick erection of all process elements through the use of highstrength bolts is considered. The advantages of this constructive system in comparison with alternative systems are shown. The basic concepts of "rational decisions" to the design, namely, the objective of the optimal management of the structure parameters, which can not only improve its basic performance indicators, but also, and most importantly, improve operational reliability, is presented.

  6. Massively Parallel Sort-Merge Joins in Main Memory Multi-Core Database Systems

    OpenAIRE

    Martina-Cezara Albutiu, Alfons Kemper, Thomas Neumann

    2012-01-01

    Two emerging hardware trends will dominate the database system technology in the near future: increasing main memory capacities of several TB per server and massively parallel multi-core processing. Many algorithmic and control techniques in current database technology were devised for disk-based systems where I/O dominated the performance. In this work we take a new look at the well-known sort-merge join which, so far, has not been in the focus of research ...

  7. Systems-Level Energy Audit for Main Complex, Construction Engineering Research Laboratory

    National Research Council Canada - National Science Library

    Lin, Mike

    2003-01-01

    ... (Buildings 1, 2, and 3) was conducted. The goals of the audit were to review energy and water use in the current main complex building, to review and inventory energy system equipment, and to devise short- and long-term energy improvement...

  8. Diagnosis of Feedwater Heater Performance Degradation using Fuzzy Approach

    International Nuclear Information System (INIS)

    Kim, Hyeonmin; Kang, Yeon Kwan; Heo, Gyunyoung; Song, Seok Yoon

    2014-01-01

    It is inevitable to avoid degradation of component, which operates continuously for long time in harsh environment. Since this degradation causes economical loss and human loss, it is important to monitor and diagnose the degradation of component. The diagnosis requires a well-systematic method for timely decision. Before this article, the methods using regression model and diagnosis table have been proposed to perform the diagnosis study for thermal efficiency in Nuclear Power Plants (NPPs). Since the regression model was numerically less-stable under changes of operating variables, it was difficult to provide good results in operating plants. Contrary to this, the diagnosis table was hard to use due to ambiguous points and to detect how it affects degradation. In order to cover the issues of previous researches, we proposed fuzzy approaches and applied it to diagnose Feedwater Heater (FWH) degradation to check the feasibility. The degradation of FWHs is not easy to be observed, while trouble such as tube leakage may bring simultaneous damage to the tube bundle. This study explains the steps of diagnosing typical failure modes of FWHs. In order to cover the technical issues of previous researches, we adopted fuzzy logic to suggest a diagnosis algorithm for the degradation of FHWs and performed feasibility study. In this paper, total 7 modes of FWH degradation modes are considered, which are High Drain Level, Low Shell Pressure, Tube Pressure Increase, Tube Fouling, Pass Partition Plate Leakage, Tube Leakage, Abnormal venting. From the literature survey and simulation, diagnosis table for FWH is made. We apply fuzzy logic based on diagnosis table. Authors verify fuzzy diagnosis for FWH degradation synthesized the random input sets from made diagnosis table. Comparing previous researches, suggested method more-stable under changes of operating variables, than regression model. On the contrary, the problem which ambiguous points and detect how it affects degradation

  9. Diagnosis of Feedwater Heater Performance Degradation using Fuzzy Approach

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonmin; Kang, Yeon Kwan; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Song, Seok Yoon [Korea Hydro and Nuclear Power, Daejeon (Korea, Republic of)

    2014-05-15

    It is inevitable to avoid degradation of component, which operates continuously for long time in harsh environment. Since this degradation causes economical loss and human loss, it is important to monitor and diagnose the degradation of component. The diagnosis requires a well-systematic method for timely decision. Before this article, the methods using regression model and diagnosis table have been proposed to perform the diagnosis study for thermal efficiency in Nuclear Power Plants (NPPs). Since the regression model was numerically less-stable under changes of operating variables, it was difficult to provide good results in operating plants. Contrary to this, the diagnosis table was hard to use due to ambiguous points and to detect how it affects degradation. In order to cover the issues of previous researches, we proposed fuzzy approaches and applied it to diagnose Feedwater Heater (FWH) degradation to check the feasibility. The degradation of FWHs is not easy to be observed, while trouble such as tube leakage may bring simultaneous damage to the tube bundle. This study explains the steps of diagnosing typical failure modes of FWHs. In order to cover the technical issues of previous researches, we adopted fuzzy logic to suggest a diagnosis algorithm for the degradation of FHWs and performed feasibility study. In this paper, total 7 modes of FWH degradation modes are considered, which are High Drain Level, Low Shell Pressure, Tube Pressure Increase, Tube Fouling, Pass Partition Plate Leakage, Tube Leakage, Abnormal venting. From the literature survey and simulation, diagnosis table for FWH is made. We apply fuzzy logic based on diagnosis table. Authors verify fuzzy diagnosis for FWH degradation synthesized the random input sets from made diagnosis table. Comparing previous researches, suggested method more-stable under changes of operating variables, than regression model. On the contrary, the problem which ambiguous points and detect how it affects degradation

  10. Applying Systems Engineering to Improve the Main Gas Turbine Exhaust System Maintenance Strategy for the CG-47 Ticonderoga Class Cruiser

    Science.gov (United States)

    2015-09-01

    national security and prosperity (U.S. Navy 2014). In perspective, oceans are the lifeblood of the planet and its entire population . The National...maintenance strategy, reliability-centered maintenance, cost, schedule, performance, growth -work, new-work, optimal fleet response plan, time-directed...76 5. Main Gas Turbine Exhaust System Growth -Work ..................77 E. RECOMMENDATIONS TO IMPROVE THE MAIN GAS TURBINE EXHAUST SYSTEM

  11. Digital Low Level RF Systems for Fermilab Main Ring and Tevatron

    Science.gov (United States)

    Chase, B.; Barnes, B.; Meisner, K.

    1997-05-01

    At Fermilab, a new Low Level RF system is successfully installed and operating in the Main Ring. Installation is proceeding for a Tevatron system. This upgrade replaces aging CAMAC/NIM components for an increase in accuracy, reliability, and flexibility. These VXI systems are based on a custom three channel direct digital synthesizer(DDS) module. Each synthesizer channel is capable of independent or ganged operation for both frequency and phase modulation. New frequency and phase values are computed at a 100kHz rate on the module's Analog Devices ADSP21062 (SHARC) digital signal processor. The DSP concurrently handles feedforward, feedback, and beam manipulations. Higher level state machines and the control system interface are handled at the crate level using the VxWorks operating system. This paper discusses the hardware, software and operational aspects of these LLRF systems.

  12. Development and application of intelligent CAE system for cyclotron main magnet

    International Nuclear Information System (INIS)

    Zhang Tianjue; Chen Yong; Fan Mingwu

    1993-01-01

    The main magnet that represents the feature of the cyclotron is the most important part in a cyclotron construction. Though there are many codes devoted to solve magnetic field computation problems, the results from them are depended on user's skill and experience very much. To help cyclotron magnet designer get acceptable result an intelligent CAE system for cyclotron main magnet design and machining has been developed. A reasonable good results in design could be get even the designer is a beginner with the help from an expert knowledge library installed in the program. The codes include following functions: 1. Intelligent CAD; 2. 2D and 3D magnetic field computation; 3. Beam dynamics analysis; 4. CAM for main magnet

  13. A portable, low-cost, 3D-printed main magnetic field system for magnetic imaging.

    Science.gov (United States)

    Iksung Kang

    2017-07-01

    In this paper, a portable, low-cost, 3D-printed system for main magnetic field is proposed to suggest a solution for accessibility problems of current magnetic imaging systems, e.g. MRI scanner, their size and cost. The system consists of twelve pairs of NdFeB N35 permanent magnets arranged in a Halbach array in a 3D-printed, cylindrical container based on FEM simulation results by COMSOL Multiphysics 4.4b. Its magnetic field homogeneity and field strength were measured by Hall sensors, WSH-135 XPAN2 by Wilson Semiconductor, and the container was printed by 3DISON H700 by Rokit. The system generated a 5-mm imaging quality FOV and main magnetic field of 120 mT with a 12 % error in the field strength. Also, a hundred dollar was enough for the manufacture of the system with a radius of 6 cm and height of 10 cm. Given the results, I believe the system will be useful for some magnetic imaging applications, e.g. EPRI and low-field MRI.

  14. Experience feedback of an operation event during the experiment of feed-water pump switch

    International Nuclear Information System (INIS)

    Sun Shuhai; Li Huasheng; Zhang Hao

    2012-01-01

    In this paper an event is summarized and analyzed, which caused the quit of the high-pressure heaters and the nuclear power rising, during the experiment of the driven feed-water pump switch. The good experience feedback on this event is brought out through gathering related information of domestic nuclear plants. (authors)

  15. Stress analysis of LOFT containment vessel attachments for the mainsteam and feedwater piping support structures

    International Nuclear Information System (INIS)

    Finicle, D.P.

    1977-01-01

    The LOFT Containment Vessel attachments for the Mainsteam and Feedwater Piping Support Structures have been analyzed for operating and faulted loading conditions. This report contains the analysis of the connections to the containment vessel for the most current design and loading. Also contained in this report is the analysis of the piping supports

  16. Overview of the main challenges for the engineering design of the test facilities system of IFMIF

    International Nuclear Information System (INIS)

    Molla, J.; Nakamura, K.

    2009-01-01

    High intense radiation fields were demanded to IFMIF to address the lack of information on effects in materials due to radiation fields with fusion reactor features. Such intense radiation fields will also produce a number of unwanted effects in exposed materials and components. The main difficulties to achieve a reliable engineering design of the Test Facilities System during the Engineering Validation and the Engineering Design phase of IFMIF now under development are reviewed in this paper. The most challenging activities will be the design of the high flux test module, the creep fatigue test module, the test cell and the remote handling system. The intense radiation fields in the irradiation area and the high availability required for IFMIF (70%) are the main reasons for these difficulties.

  17. Robust Backstepping Control for Cold Rolling Main Drive System with Nonlinear Uncertainties

    Directory of Open Access Journals (Sweden)

    Xu Yang

    2013-01-01

    Full Text Available The nonlinear model of main drive system in cold rolling process, which considers the influence with parameter uncertainties such as clearance and variable friction coefficient, as well as external disturbance by roll eccentricity and variation of strip material quality, is built. By transformation, the lower triangular structure form of main drive system is obtained. The backstepping algorithm based on signal compensation is proposed to design a linear time-invariant (LTI robust controller, including a nominal controller and a robust compensator. A comparison with PI controller shows that the controller has better disturbance attenuation performance and tracking behaviors. Meanwhile, according to its LTI characteristic, the robust controller can be realized easily; therefore it is also appropriated to high speed dynamic rolling process.

  18. Main steam system piping response under safety/relief valve opening events

    International Nuclear Information System (INIS)

    Swain, E.O.; Esswein, G.A.; Hwang, H.L.; Nieh, C.T.

    1980-01-01

    The stresses in the main steam branch pipe of a Boiling Water Reactor due to safety/relief valve blowdown has been measured from an in situ piping system test. The test results were compared with analytical results. The predicted stresses using the current state of art analytical methods used for BWR SRV discharge transient piping response loads were found to be conservative when compared to the measured stress values. 3 refs

  19. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    International Nuclear Information System (INIS)

    Ignatiev, Victor; Feynberg, Olga; Merzlyakov, Aleksandr; Surenkov, Aleksandr; Subbotin, Vladimir; Zakirov, Raul; Toropov, Andrey; Panov, Aleksandr; Afonichkin, Valery

    2008-01-01

    To examine and demonstrate the feasibility of molten salt reactors (MSR) to reduce long lived waste toxicity and to produce efficiently electricity in closed fuel cycle, some national and international studies were initiated last years. In this paper main focus is placed on experimental evaluation of single stream Molten Salt Actinide Recycler and Transmuter (MOSART) system fuelled with different compositions of plutonium plus minor actinide trifluoride (AnF 3 ) from LWR spent nuclear fuel without U-Th support. This paper summarizes main experimental results of ISTC-1606 related to physical and chemical properties of fuel salt, container materials for fuel circuit, and fuel salt clean up of MOSART system. As result of ISTC-1606 studies claim is made, that the 7 Li,Na,Be/F and 7 Li,Be/F solvents selected for primary system appear to resolve main reactor physics, thermal hydraulics, materials compatibility, fuel salt clean up and safety problems as applied to the MOSART concept development. The created experimental facilities and the database on properties of fuel salt mixtures and container materials are used for a choice and improvement fuel salts and coolants for new applications of this high temperature technology for sustainable nuclear power development. (authors)

  20. Main Experimental Results of ISTC-1606 for Recycling and Transmutation in Molten Salt Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, Victor; Feynberg, Olga; Merzlyakov, Aleksandr; Surenkov, Aleksandr [Russian Research Center - Kurchatov Institute, Kurchatov sq. 1, Moscow, RF, 123182 (Russian Federation); Subbotin, Vladimir; Zakirov, Raul; Toropov, Andrey; Panov, Aleksandr [Russian Federal Nuclear Center - Institute of Technical Physics, Snezhinsk (Russian Federation); Afonichkin, Valery [Institute of High-Temperature Electrochemistry, Ekaterinburg (Russian Federation)

    2008-07-01

    To examine and demonstrate the feasibility of molten salt reactors (MSR) to reduce long lived waste toxicity and to produce efficiently electricity in closed fuel cycle, some national and international studies were initiated last years. In this paper main focus is placed on experimental evaluation of single stream Molten Salt Actinide Recycler and Transmuter (MOSART) system fuelled with different compositions of plutonium plus minor actinide trifluoride (AnF{sub 3}) from LWR spent nuclear fuel without U-Th support. This paper summarizes main experimental results of ISTC-1606 related to physical and chemical properties of fuel salt, container materials for fuel circuit, and fuel salt clean up of MOSART system. As result of ISTC-1606 studies claim is made, that the {sup 7}Li,Na,Be/F and {sup 7}Li,Be/F solvents selected for primary system appear to resolve main reactor physics, thermal hydraulics, materials compatibility, fuel salt clean up and safety problems as applied to the MOSART concept development. The created experimental facilities and the database on properties of fuel salt mixtures and container materials are used for a choice and improvement fuel salts and coolants for new applications of this high temperature technology for sustainable nuclear power development. (authors)

  1. Massively Parallel Sort-Merge Joins in Main Memory Multi-Core Database Systems

    OpenAIRE

    Albutiu, Martina-Cezara; Kemper, Alfons; Neumann, Thomas

    2012-01-01

    Two emerging hardware trends will dominate the database system technology in the near future: increasing main memory capacities of several TB per server and massively parallel multi-core processing. Many algorithmic and control techniques in current database technology were devised for disk-based systems where I/O dominated the performance. In this work we take a new look at the well-known sort-merge join which, so far, has not been in the focus of research in scalable massively parallel mult...

  2. 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.)

  3. Issues regarding Risk Effect Analysis of Digitalized Safety Systems and Main Risk Contributors

    International Nuclear Information System (INIS)

    Kang, Hyun Gook; Jang, Seung-Cheol

    2008-01-01

    Risk factors of safety-critical digital systems affect overall plant risk. In order to assess this risk effect, a risk model of a digitalized safety system is required. This article aims to provide an overview of the issues when developing a risk model and demonstrate their effect on plant risk quantitatively. Research activities in Korea for addressing these various issues, such as the software failure probability and the fault coverage of self monitoring mechanism are also described. The main risk contributors related to the digitalized safety system were determined in a quantitative manner. Reactor protection system and engineered safety feature component control system designed as part of the Korean Nuclear I and C System project are used as example systems. Fault-tree models were developed to assess the failure probability of a system function which is designed to generate an automated signal for actuating both of the reactor trip and the complicated accident-mitigation actions. The developed fault trees were combined with a plant risk model to evaluate the effect of a digitalized system's failure on the plant risk. (authors)

  4. Main Strategies of Plant Expression System Glycoengineering for Producing Humanized Recombinant Pharmaceutical Proteins.

    Science.gov (United States)

    Rozov, S M; Permyakova, N V; Deineko, E V

    2018-03-01

    Most the pharmaceutical proteins are derived not from their natural sources, rather their recombinant analogs are synthesized in various expression systems. Plant expression systems, unlike mammalian cell cultures, combine simplicity and low cost of procaryotic systems and the ability for posttranslational modifications inherent in eucaryotes. More than 50% of all human proteins and more than 40% of the currently used pharmaceutical proteins are glycosylated, that is, they are glycoproteins, and their biological activity, pharmacodynamics, and immunogenicity depend on the correct glycosylation pattern. This review examines in detail the similarities and differences between N- and O-glycosylation in plant and mammalian cells, as well as the effect of plant glycans on the activity, pharmacokinetics, immunity, and intensity of biosynthesis of pharmaceutical proteins. The main current strategies of glycoengineering of plant expression systems aimed at obtaining fully humanized proteins for pharmaceutical application are summarized.

  5. A model for the Space Shuttle Main Engine High Pressure Oxidizer Turbopump shaft seal system

    Science.gov (United States)

    Paxson, Daniel E.

    1990-01-01

    A model of the High Pressure Oxidizer Turbopump (HPOTP) shaft seal system on the Space Shuttle Main Engine (SSME) is described. The model predicts the fluid properties and flow rates throughout this system for a number of conditions simulating failed seals. The results agree well with qualitative expectations and redline values but cannot be verified with actual data due to the lack thereof. The results indicate that each failure mode results in a unique distribution of properties throughout the seal system and can therefore be individually identified given the proper instrumentation. Furthermore, the detection process can be built on the principle of qualitative reasoning without the use of exact fluid property values. A simplified implementation of the model which does not include the slinger/labyrinth seal combination has been developed and will be useful for inclusion in a real-time diagnostic system.

  6. New data acquisition system for beam loss monitor used in J-PARC main ring

    Science.gov (United States)

    Satou, K.; Toyama, T.; Kamikubota, N.; Yoshida, S.; Matsushita, J.; Wakita, T.; Sugiyama, M.; Morino, T.

    2018-04-01

    A new data acquisition system has been developed continually as a part of the development of a new beam loss monitor (BLM) system for the J-PARC main ring. This development includes a newly designed front-end isolation amp that uses photo-couplers and a VME-based new analog-to-digital converter (ADC) system. Compared to the old amp, the new amp has a 10 times higher conversion impedance for the input current to the output voltage; this value is 1 M Ω. Moreover, the bandwidth was improved to from DC to 50 kHz, which is about two orders of magnitude greater than the previously used bandwidth. The theoretical estimations made in this study roughly agree with the frequency response obtained for the new system. The new ADC system uses an on-board field-programmable gate array chip for signal processing. By replacing the firmware of this chip, changes pertaining to future accelerator upgrade plans may be introduced into the new ADC system; in addition, the ADC system can be used in other applications. The sampling speed of the system is 1 MS/s, and it exhibits a 95 dBc spurious-free dynamic range and 16.5 effective number of bits. The obtained waveform and integrated charge data are compared with two reference levels in the ADC system. If the data exceeds the reference level, the system generates an alarm to dump the beams. By using the new data acquisition system, it was proved that the new BLM system shows a wide dynamic range of 160 dB. In this study, the details of the new data acquisition system are described.

  7. Impact of the operation of non-displaced feedwater heaters on the performance of Solar Aided Power Generation plants

    International Nuclear Information System (INIS)

    Qin, Jiyun; Hu, Eric; Nathan, Graham J.

    2017-01-01

    Highlights: • Impact of non-displaced feedwater heater on plant’s performance has been evaluated. • Two operation strategies for non-displaced feedwater heater has been proposed. • Constant temperature strategy is generally better. • Constant mass flow rate strategy is suit for rich solar thermal input. - Abstract: Solar Aided Power Generation is a technology in which low grade solar thermal energy is used to displace the high grade heat of the extraction steam in a regenerative Rankine cycle power plant for feedwater preheating purpose. The displaced extraction steam can then expand further in the steam turbine to generate power. In such a power plant, using the (concentrated) solar thermal energy to displace the extraction steam to high pressure/temperature feedwater heaters (i.e. displaced feedwater heaters) is the most popular arrangement. Namely the extraction steam to low pressure/temperature feedwater heaters (i.e. non-displaced feedwater heaters) is not displaced by the solar thermal energy. In a Solar Aided Power Generation plants, when solar radiation/input changes, the extraction steam to the displaced feedwater heaters requires to be adjusted according to the solar radiation. However, for the extraction steams to the non-displaced feedwater heaters, it can be either adjusted accordingly following so-called constant temperature strategy or unadjusted i.e. following so-called constant mass flow rate strategy, when solar radiation/input changes. The previous studies overlooked the operation of non-displaced feedwater heaters, which has also impact on the whole plant’s performance. This paper aims to understand/reveal the impact of the two different operation strategies for non-displaced feedwater heaters on the plant’s performance. In this paper, a 300 MW Rankine cycle power plant, in which the extraction steam to high pressure/temperature feedwater heaters is displaced by the solar thermal energy, is used as study case for this purpose. It

  8. Performance of Anidolic Daylighting Systems in tropical climates - Parametric studies for identification of main influencing factors

    Energy Technology Data Exchange (ETDEWEB)

    Linhart, Friedrich; Scartezzini, Jean-Louis [Solar Energy and Building Physics Laboratory (LESO-PB), Swiss Federal Institute of Technology in Lausanne (EPFL), 1015 Lausanne (Switzerland); Wittkopf, Stephen K. [School of Design and Environment, Department of Architecture, National University of Singapore, 4 Architecture Drive, Singapore 117566 (Singapore)

    2010-07-15

    Making daylight more available in buildings is highly desirable, not only for reasons of energy-efficiency, but also for improvement of occupants' health and well-being. Core-daylighting, that is daylight provision in areas situated at considerable distances from facades and windows, is currently one of the main challenges in sustainable building design. Anidolic Daylighting Systems (ADSs) are one very promising technology in the field of core-daylighting, but commercial solutions that are not only well-performing but also financially competitive are not yet widely available. This article presents results of parametric studies on Anidolic Integrated Ceilings (AICs), a special type of ADS, for identification of main influencing factors. The article describes a reliable method for simulating ADS and AIC performance under given sky conditions. Various simulation results for the example location Singapore are discussed in detail, it is concluded that the main influencing factors are coating material, system dimensions and external obstruction, and those influencing factors' potential impacts are quantified. It is shown that AIC overall efficiencies can reach up to almost 50% in Singapore. The essentially new results presented in this article can be of great help to architects, engineers and scientists in the future, when it comes to precisely dimensioning ADS for various buildings and daylight conditions. (author)

  9. CONCEPT OF "GENERIC ORIGIN" AND CHRISTIAN VALUES AS THE MAIN CHARACTERISTICS OF PAVEL VASILYEV'S ARTISTIC SYSTEM

    Directory of Open Access Journals (Sweden)

    Nina Viktorovna Popova

    2014-11-01

    Full Text Available This article contemplates the system of moral and aesthetic values which predetermined the main characteristics of Pavel Vasilyev’s artistic system, and traces the main stages of his creative biography. The main attention is paid to his key works – poems The Song about Death of Cossack Army, Salt Revolt, Khristolyubovskiye Printed Cottons, Summer, and August, as well as the poems written during the period of 1930–1937 (Farewell to Friends, Stonemason, Old Moscow, Serafi m Dagaev’s Town, Horse, Camel, Injured Song, Oath on Chalice. The article raises the question about the most important role of the concept of “generic origin”, Christian values and Christian symbolism in Vasilyev’s poetics and studies the problem of using various forms of verbal folk art in the art structure of his poems for the reflection of Russian orthodox traditions. The axiological approach and methods of historical typological and structural semiotic methods for studying art texts are used.

  10. ANALYSIS OF THE MAIN STRUCTURAL CHARACTERISTICS OF THE ROMANIAN FINANCIAL SYSTEM

    Directory of Open Access Journals (Sweden)

    Marius Constantin Apostoaie

    2011-12-01

    Full Text Available To know and to understand the factors within a financial system that significantly influences the institutions behaviour and the market responses to the central bank actions is of great importance for the monetary authorities and for the achievement of their monetary policy objectives. In Romania, the National Bank must pay special attention to financial institutions given their significant role in the process of financial intermediation (bank-based system. This study aims to conduct an investigation of the main coordinates of the Romanian financial system that are relevant for the study of the monetary policy transmission mechanism. The study focuses on the following features of the Romanian financial system: financial deepening, the existence of a bank-based system, the concentration degree of the Romanian banking sector, the balance sheet structure of monetary financial institutions and the soundness of the banking system. The study also considers the effects of the recent financial market turmoil on the financial system. The span of the empirical study covers the period after 2000 but brings to the forefront the period before and after October 2008. We believe that this research provides useful information for a future analysis of the Romanian monetary policy transmission mechanism.

  11. Digitized operator evaluation system for main control room of nuclear power plant

    International Nuclear Information System (INIS)

    Chen Yu; Yan Shengyuan; Chen Wenlong

    2014-01-01

    In order to evaluate the human-machine system matching relation of main control room in nuclear power plant accurately and efficiently, the expression and parameters of operator human body model were analyzed, and the evaluation required function of digital operator was determined. Based on the secondary development technology, the digital operator evaluation body model was developed. It could choose generation, gender, operation posture, single/eyes horizon, and left/right hand up to the domain according to the needs of specific evaluation, it was used to evaluate whether display information can be visible and equipment can be touch, and it also has key evaluation functions such as workspace and character visibility at the same time. The examples show that this method can complete the evaluation work of human-machine matching relation for main control room of nuclear power plant accurately, efficiently and quickly, and achieve the most optimal human-machine coordination relationship. (authors)

  12. Introduction of a backup system for data and servers of main IT infrastructure services

    International Nuclear Information System (INIS)

    Hirayama, Takashi; Kannari, Masaaki

    2013-06-01

    The optimization of the JAEA network system has been promoted in accordance with the optimization plan which has the fundamental principles of ensuring its dependability, information security and usability. In respect to ensuring the dependability, we addressed to a) the reduction of both trouble probability and recovery time, and b) an execution of the business continuity plan in time of large-scale earthquake. For the latter, we installed an e-mail backup server and an alternate connection to the internet in Kansai Photon Science Institute (Kizu-area) based on lesson learned from the experience of the Great East Japan Earthquake on March 11, 2011. In addition, we introduced a backup system for data and servers of other main IT infrastructure services. This report documents the configuration and operation of the backup system. (author)

  13. Units 3 and 4 steam generators new water level control system

    International Nuclear Information System (INIS)

    Dragoev, D.; Genov, St.

    2001-01-01

    The Steam Generator Water Level Control System is one of the most important for the normal operation systems, related to the safety and reliability of the units. The main upgrading objective for the SG level and SGWLC System modernization is to assure an automatic maintaining of the SG level within acceptable limits (below protections and interlocks) from 0% to 100% of the power in normal operation conditions and in case of transients followed by disturbances in the SG controlled parameters - level, steam flow, feedwater flow and/or pressure/temperature. To achieve this objective, the computerized controllers of new SG water level control system follows current computer control technology and is implemented together with replacement of the feedwater control valves and the needed I and C equipment. (author)

  14. Development of Alarm System link Drawing for Operation Support for APR1400 Digital Main Control Room

    International Nuclear Information System (INIS)

    Kim, Ki-Hwan

    2016-01-01

    Digitalized MMI(Man-Machine Interface) including Digital Main Control Room(MCR) and digital I and C system was being applied for SKN 3 and 4 Nuclear Power Plant(NPP) and subsequent APR1400 NPP type. But, operators can not easily find instrument for alarm immediately. Therefore, Alarm system is required to easily find instrument for Alarm. For this implementation, we will plan system design considering design feature without affecting network load and CPU load. We have developed Alarm system link drawing for digital MCR. Operators of the digitalized MCR navigates from their consoles to the drawings related to the plant alarms and their instruments or the operation status. Such method gives cognitive load to the operators having to travel to different locations in finding the related information. Screen Sharing System, which is the fundamental technique for Drawing Interconnection Alarm System is close to completion, and it should be functionally tested and verified by the human factor engineering. For the actual application to the operating plants, the drawings to be interconnected to the alarms and the opinions from the operators/maintenance departments for designating alarm number should be surveyed, Also, another function that allows the access to the alarm related drawings not only from the MCR but also from the other offices

  15. Development of Alarm System link Drawing for Operation Support for APR1400 Digital Main Control Room

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki-Hwan [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    Digitalized MMI(Man-Machine Interface) including Digital Main Control Room(MCR) and digital I and C system was being applied for SKN 3 and 4 Nuclear Power Plant(NPP) and subsequent APR1400 NPP type. But, operators can not easily find instrument for alarm immediately. Therefore, Alarm system is required to easily find instrument for Alarm. For this implementation, we will plan system design considering design feature without affecting network load and CPU load. We have developed Alarm system link drawing for digital MCR. Operators of the digitalized MCR navigates from their consoles to the drawings related to the plant alarms and their instruments or the operation status. Such method gives cognitive load to the operators having to travel to different locations in finding the related information. Screen Sharing System, which is the fundamental technique for Drawing Interconnection Alarm System is close to completion, and it should be functionally tested and verified by the human factor engineering. For the actual application to the operating plants, the drawings to be interconnected to the alarms and the opinions from the operators/maintenance departments for designating alarm number should be surveyed, Also, another function that allows the access to the alarm related drawings not only from the MCR but also from the other offices.

  16. Dynamic optimization of maintenance and improvement planning for water main system: Periodic replacement approach

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Woo; Choi, Go Bong; Lee, Jong Min [Seoul National University, Seoul (Korea, Republic of); Suh, Jung Chul [Samchully Corporation, Seoul (Korea, Republic of)

    2016-01-15

    This paper proposes a Markov decision process (MDP) based approach to derive an optimal schedule of maintenance, rehabilitation and replacement of the water main system. The scheduling problem utilizes auxiliary information of a pipe such as the current state, cost, and deterioration model. The objective function and detailed algorithm of dynamic programming are modified to solve the periodic replacement problem. The optimal policy evaluated by the proposed algorithm is compared to several existing policies via Monte Carlo simulations. The proposed decision framework provides a systematic way to obtain an optimal policy.

  17. Intelligent CAE system of CYCLONE type cyclotron main magnet and its applications

    International Nuclear Information System (INIS)

    Zhang Tianjue; Chen Yong; Fan Mingwu

    1993-01-01

    The main magnet that features the cyclotron is the most important part in a cyclotron construction. Though there are many codes devoted to solving magnetic field computation problems, the results depend on the user's skill and experience very much. To help a cyclotron magnet designer get acceptable results, an intelligent CAE system for CYCLONE type cyclotron magnet design and machining has been developed. Reasonable good results could be got even the designer with the help from an expert knowledge library installed in the program

  18. A main amplifier circuit and data acquisition system for charged particle detector array

    International Nuclear Information System (INIS)

    Hao Rui; Ge Yucheng

    2011-01-01

    The charged particle detector array has huge amounts of signal and needs high counting rate. To meet the requirements, a main amplifier and analog-to-digital conversion circuit based on high-speed op-amp chips and ADC chip was designed. A 51-MCU was used to control the circuit of ADC and the USB communication chip. The signals were digitized and uploaded by the MCU-ADC-USB circuit. The whole system has a compact hardware structure and a reasonable controlling software, which meet the design requirements. (authors)

  19. Duct corrosion in the ventilating air conditioning system for Main Control Room

    International Nuclear Information System (INIS)

    Yamada, Kohei; Kobayashi, Takashi; Minami, Akiko; Fukuba, Kazushi

    2014-01-01

    Higashidori Nuclear Power Station, start-of-operation in December 2005, is a relatively new plant. We decided to get original data of air duct condition to determine maintenance policy of air duct, because planned maintenance of air duct has never been done and the corrosion of air duct has occurred in other plant. In January 2014, we found a corrosion-hole at the downstream of the inlet damper in the ventilating air conditioning system for Main Control Room (MCR). We supposed that the cause of rapid corrosion is related to the characteristic environment of this site. (author)

  20. Intelligent CAE system of CYCLONE type cyclotron main magnet and its applications

    International Nuclear Information System (INIS)

    Zhang Tianjue; Chen Yong; Fan Mingwu

    1992-07-01

    The main magnet that represents the feature of the cyclotron is the most important part in a cyclotron construction. Though there are many codes devoted to solve magnetic field computation problems, the results from them depend on the user's skill and experience very much. To help cyclotron magnet designer to get acceptable results, an intelligent CAE (computer aided engineering) system for CYCLONE type cyclotron magnet design and machining has been developed. A reasonable good results in the design could be got even if the designer is a beginner, because of the help from an expert knowledge library installed in the program

  1. Migration of Older to New Digital Control Systems in Nuclear Power Plant Main Control Rooms

    Energy Technology Data Exchange (ETDEWEB)

    Kovesdi, Casey Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Joe, Jeffrey Clark [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-04-01

    The United States (U.S.) Department of Energy (DOE) Office of Nuclear Energy (NE) has the primary mission to advance nuclear power by resolving socio-technical issues through research and development (R&D). One DOE-NE activity supporting this mission is the Light Water Reactor Sustainability (LWRS) program. LWRS has the overall objective to sustain the operation of existing commercial nuclear power plants (NPPs) through conducting R&D across multiple “pathways,” or R&D focus areas. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies pathway conducts targeted R&D to address aging and reliability concerns with the legacy instrumentation and control (I&C) and related information systems in operating U.S. NPPs. This work involves (1) ensuring that legacy analog II&C systems are not life-limiting issues for the LWR fleet, and (2) implementing digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Under the LWRS Advanced II&C pathway, Human Factors experts at Idaho National Laboratory (INL) have been conducting R&D in support of NPP main control room (MCR) modernization activities. Work in prior years has focused on migrating analog I&C systems to new digital I&C systems (). In fiscal year 2016 (FY16), one new focus area for this research is migrating older digital I&C systems to new and advanced digital I&C systems. This report summarizes a plan for conducting a digital-to-digital migration of a legacy digital I&C system to a new digital I&C system in support of control room modernization activities.

  2. Migration of Older to New Digital Control Systems in Nuclear Power Plant Main Control Rooms

    International Nuclear Information System (INIS)

    Kovesdi, Casey Robert; Joe, Jeffrey Clark

    2016-01-01

    The United States (U.S.) Department of Energy (DOE) Office of Nuclear Energy (NE) has the primary mission to advance nuclear power by resolving socio-technical issues through research and development (R&D). One DOE-NE activity supporting this mission is the Light Water Reactor Sustainability (LWRS) program. LWRS has the overall objective to sustain the operation of existing commercial nuclear power plants (NPPs) through conducting R&D across multiple ''pathways,'' or R&D focus areas. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies pathway conducts targeted R&D to address aging and reliability concerns with the legacy instrumentation and control (I&C) and related information systems in operating U.S. NPPs. This work involves (1) ensuring that legacy analog II&C systems are not life-limiting issues for the LWR fleet, and (2) implementing digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Under the LWRS Advanced II&C pathway, Human Factors experts at Idaho National Laboratory (INL) have been conducting R&D in support of NPP main control room (MCR) modernization activities. Work in prior years has focused on migrating analog I&C systems to new digital I&C systems (). In fiscal year 2016 (FY16), one new focus area for this research is migrating older digital I&C systems to new and advanced digital I&C systems. This report summarizes a plan for conducting a digital-to-digital migration of a legacy digital I&C system to a new digital I&C system in support of control room modernization activities.

  3. Main Steam Line Break Analysis for the Fully Passive Safety System of SMART

    International Nuclear Information System (INIS)

    Kim, Seong Wook; Chun, Ji Han; Bae, Kyoo Hwan; Kim, Keung Koo

    2013-01-01

    The standard design approval of SMART (System-integrated Modular Advanced ReacTor) developed by KAERI and KEPCO consortium was issued on July 4, 2012. Although SMART has enhanced safety compared to the conventional reactor, there is a demand to meet the 'passive safety performance requirements' after the Fukushima accident. The passive safety performance requirements are the capabilities to maintain the plant at a safe shutdown condition for a minimum of 72 hours without AC power supply or operator action in case of design basis accident (DBA). To satisfy the requirements, KAERI is developing a safety enhanced SMART by adopting a passive safety injection system. The passive safety injection system developed for SMART is a gravity-driven injection system, which consists of four trains, each of which includes a pressure balance line, core makeup tank (CMT), safety injection tank (SIT) and injection line. The CMT plays an important role to inject borated water into the RCS to prevent or dissolve the return to power (re-criticality) condition during the event of increase in heat removal by the secondary system. The main steam line break accident (MSLB) is the most limiting accident for an increase in heat removal by the secondary system. In this study, the safety analysis results of MSLBs at hot full power condition and at hot zero power condition in view of re-criticality are given. The MSLB accident has been analyzed for the SMART adopting fully passive safety system in the aspect of re-criticality. The results show that the core remains subcritical condition throughout the transient due to the borated water injected by the CMT. As further works, many kinds of analyses and sensitivity studies should be performed for the design establishment and improvement of the fully passive system of SMART

  4. LARGE SCALE DISTRIBUTED PARAMETER MODEL OF MAIN MAGNET SYSTEM AND FREQUENCY DECOMPOSITION ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    ZHANG,W.; MARNERIS, I.; SANDBERG, J.

    2007-06-25

    Large accelerator main magnet system consists of hundreds, even thousands, of dipole magnets. They are linked together under selected configurations to provide highly uniform dipole fields when powered. Distributed capacitance, insulation resistance, coil resistance, magnet inductance, and coupling inductance of upper and lower pancakes make each magnet a complex network. When all dipole magnets are chained together in a circle, they become a coupled pair of very high order complex ladder networks. In this study, a network of more than thousand inductive, capacitive or resistive elements are used to model an actual system. The circuit is a large-scale network. Its equivalent polynomial form has several hundred degrees. Analysis of this high order circuit and simulation of the response of any or all components is often computationally infeasible. We present methods to use frequency decomposition approach to effectively simulate and analyze magnet configuration and power supply topologies.

  5. Main results on the RF amplitude and phase regulation systems in operation at GANIL

    International Nuclear Information System (INIS)

    Joubert, A.; Ducoudret, B.; Labiche, J.C.; Loyant, J.M.

    1984-06-01

    The general features of the amplitude and phase regulations and their control systems are briefly reviewed. These feedback control systems are fully under the control of the main computer aided by dedicated CAMAC microprocessors for actions such as starting, parameters tuning or phase stability surveying. Numerous results obtained with spectrum analysis method give the actual RF purity and the residual modulation and crossmodulation noise level for all RF signals picked up in the RF resonators. A typical value for the noise immunity is 80 dB below the carrier at 100 Hz deviation. Another set of results gives the actual long term phase drift between resonators (< 0.2 RF degree within 6 hours). The stability of the RF phases is confirmed by on line beam phase measurements

  6. A System for Series Magnetic Measurements of the LHC Main Quadrupoles

    CERN Document Server

    Smirnov, N; Chiusano, F; Dunkel, O; Legrand, P; Schloss, S; Schnizer, P; Sievers, P

    2002-01-01

    More than 400 twin aperture lattice quadrupoles are needed for the Large Hadron Collider (LHC) which is under construction at CERN. The main quadrupole is assembled with correction magnets in a common cryostat called the Short Straight Section (SSS). We plan to measure all SSS's in cold conditions with an unprecedented accuracy: integrated gradient of the field within 150 ppm, harmonics in a range of 1 to 5 ppm, magnetic axis of all elements within 0.1 mm and their field direction within 0.2 mrad. In this paper we describe the automatic measurement system that we have designed, built and calibrated. Based on the results obtained on the two first prototypes of the SSS's (SSS3 and SSS4) we show that this system meets all above requirements.

  7. Characterizing Observed Limit Cycles in the Cassini Main Engine Guidance Control System

    Science.gov (United States)

    Rizvi, Farheen; Weitl, Raquel M.

    2011-01-01

    The Cassini spacecraft dynamics-related telemetry during long Main Engine (ME) burns has indicated the presence of stable limit cycles between 0.03-0.04 Hz frequencies. These stable limit cycles cause the spacecraft to possess non-zero oscillating rates for extended periods of time. This indicates that the linear ME guidance control system does not model the complete dynamics of the spacecraft. In this study, we propose that the observed limit cycles in the spacecraft dynamics telemetry appear from a stable interaction between the unmodeled nonlinear elements in the ME guidance control system. Many nonlinearities in the control system emerge from translating the linear engine gimbal actuator (EGA) motion into a spacecraft rotation. One such nonlinearity comes from the gear backlash in the EGA system, which is the focus of this paper. The limit cycle characteristics and behavior can be predicted by modeling this gear backlash nonlinear element via a describing function and studying the interaction of this describing function with the overall dynamics of the spacecraft. The linear ME guidance controller and gear backlash nonlinearity are modeled analytically. The frequency, magnitude, and nature of the limit cycle are obtained from the frequency response of the ME guidance controller and nonlinear element. In addition, the ME guidance controller along with the nonlinearity is simulated. The simulation response contains a limit cycle with similar characterstics as predicted analytically: 0.03-0.04 Hz frequency and stable, sustained oscillations. The analytical and simulated limit cycle responses are compared to the flight telemetry for long burns such as the Saturn Orbit Insertion and Main Engine Orbit Trim Maneuvers. The analytical and simulated limit cycle characteristics compare well with the actual observed limit cycles in the flight telemetry. Both have frequencies between 0.03-0.04 Hz and stable oscillations. This work shows that the stable limit cycles occur

  8. Independent Orbiter Assessment (IOA): Assessment of the orbiter main propulsion system FMEA/CIL, volume 1

    Science.gov (United States)

    Slaughter, B. C.

    1988-01-01

    The results of the Independent Orbiter Assessment (IOA) of the Failure Modes and Effects Analysis (FMEA) and Critical Items List (CIL) are presented. The IOA effort first completed an analysis of the Main Propulsion System (MPS) hardware, generating draft failure modes and potential critical items. To preserve independence, this analysis was accomplished without reliance upon the results contained within the NASA FMEA/CIL documentation. The IOA results were then compared to available data from the Rockwell Downey/NASA JSC FMEA/CIL review. The Orbiter MPS is composed of the Propellant Management Subsystem (PMS) consisting of the liquid oxygen (LO2) and liquid hydrogen (LH2) subsystems and the helium subsystem. The PMS is a system of manifolds, distribution lines, and valves by which the liquid propellants pass from the External Tank to the Space Shuttle Main Engine (SSME). The helium subsystem consists of a series of helium supply tanks and their associated regulators, control valves, and distribution lines. Volume 1 contains the MPS description, assessment results, ground rules and assumptions, and some of the IOA worksheets.

  9. [Main principles of rapid diagnosis and intensive care in emergency states: their realization in expert systems].

    Science.gov (United States)

    Zislin, B D; Bazhenov, A M; Belkin, A A; Bazylev, S V; Badaev, F I; Trifonov, Iu O

    1997-01-01

    A retrospective analysis of 543 case histories over 1980-1990 in the town of Yekaterinburg and analysis of published data permitted the authors to single out the signs characterizing the most frequent syndromes requiring urgent intensive care. By either diagnostic value, these signs are distributed into main, accessory, and ruling out. An expert system has been created, making use of the productive-Freimont's approach to representing information on the basis of blurred multiplicities and ambiguous logics. The diagnosis was made by stages: first the main signs were analyzed, determining the severity of patient's status, then (after first aid was rendered) accessory and ruling out signs, which help make the diagnosis more precise. The system was tried in 231 patients, 102 of these with acute respiratory failure, 63 with acute hemodynamic insufficiency, and 66 with acute cerebral insufficiency. Primary diagnosis of the underlying syndrome was correct in 87-89% of cases, of the concomitant syndrome in 92-97%. Repeated evaluations (in 1-3 and 24 h) taking account of the time course of the symptoms and of the results of unsophisticated instrumental examinations increased the share of correct diagnoses to 92-96%.

  10. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10{sup -4} Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  11. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    International Nuclear Information System (INIS)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10 -4 Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  12. Imitative modeling automatic system Control of steam pressure in the main steam collector with the influence on the main Servomotor steam turbine

    Science.gov (United States)

    Andriushin, A. V.; Zverkov, V. P.; Kuzishchin, V. F.; Ryzhkov, O. S.; Sabanin, V. R.

    2017-11-01

    The research and setting results of steam pressure in the main steam collector “Do itself” automatic control system (ACS) with high-speed feedback on steam pressure in the turbine regulating stage are presented. The ACS setup is performed on the simulation model of the controlled object developed for this purpose with load-dependent static and dynamic characteristics and a non-linear control algorithm with pulse control of the turbine main servomotor. A method for tuning nonlinear ACS with a numerical algorithm for multiparametric optimization and a procedure for separate dynamic adjustment of control devices in a two-loop ACS are proposed and implemented. It is shown that the nonlinear ACS adjusted with the proposed method with the regulators constant parameters ensures reliable and high-quality operation without the occurrence of oscillations in the transient processes the operating range of the turbine loads.

  13. Main components and characteristics of landslide early warning systems operational worldwide

    Science.gov (United States)

    Piciullo, Luca; Cepeda, José

    2017-04-01

    During the last decades the number of victims and economic losses due to natural hazards are dramatically increased worldwide. The reason can be mainly ascribed to climate changes and urbanization in areas exposed at high level of risk. Among the many mitigation measures available for reducing the risk to life related to natural hazards, early warning systems certainly constitute a significant cost-effective option available to the authorities in charge of risk management and governance. The aim is to help and protect populations exposed to natural hazards, reducing fatalities when major events occur. Landslide is one of the natural hazards addressed by early warning systems. Landslide early warning systems (LEWSs) are mainly composed by the following four components: set-up, correlation laws, decisional algorithm and warning management. Within this framework, the set-up includes all the preliminary actions and choices necessary for designing a LEWS, such as: the area covered by the system, the types of landslides and the monitoring instruments. The monitoring phase provides a series of important information on different variables, considered as triggering factors for landslides, in order to define correlation laws and thresholds. Then, a decisional algorithm is necessary for defining the: number of warning levels to be employed in the system, decision making procedures, and everything else system managers may need for issuing warnings in different warning zones. Finally the warning management is composed by: monitoring and warning strategy; communication strategy; emergency plan and, everything connected to the social sphere. Among LEWSs operational worldwide, two categories can be defined as a function of the scale of analysis: "local" and "territorial" systems. The scale of analysis influences several actions and aspects connected to the design and employment of the system, such as: the actors involved, the monitoring systems, type of landslide phenomena

  14. Users of the main smartphone operating systems (iOS, Android) differ only little in personality.

    Science.gov (United States)

    Götz, Friedrich M; Stieger, Stefan; Reips, Ulf-Dietrich

    2017-01-01

    The increasingly widespread use of mobile phone applications (apps) as research tools and cost-effective means of vast data collection raises new methodological challenges. In recent years, it has become a common practice for scientists to design apps that run only on a single operating system, thereby excluding large numbers of users who use a different operating system. However, empirical evidence investigating any selection biases that might result thereof is scarce. Henceforth, we conducted two studies drawing from a large multi-national (Study 1; N = 1,081) and a German-speaking sample (Study 2; N = 2,438). As such Study 1 compared iOS and Android users across an array of key personality traits (i.e., well-being, self-esteem, willingness to take risks, optimism, pessimism, Dark Triad, and the Big Five). Focusing on Big Five personality traits in a broader scope, in addition to smartphone users, Study 2 also examined users of the main computer operating systems (i.e., Mac OS, Windows). In both studies, very few significant differences were found, all of which were of small or even tiny effect size mostly disappearing after sociodemographics had been controlled for. Taken together, minor differences in personality seem to exist, but they are of small to negligible effect size (ranging from OR = 0.919 to 1.344 (Study 1), ηp2 = .005 to .036 (Study 2), respectively) and may reflect differences in sociodemographic composition, rather than operating system of smartphone users.

  15. Formation of Main Elements of the System Motivation of Human Capital of Construction Enterprises

    Directory of Open Access Journals (Sweden)

    Malykhina Oksana M.

    2014-02-01

    Full Text Available The article focuses on a necessity to develop system motivation of human capital of construction enterprises. The author gives her own vision of this process on the basis of the system situational approach. The article identifies main requirements, which are made stronger with the necessity of establishment of the system motivation. It generalises the aggregate of components of internal and external environments, which form the structure, requirements to functioning and directions of managerial impact on the genesis of this process, which would facilitate achievement of long-term goals of construction enterprises on the basis of the innovation model of development. Efficiency of functioning of the system motivation of human capital of construction enterprises would depend on accounting in the process of its formation and development of such basic properties as: target orientation, synergetic integrity, availability of the internal organisational structure, ability to restore and resistance to influence of external environment factors, flexibility and dynamism and ecologo-socio-economic efficiency.

  16. Simplified Models for Analysis and Design of the Control System Main Loops of CAREM Reactor

    International Nuclear Information System (INIS)

    Etchepareborda, Andres; Flury, Celso

    2000-01-01

    The target of this work is to show a few models developed for control analysis and design of the reactor CAREM's main control loops within a broad range of power (between 40 % and 100%).By one side, it is shown the main features of a analytic model programed in MATLAB.This model is based on fitting steady state points at different power levels of the CAREM's RETRAN model.By the other side, it is shown linear models of black-box type denoting the perturbed behavior of the system for each level power point.These models are identified from temporal responses of CAREM's RETRAN model to perturbed input signals over the different steady power level points.Then the dynamics of these models are verified contrasting the temporal responses of the RETRAN model versus the responses of the MATLAB model and the identified models, in each steady power level point.Also are contrasting the frequency response of the linearization of MATLAB model versus the frequency response of the identified models, in each steady power level point.Either the MATLAB model as the identified models are good enough for the control analysis and design of the three main control loops.The MATLAB model has a few differences against the RETRAN model in the primary pressure output variable, that it must be taken into account in the design of this control loop if this model is used.The aim of these models is to represent in a satisfactory way the dynamics of the plant for a later control analysis and design of the control loops in a frequency range between 0.01 rad/seg and 0.3 rad/seg, and a power range between 40 % and 100 %

  17. Remote Visual Testing (RVT) for the diagnostic inspection of feedwater heaters

    International Nuclear Information System (INIS)

    Nugent, M.J.; Pellegrino, B.A.

    1993-01-01

    Feedwater heaters are an important component in the overall plant heat rate, reliability, availability, performance and maintenance considerations at power stations. The ability to diagnose heater problems in-situ properly can lead to: (1) Preventative plugging of damaged, but unfailed tubes; (2) In-place repair procedures; (3) Incorporation of corrective actions into replacement designs or heater/unit operations. The benefits and limitations of Non-Destructive Testing (NDT) on feedwater heaters are briefly reviewed. All Remote Visual Testing (RVT) including borescopes, fiberscopes, videoborescopes and Closed Circuit Television (CCTV) cameras are discussed along with currently accepted formats for documentation. The benefits of a comprehensive in-place inspection involving Remote Visual Testing are discussed in relationship to its diagnostic capabilities. The results of eight post-service heater inspections are discussed along with the root cause of failure of seven unique failure mechanisms. These inspections, including FWH access, RVT tool and data analysis, are detailed. 13 figs

  18. Power-feedwater temperature operating domain for Sbwr applying Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar M, L. A.; Quezada G, S.; Espinosa M, E. G.; Vazquez R, A.; Varela H, J. R.; Cazares R, R. I.; Espinosa P, G., E-mail: sequega@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2014-10-15

    In this work the analyses of the feedwater temperature effects on reactor power in a simplified boiling water reactor (Sbwr) applying a methodology based on Monte Carlo simulation is presented. The Monte Carlo methodology was applied systematically to establish operating domain, due that the Sbwr are not yet in operation, the analysis of the nuclear and thermal-hydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. The results show that the reactor power is inversely proportional to the temperature of the feedwater, reactor power changes at 8% when the feed water temperature changes in 8%. (Author)

  19. Power-feedwater temperature operating domain for Sbwr applying Monte Carlo simulation

    International Nuclear Information System (INIS)

    Aguilar M, L. A.; Quezada G, S.; Espinosa M, E. G.; Vazquez R, A.; Varela H, J. R.; Cazares R, R. I.; Espinosa P, G.

    2014-10-01

    In this work the analyses of the feedwater temperature effects on reactor power in a simplified boiling water reactor (Sbwr) applying a methodology based on Monte Carlo simulation is presented. The Monte Carlo methodology was applied systematically to establish operating domain, due that the Sbwr are not yet in operation, the analysis of the nuclear and thermal-hydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. The results show that the reactor power is inversely proportional to the temperature of the feedwater, reactor power changes at 8% when the feed water temperature changes in 8%. (Author)

  20. Leak before break evaluation for main steam piping system made of SA106 Gr.C

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Kyoung Mo; Jee, Kye Kwang; Pyo, Chang Ryul; Ra, In Sik [Korea Power Engineering Company, Seoul (Korea, Republic of)

    1997-04-01

    The basis of the leak before break (LBB) concept is to demonstrate that piping will leak significantly before a double ended guillotine break (DEGB) occurs. This is demonstrated by quantifying and evaluating the leak process and prescribing safe shutdown of the plant on the basis of the monitored leak rate. The application of LBB for power plant design has reduced plant cost while improving plant integrity. Several evaluations employing LBB analysis on system piping based on DEGB design have been completed. However, the application of LBB on main steam (MS) piping, which is LBB applicable piping, has not been performed due to several uncertainties associated with occurrence of steam hammer and dynamic strain aging (DSA). The objective of this paper is to demonstrate the applicability of the LBB design concept to main steam lines manufactured with SA106 Gr.C carbon steel. Based on the material properties, including fracture toughness and tensile properties obtained from the comprehensive material tests for base and weld metals, a parametric study was performed as described in this paper. The PICEP code was used to determine leak size crack (LSC) and the FLET code was used to perform the stability assessment of MS piping. The effects of material properties obtained from tests were evaluated to determine the LBB applicability for the MS piping. It can be shown from this parametric study that the MS piping has a high possibility of design using LBB analysis.

  1. The main microelements and phosphorus content of sediments formed in a drinking water supply system

    Directory of Open Access Journals (Sweden)

    Marina Valentukeviciene

    2016-11-01

    Full Text Available Groundwater is the only source for drinking water supply in Lithuania. Twenty water intakes exploiting Quaternary aquifers are operating in Vilnius City. The main aim of this study was to characterize the heavy metal content of internal pipeline sediments in the water supply network. It also provides a new insight into the accumulation of phosphorus and its variation in pipeline sediments in the study area. The results of this research reflect the level of heavy metals that accumulated during the water supply process. The main microelements detected were lead, nickel, zinc and copper. The research results will be useful for conducting preliminary evaluations of possible microelement accumulation in other similar water supply systems. The evaluation of water supply sediments is considered as one of the most important activities associated with a water safety approach. The results of this research indicate the dependence between phosphorus accumulation and Pb, Cr, Zn, Ni and Cu quantities in the internal sediments of water supply pipelines.

  2. THE MULTIPLE PRE-MAIN-SEQUENCE SYSTEM HBC 515 IN L1622

    International Nuclear Information System (INIS)

    Reipurth, Bo; Aspin, Colin; Herbig, George

    2010-01-01

    The bright pre-main-sequence star HBC 515 (HD 288313) located in the L1622 cometary cloud in Orion has been studied extensively with optical/infrared imaging and ultraviolet/optical/infrared spectroscopy. The spectra indicate that HBC 515 is a weakline T Tauri star (TTS) of spectral type K2V. Adaptive optics imaging in the K band reveals that HBC 515 is a binary with two equally bright components separated by 0.''5. A very faint third component is found 5'' to the northwest. Spitzer IRAC and MIPS observations show that at mid-infrared wavelengths this third source dominates the system, suggesting that it is a protostar still embedded in the nascent cloud of HBC 515. The close association of a weakline TTS with a newborn protostar in a multiple system is noteworthy. Two nearby TTSs are likely associated with the HBC 515 multiple system, and the dynamical evolution of the complex that would lead to such a configuration is considered.

  3. Limbic system, the main focus of dementia syndrome; A study with MRI and PET

    Energy Technology Data Exchange (ETDEWEB)

    Matsuzawa, Taiju [Morinosato Hospital, Atsugi, Kanagawa (Japan)

    1990-12-01

    Alzheimer disease and multi-infarct dementia are two entirely different diseases producing almost the same abnormalities as dementia syndrome. The statistical studies with MRI to locate the focus of dementia syndrome in the neocortex was an absolute failure. With MRI there is drastic atrophy and destruction of the amygdala and hippocampus suggesting the limbic system as the focus of dementia syndrome. Destruction of the limbic system in particular amygdala and hippocampus produced the functional obstruction brought about by the marked reduction in the glucose utilization with PET in the bilateral temporal, parietal and occipital association cortices. Although this type constitutes only about 1/5 of all dementia patients. It is considered the fundamental type of dementia syndrome. Aside from this, there is a type wherein simultaneous and symmetrical reductions in glucose utilization of the frontal association cortex and the motor association cortex in the anterior part of the neocortex. This is referred to as type II. It constitutes about 4/5 of all dementia patients which is far more than type I. Based on these results, it is thought that limbic system is the main focus of dementia syndrome. (author).

  4. Operating experience with LAMPF main beam lines instrumentation and control system

    International Nuclear Information System (INIS)

    van Dyck, O.B.; Harvey, A.; Howard, H.H.; Roeder, D.L.

    1975-01-01

    Instrumentation and control (I and C) for the Los Alamos Clinton P. Anderson Meson Physics Facility (LAMPF) main beam line is based upon central computer control through remote stations which provide input and output to most devices. Operating experience shows that the ability of the computer to give high-quality graphical presentation of the measurements enhances operator performance and instrument usefulness. Experience also shows that operator efficiency degrades rapidly with increasing instrument response time, that is, with increasing delay between the time a control is changed and the result can be observed. For this reason, instrumentation upgrade includes speeding up data acquisition and display times to under 10 s. Similarly, television-viewed phosphors are being retained where possible since their instantaneous response is very useful. Other upgrading of the instrumentation system is planned to improve data accuracy, reliability, redundancy, and instrument radiation tolerance. Past experience is being applied in adding or relocating devices to simplify tuning procedures. (U.S.)

  5. Use of an expert system data analysis manager for space shuttle main engine test evaluation

    Science.gov (United States)

    Abernethy, Ken

    1988-01-01

    The ability to articulate, collect, and automate the application of the expertise needed for the analysis of space shuttle main engine (SSME) test data would be of great benefit to NASA liquid rocket engine experts. This paper describes a project whose goal is to build a rule-based expert system which incorporates such expertise. Experiential expertise, collected directly from the experts currently involved in SSME data analysis, is used to build a rule base to identify engine anomalies similar to those analyzed previously. Additionally, an alternate method of expertise capture is being explored. This method would generate rules inductively based on calculations made using a theoretical model of the SSME's operation. The latter rules would be capable of diagnosing anomalies which may not have appeared before, but whose effects can be predicted by the theoretical model.

  6. TQC works in newly-built nuclear power plant and main electric power system plannings

    International Nuclear Information System (INIS)

    Akiyama, Yoshihisa; Kawakatsu, Tadashi; Hashimoto, Yasuo

    1985-01-01

    In the Kansai Electric Power Co., Inc., TQC has been introduced to solve such major problems in nuclear power generation as the securing of nuclear power reliability, the suppression of rises in the costs, the reduction in long periods of power failure and the promotion in siting of nuclear power plants. It is thus employed as a means of the ''creation of a slim and tough business constitution''. The state of activities in Kansai Electric are described in quality assurance of a newly-built nuclear power plant and in raising the reliability of the main electric power system to distribute the generated nuclear power and further the future prospects are explained. (Mori, K.)

  7. Vibration reduction at the engine-pumps assembly of the main moderator system

    International Nuclear Information System (INIS)

    Holostencu, Adriana; Dinu, Marius

    2005-01-01

    Problem of decreasing vibrations in the main moderator motor at Cernavoda NPP - Unit 1 is presented. The moderator pumps are of centrifugal, vertical, single stage and double suction type. Each pump is provided with a main motor (690 Kw, 50 Hz, 6 KV) capable of full rotative speed and a secondary motor, also known as 'pony motor' (15 Kw, 50 Hz, 380 V) capable of a quarter of the full speed. At starting-commissioning stage of Cernavoda NPP, the vibration level in the moderator pump - motor assembly had an average value of 6 mm/s with spurious peaks up to 8 mm/s. It has to be mentioned that operation with a high vibration level may lead to: - a premature wear of the motor bearings; - extra stresses and fatigue in the material of pump and associated pipes. In order to maintain vibration speed to the design limit, the NPP personnel have started investigations since 1997. The main activities were: - verification of the vibration measuring loops; - checking the torque of the bolts that hold the motor's case; - measuring the start-up and nominal currents of the motors in order to determine any phase unbalance; - adjusting the spring hangers from the pumps discharge with simultaneous monitoring of the motor vibration level; - installation of rubber pads in the gap between the motor lugs and the existing seismic supports. None of these actions revealed deviations from the installation requirements or operating parameters. In 1999, a contract with EUROTEST S. A and Stevenson was signed, in order to find a solution to reduce the vibration level. The EUROTEST/ Stevenson action plan contains de following main activities: 1. Creating a calculus model of the moderator system, based on the design drawings provided by Cernavoda NPP; 2. Preparing and performing the vibration measurements in various points of the system; 3. Analyzing the measurements results; 4. Calibrating the calculus model created in step 1, according with the field measurements; 5. Propose a solution to reduce

  8. STUDY ON THE MAIN ASPECT RELATED TO COSTS, GENERATED BY INTEGRATED SYSTEMS IN THE BUSINESS

    Directory of Open Access Journals (Sweden)

    CEAUSESCU AURELIAN IONUT

    2014-08-01

    Full Text Available Under the conditions in which the world for some time past threshold Internet era, many Romanian companies yet reducing management applications accounting programs to CIEL, Win a Mentor or ASCON. It is true that the statement should not be generalized, so that, however, the market Enterprise Resource Planning (ERP has evolved, but we must admit that we are still far from being connected to the reality generated by globalisation of technological markets. In other words, for developed countries in terms of technology, ERP constitute the pinnacle of development over five decades of the techniques of economic administration support beneficiary technologies of information. Main challenge consists of the economic integration of all processes and optimize available resources of a global market.. In our country, however, things are different, and about a tradition in Romanian develop IT systems we can't talk about, though without too much excitement..

  9. Qualitative and Quantitative Analysis of Organic Impurities in Feedwater of a Heat-Recovery Steam Generator

    Science.gov (United States)

    Chichirov, A. A.; Chichirova, N. D.; Filimonova, A. A.; Gafiatullina, A. A.

    2018-03-01

    In recent years, combined-cycle units with heat-recovery steam generators have been constructed and commissioned extensively in the European part of Russia. By the example of the Kazan Cogeneration Power Station no. 3 (TETs-3), an affiliate of JSC TGK-16, the specific problems for most power stations with combined-cycle power units that stem from an elevated content of organic impurities in the feedwater of the heat-recovery steam generator (HRSG) are examined. The HRSG is fed with highly demineralized water in which the content of organic carbon is also standardized. It is assumed that the demineralized water coming from the chemical water treatment department of TETs-3 will be used. Natural water from the Volga River is treated to produce demineralized water. The results of a preliminary analysis of the feedwater demonstrate that certain quality indices, principally, the total organic carbon, are above the standard values. Hence, a comprehensive investigation of the feedwater for organic impurities was performed, which included determination of their structure using IR and UV spectroscopy techniques, potentiometric measurements, and element analysis; determination of physical and chemical properties of organic impurities; and prediction of their behavior in the HRSG. The estimation of the total organic carbon revealed that it exceeded the standard values in all sources of water comprising the feedwater for the HRSG. The extracted impurities were humic substances, namely, a mixture of humic and fulvic acids in a 20 : 80 ratio, respectively. In addition, an analysis was performed of water samples taken at all intermediate stages of water treatment to study the behavior of organic substances in different water treatment processes. An analysis of removal of the humus substances in sections of the water treatment plant yielded the concentration of organic substances on the HRSG condensate. This was from 100 to 150 μg/dm3. Organic impurities in boiler water can induce

  10. Developing the optimum boiler water and feedwater treatment for fossil plants

    Energy Technology Data Exchange (ETDEWEB)

    Dooley, B [Electric Power Research Inst., Palo Alto, California (United States)

    1996-12-01

    Over the last two years a new set of cycle chemistry guidelines has been developed for each of the treatments used in fossil plants. These revisions have been based on research conducted over the last ten years, much at the international collaborative level. By careful selection and optimization of the boiler water and feedwater treatments, it will be possible to accrue large financial, maintenance, availability and performance improvements. (au) 14 refs.

  11. Device for the analysis of feedwater and condensation samples from power plants

    International Nuclear Information System (INIS)

    Mostofin, A.A.; Sorokina, N.S.

    1978-01-01

    An improved version of a device for automatic measurement of the salt and NH 3 contents of feedwater and condensate samples from nuclear power plants is described. Only one sample is required for determining both values. The invention proposes on the one hand to change the dimensions of a throttle opening and on the other to install a second measuring instrument (conductivity measuring instrument). (UWI) [de

  12. Evaluation of heatup and recovery in a loss of feedwater accident with multiple failure

    International Nuclear Information System (INIS)

    Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung

    1991-01-01

    A loss of feedwater accident with multiple failure has been studied in order to identify the potential severity of the accident when compared with the design basis accident in PWR. The PCS heatup and recovery mode in a LOFA with multiple failure was evaluated using the LOFT L9-1/L3-3 experiment. From experimental result, 4 separable subphase were identified and the associated phenomena were also addressed

  13. Water-hammer in the feed-water pipes for PWR steam generators

    International Nuclear Information System (INIS)

    Gonnet, Bernard; Leroy, Claude; Oullion, Jean; Yazidjian, J.-C.

    1979-01-01

    PWR boiler water feed pipes have been known for several years to be affected by violent water-hammer during start-ups and operation of the plant. In view of the varying results of corrective design modifications in America and Europe, FRAMATOME undertook an experimental research programme which resulted in the adoption of cruciform tubes on the feed-water distributor as the most reliable solution. Subsequent tests at Fessenheim I confirmed the effectiveness of this device [fr

  14. Interim status report on the revision of ASME PTC 12.1 -- closed feedwater heaters

    International Nuclear Information System (INIS)

    Stellern, J.L.; Hoobler, J.V.; Milton, J.W.; Welch, T.; Kona, C.; Thompson, H.N.; Tsou, J.L.

    1993-01-01

    The ASME Performance Test Code (PTC) 12.1-1978 for the performance testing of feedwater heaters is being revised extensively and updated. The committee anticipates that the final draft of the proposed Code will be ready for industry review in 1993. This Code revision will greatly enhance the usefulness and cost effectiveness of feedwater heater performance testing. This paper has been prepared to report on the progress of the committee and to disseminate information on the nature of the revision. Included in this paper are some of the notable changes intended for the Code. The most extensive change is the calculation method, which is described in step-by-step detail. An approach is also described for using ultrasonic flow techniques to test individual or split-string feedwater heaters, when flow nozzles are not available. Additionally some educational information on the use and limitations of ultrasonic measurement instrumentation is included. Discussion is also included on the required uncertainty analysis. 3 refs., 2 figs., 2 tabs

  15. THE QUADRUPLE PRE-MAIN-SEQUENCE SYSTEM LkCa 3: IMPLICATIONS FOR STELLAR EVOLUTION MODELS

    International Nuclear Information System (INIS)

    Torres, Guillermo; Latham, David W.; Ruíz-Rodríguez, Dary; Prato, L.; Wasserman, Lawrence H.; Badenas, Mariona; Schaefer, G. H.; Mathieu, Robert D.

    2013-01-01

    We report the discovery that the pre-main-sequence (PMS) object LkCa 3 in the Taurus-Auriga star-forming region is a hierarchical quadruple system of M stars. It was previously known to be a close (∼0.''5) visual pair, with one component being a moderately eccentric 12.94 day single-lined spectroscopic binary. A re-analysis of archival optical spectra complemented by new near-infrared (NIR) spectroscopy shows both visual components to be double lined; the second one has a period of 4.06 days and a circular orbit. In addition to the orbital elements, we determine optical and NIR flux ratios, effective temperatures, and projected rotational velocities for all four stars. Using existing photometric monitoring observations of the system that had previously revealed the rotational period of the primary in the longer-period binary, we also detect the rotational signal of the primary in the 4.06 day binary, which is synchronized with the orbital motion. With only the assumption of coevality, a comparison of all of these constraints with current stellar evolution models from the Dartmouth series points to an age of 1.4 Myr and a distance of 133 pc, consistent with previous estimates for the region and suggesting that the system is on the near side of the Taurus complex. Similar comparisons of the properties of LkCa 3 and the well-known quadruple PMS system GG Tau with the widely used models from the Lyon series for a mixing length parameter of α ML = 1.0 strongly favor the Dartmouth models

  16. Users of the main smartphone operating systems (iOS, Android differ only little in personality.

    Directory of Open Access Journals (Sweden)

    Friedrich M Götz

    Full Text Available The increasingly widespread use of mobile phone applications (apps as research tools and cost-effective means of vast data collection raises new methodological challenges. In recent years, it has become a common practice for scientists to design apps that run only on a single operating system, thereby excluding large numbers of users who use a different operating system. However, empirical evidence investigating any selection biases that might result thereof is scarce. Henceforth, we conducted two studies drawing from a large multi-national (Study 1; N = 1,081 and a German-speaking sample (Study 2; N = 2,438. As such Study 1 compared iOS and Android users across an array of key personality traits (i.e., well-being, self-esteem, willingness to take risks, optimism, pessimism, Dark Triad, and the Big Five. Focusing on Big Five personality traits in a broader scope, in addition to smartphone users, Study 2 also examined users of the main computer operating systems (i.e., Mac OS, Windows. In both studies, very few significant differences were found, all of which were of small or even tiny effect size mostly disappearing after sociodemographics had been controlled for. Taken together, minor differences in personality seem to exist, but they are of small to negligible effect size (ranging from OR = 0.919 to 1.344 (Study 1, ηp2 = .005 to .036 (Study 2, respectively and may reflect differences in sociodemographic composition, rather than operating system of smartphone users.

  17. Users of the main smartphone operating systems (iOS, Android) differ only little in personality

    Science.gov (United States)

    Stieger, Stefan; Reips, Ulf-Dietrich

    2017-01-01

    The increasingly widespread use of mobile phone applications (apps) as research tools and cost-effective means of vast data collection raises new methodological challenges. In recent years, it has become a common practice for scientists to design apps that run only on a single operating system, thereby excluding large numbers of users who use a different operating system. However, empirical evidence investigating any selection biases that might result thereof is scarce. Henceforth, we conducted two studies drawing from a large multi-national (Study 1; N = 1,081) and a German-speaking sample (Study 2; N = 2,438). As such Study 1 compared iOS and Android users across an array of key personality traits (i.e., well-being, self-esteem, willingness to take risks, optimism, pessimism, Dark Triad, and the Big Five). Focusing on Big Five personality traits in a broader scope, in addition to smartphone users, Study 2 also examined users of the main computer operating systems (i.e., Mac OS, Windows). In both studies, very few significant differences were found, all of which were of small or even tiny effect size mostly disappearing after sociodemographics had been controlled for. Taken together, minor differences in personality seem to exist, but they are of small to negligible effect size (ranging from OR = 0.919 to 1.344 (Study 1), ηp2 = .005 to .036 (Study 2), respectively) and may reflect differences in sociodemographic composition, rather than operating system of smartphone users. PMID:28467473

  18. Analysis of ultrasound propagation in high-temperature nuclear reactor feedwater to investigate a clamp-on ultrasonic pulse doppler flowmeter

    International Nuclear Information System (INIS)

    Tezuka, Kenichi; Mori, Michitsugu; Wada, Sanehiro; Aritomi, Masanori; Kikura, Hiroshige; Sakai, Yukihiro

    2008-01-01

    The flow rate of nuclear reactor feedwater is an important factor in the operation of a nuclear power reactor. Venturi nozzles are widely used to measure the flow rate. Other types of flowmeters have been proposed to improve measurement accuracy and permit the flow rate and reactor power to be increased. The ultrasonic pulse Doppler system is expected to be a candidate method because it can measure the flow profile across the pipe cross section, which changes with time. For accurate estimation of the flow velocity, the incidence angle of ultrasound entering the fluid should be estimated using Snell's law. However, evaluation of the ultrasound propagation is not straightforward, especially for a high-temperature pipe with a clamp-on ultrasonic Doppler flowmeter. The ultrasound beam path may differ from what is expected from Snell's law due to the temperature gradient in the wedge and variation in the acoustic impedance between interfaces. Recently, simulation code for ultrasound propagation has come into use in the nuclear field for nondestructive testing. This article analyzes and discusses ultrasound propagation, using 3D-FEM simulation code plus the Kirchhoff method, as it relates to flow profile measurement in nuclear reactor feedwater with the ultrasonic pulse Doppler system. (author)

  19. Simulation research on operation scheme of dissymmetrical main engine of CODOG propulsion system

    Directory of Open Access Journals (Sweden)

    HUANG Bin

    2018-02-01

    Full Text Available [Objectives] How to maintain propulsion capability in a CODOG propulsion system damage situation has important significance. [Methods] A ‘Hull-Engine-CPP-Rudder’ simulation model of a CODOG marine power plant is established on Simulink using the modularized method, and a dissymmetrical main engine urgent working mode is proposed and simulated. [Results] The results show that in the dissymmetrical working mode, two different engines cannot work simultaneously at designed capacity. However, by adjusting the pitch of the CPP, one engine can work at designed capacity and the other can work at partial load capacity; under this working mode, if high speed is demanded, the gas turbine should work at designed capacity. The CPP pitch driven by diesel should be maintained at a high value near the maximum. The maximum speed of this working mode is 84.4% of the designed speed, which is higher than the speed of the single shaft working mode driven by a gas turbine. [Conclusions] The research results of this paper can provide useful references for the design of ship propulsion systems.

  20. Effectiveness of experimental system in children’s mastering of main manual motor skills

    Directory of Open Access Journals (Sweden)

    L. L. Galamandjuk

    2015-07-01

    Full Text Available Purpose: determine effectiveness of simulated system of pre-school age children motor functioning’s preventive development for mastering of main movements for manual skills. Material: in the research five years’ children (control group n=150 and experimental group n=120 participated. Results: it was determined that transition from uncomfortable to comfortable for a child conditions of exercises’ fulfillment facilitates quicker formation of required motor program. It is connected with the fact that, independent on orientation of manual motor asymmetry progressing of semi-spheres’ interaction takes place. This interaction is an important condition of increasing of child functioning’s effectiveness in different aspects. Achievement of such result was also facilitated by physical exercises, which children practiced at home. Functioning of physical culture instructor and kindergarten teachers was also important: they formed parents’ conscious position concerning importance of such trainings; recommended effective means and methods. Conclusions: application of the offered system ensures much better result than traditional approach to this problem. It is one of keys to prevention of negative tendencies in development of pre-school age children.

  1. Community Energy Systems and the Law of Public Utilities. Volume Twenty-one. Maine

    Energy Technology Data Exchange (ETDEWEB)

    Feurer, D.A.; Weaver, C.L.

    1981-01-01

    A detailed description is given of the laws and programs of the State of Maine governing the regulation of public energy utilities, the siting of energy generating and transmission facilities, the municipal franchising of public energy utilities, and the prescription of rates to be charged by utilities including attendant problems of cost allocations, rate base and operating expense determinations, and rate of return allowances. These laws and programs are analyzed to identify impediments which they may present to the implementation of Integrated Community Energy Systems (ICES). This report is one of fifty-one separate volumes which describe such regulatory programs at the Federal level and in each state as background to the report entitled Community Energy Systems and the Law of Public Utilities - Volume One: An Overview. This report also contains a summary of a strategy described in Volume One - An Overview for overcoming these impediments by working within the existing regulatory framework and by making changes in the regulatory programs to enhance the likelihood of ICES implementation.

  2. HTTR hydrogen production system. Structure and main specifications of mock-up test facility (Contract research)

    International Nuclear Information System (INIS)

    Kato, Michio; Aita, Hideki; Inagaki, Yoshiyuki; Hayashi, Koji; Ohashi, Hirofumi; Sato, Hiroyuki; Iwatsuki, Jin; Takada, Shoji; Inaba, Yoshitomo

    2007-03-01

    The mock-up test facility was fabricated to investigate performance of the steam generator for mitigation of the temperature fluctuation of helium gas and transient behavior of the hydrogen production system for HTTR and to obtain experimental data for verification of a dynamic analysis code. The test facility has an approximate hydrogen production capacity of 120Nm 3 /h and the steam reforming process of methane; CH 4 +H 2 O=3H 2 +CO, was used for hydrogen production of the test facility. An electric heater was used as a heat source instead of the reactor in order to heat helium gas up to 880degC (4MPa) at the chemical reactor inlet which is the same temperature as the HTTR hydrogen production system. Fabrication of the test facility was completed in February in 2002, and seven cycle operations were carried out from March in 2002 to December in 2004. This report describes the structure and main specifications of the test facility. (author)

  3. Genetic dissection of pheromone processing reveals main olfactory system-mediated social behaviors in mice.

    Science.gov (United States)

    Matsuo, Tomohiko; Hattori, Tatsuya; Asaba, Akari; Inoue, Naokazu; Kanomata, Nobuhiro; Kikusui, Takefumi; Kobayakawa, Reiko; Kobayakawa, Ko

    2015-01-20

    Most mammals have two major olfactory subsystems: the main olfactory system (MOS) and vomeronasal system (VNS). It is now widely accepted that the range of pheromones that control social behaviors are processed by both the VNS and the MOS. However, the functional contributions of each subsystem in social behavior remain unclear. To genetically dissociate the MOS and VNS functions, we established two conditional knockout mouse lines that led to either loss-of-function in the entire MOS or in the dorsal MOS. Mice with whole-MOS loss-of-function displayed severe defects in active sniffing and poor survival through the neonatal period. In contrast, when loss-of-function was confined to the dorsal MOB, sniffing behavior, pheromone recognition, and VNS activity were maintained. However, defects in a wide spectrum of social behaviors were observed: attraction to female urine and the accompanying ultrasonic vocalizations, chemoinvestigatory preference, aggression, maternal behaviors, and risk-assessment behaviors in response to an alarm pheromone. Functional dissociation of pheromone detection and pheromonal induction of behaviors showed the anterior olfactory nucleus (AON)-regulated social behaviors downstream from the MOS. Lesion analysis and neural activation mapping showed pheromonal activation in multiple amygdaloid and hypothalamic nuclei, important regions for the expression of social behavior, was dependent on MOS and AON functions. Identification of the MOS-AON-mediated pheromone pathway may provide insights into pheromone signaling in animals that do not possess a functional VNS, including humans.

  4. Long-Range Untethered Real-Time Live Gas Main Robotic Inspection System

    Energy Technology Data Exchange (ETDEWEB)

    Hagen Schempf; Daphne D' Zurko

    2004-10-31

    Under funding from the Department of Energy (DOE) and the Northeast Gas Association (NGA), Carnegie Mellon University (CMU) developed an untethered, wireless remote controlled inspection robot dubbed Explorer. The project entailed the design and prototyping of a wireless self-powered video-inspection robot capable of accessing live 6- and 8-inch diameter cast-iron and steel mains, while traversing turns and Ts and elbows under real-time control with live video feedback to an operator. The design is that of a segmented actively articulated and wheel-leg powered robot design, with fisheye imaging capability and self-powered battery storage and wireless real-time communication link. The prototype was functionally tested in an above ground pipe-network, in order to debug all mechanical, electrical and software subsystems, and develop the necessary deployment and retrieval, as well as obstacle-handling scripts. A pressurized natural gas test-section was used to certify it for operation in natural gas at up to 60 psig. Two subsequent live-main field-trials in both cast-iron and steel pipe, demonstrated its ability to be safely launched, operated and retrieved under real-world conditions. The system's ability to safely and repeatably exidrecover from angled and vertical launchers, traverse multi-thousand foot long pipe-sections, make T and varied-angle elbow-turns while wirelessly sending live video and handling command and control messages, was clearly demonstrated. Video-inspection was clearly shown to be a viable tool to understand the state of this critical buried infrastructure, irrespective of low- (cast-iron) or high-pressure (steel) conditions. This report covers the different aspects of specifications, requirements, design, prototyping, integration and testing and field-trialing of the Explorer platform.

  5. Applications of equivalent linearization approaches to nonlinear piping systems

    International Nuclear Information System (INIS)

    Park, Y.; Hofmayer, C.; Chokshi, N.

    1997-01-01

    The piping systems in nuclear power plants, even with conventional snubber supports, are highly complex nonlinear structures under severe earthquake loadings mainly due to various mechanical gaps in support structures. Some type of nonlinear analysis is necessary to accurately predict the piping responses under earthquake loadings. The application of equivalent linearization approaches (ELA) to seismic analyses of nonlinear piping systems is presented. Two types of ELA's are studied; i.e., one based on the response spectrum method and the other based on the linear random vibration theory. The test results of main steam and feedwater piping systems supported by snubbers and energy absorbers are used to evaluate the numerical accuracy and limitations

  6. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  7. Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

    International Nuclear Information System (INIS)

    Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

    2012-01-01

    KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described

  8. Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

    Energy Technology Data Exchange (ETDEWEB)

    Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described.

  9. Replacement of heavy components of the Main Primary System (MPS). Recent innovations made by Framatome ANP

    International Nuclear Information System (INIS)

    Chanussot, J.M.; Thevenet, R.

    2006-01-01

    The replacement of heavy components is the result of widespread stress corrosion of Inconel 600 (and alloys 82/182) in the primary system. Following the corrosion of steam generator tubes, which led to the first steam generator replacement (SGR) operations, work has begun on reactor vessel head replacements (RVHR) and pressurizer replacements, pending treatment of the dissimilar butt welds (DBW). The replacement of sections of the Main Primary System is one way of optimizing plant lifetime besides preventing stress corrosion in primary components and combating phenomena as thermal fatigue or certain metallurgical weaknesses. The increase in the number of major operations carried out in the reactor building on PWR and BWR reactor geometries has called for major technical innovations. AREVA, has over the past twenty years carried out more than forty SGRs throughout the world, including the supply of 115 steam generators of all types. This equates to over 60% of all SGR operations performed in Europe and the USA. Today, most SGR operations are performed in the USA, though some still continue in Europe and Asia. Operations have been planned until 2015. The first RVHR operations were carried out in France in 1993, and have been widespread in Europe since 1996 (Spain and Belgium), and in the USA since 2002 (Davis Besse). Since 1993 AREVA, through its subsidiaries Framatome ANP, Jeumont SA and SGT, has performed 72 RVHRs, i.e. most of those carried out in Europe and the USA. Today, RVHR operations take place throughout the world and have been planned until 2014. The first pressurizer replacement took place in the fall of 2005 in Saint Lucie, USA. Others are to follow (Fort Calhoun, Millestone). A significant number of partial or complete pressurizer replacements are now being scheduled. Special focus must be given to this new operation, as a decision is required between the following types of replacements, taking into account the number of pressurizer heaters and the

  10. Corrosion behaviour of a stream generator tube material in simulated steam generator feedwater containing chlorides and sulphates

    Energy Technology Data Exchange (ETDEWEB)

    Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Sirkiae, P.; Yliniemi, K. [VTT Manufacturing Technology, Espoo (Finland); Buddas, T.; Halin, M.; Tompuri, K. [Fortum Power and Heat Oy, Loviisa Power Plant (Finland)

    2002-07-01

    The goal of the present work has been to assess the effect of relatively high concentrations of anionic impurities (Cl{sup -}, SO{sub 4}{sup 2-}) on the corrosion behaviour of Ti-stabilised stainless steel SG tubes in simulated steam generator feed-water. The main observations of this work can be summarised as follows: Sulphate ions seem to be more aggressive than chloride ions towards the primary passive film on 08X18H10T stainless steel. The results may indicate that it is more important to have a low concentration of sulphate ions than of chloride ions in secondary side water when the effects of chemical conditions on tube degradation are considered. The presence of chloride ions seems to weaken the detrimental effect of sulphate ions on the stability of oxide films growing on 08X18H10T stainless steel. No localised corrosion features of 08X18H10T stainless steel were detected in the voltammetric and impedance measurements in solutions containing up to 5000 ppb sulphates, chlorides or both of the anions. (authors)

  11. Corrosion behaviour of a stream generator tube material in simulated steam generator feedwater containing chlorides and sulphates

    International Nuclear Information System (INIS)

    Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Sirkiae, P.; Yliniemi, K.; Buddas, T.; Halin, M.; Tompuri, K.

    2002-01-01

    The goal of the present work has been to assess the effect of relatively high concentrations of anionic impurities (Cl - , SO 4 2- ) on the corrosion behaviour of Ti-stabilised stainless steel SG tubes in simulated steam generator feed-water. The main observations of this work can be summarised as follows: Sulphate ions seem to be more aggressive than chloride ions towards the primary passive film on 08X18H10T stainless steel. The results may indicate that it is more important to have a low concentration of sulphate ions than of chloride ions in secondary side water when the effects of chemical conditions on tube degradation are considered. The presence of chloride ions seems to weaken the detrimental effect of sulphate ions on the stability of oxide films growing on 08X18H10T stainless steel. No localised corrosion features of 08X18H10T stainless steel were detected in the voltammetric and impedance measurements in solutions containing up to 5000 ppb sulphates, chlorides or both of the anions. (authors)

  12. Bright future of photovoltaic-hybrid systems as main option for electricity generation in remote communities

    Energy Technology Data Exchange (ETDEWEB)

    Zahedi, Ahmad [Solar Energy Applications Research Group (Australia)

    2000-07-01

    The most common power option for remotely located communities, facilities, schools, etc., is the engine generator powered by diesel fuel. Over the past 15 years, many remote communities with limited and costly site access for maintenance and fuel delivery have had their engine-based power systems modified to photovoltaic hybrid power systems. As a result, hybrid power systems with photovoltaic as the main generator are becoming the preferred power option. The reasons for this change are simple: the engine-based power systems require regular oil and filter changes (in average after 150 hrs of operation); the maintenance cost is relatively high; the cost of travel to and from the site to perform maintenance is restricted during certain time of the year and can be more expensive than the actual maintenance itself. Photovoltaic generators are gradually replacing the diesel generators and thus are becoming the primary source in remote communities. As electricity is required for 24 hours of operation and photovoltaic are not able to generate power for 24 h, batteries are added to the system as storage units, and the diesel generators are used as a back-up power supply. The objective of this paper is to present the results obtained from a study which has been carried out on a PV-hybrid power system from the desired performance point of view. [Spanish] La opcion mas comun de energia para las comunidades, instalaciones, escuelas, etc. localizadas en lugares remotos, es el generador que utiliza diesel como combustible. En los ultimos 15 anos, muchas comunidades remotas con acceso limitado y costoso para el mantenimiento y la entrega de combustible han modificado sus sistemas de energia basados en motores por sistemas de energia hibridos fotovoltaicos. Como resultado, los sistemas hibridos de energia con generadores fotovoltaicos como principal generador se estan convirtiendo en la opcion preferida de generacion de electricidad. Las razones para este cambio son simples: los

  13. Time Variability of the Dust Sublimation Zones in Pre-Main Sequence Disk Systems

    Science.gov (United States)

    Sitko, Michael L.; Carpenter, W. J.; Grady, C. A.; Russel, R. W.; Lynch, D. K.; Rudy, R. J.; Mazuk, S. M.; Venturini, C. C.; Kimes, R. L.; Beerman, L. C.; hide

    2007-01-01

    The dust sublimation zone (DSZ) is the region of pre-main sequence (PMS) disks where dust grains most easily anneal, sublime, and condense out of the gas. Because of this, it is a location where crystalline material may be enhanced and redistributed throughout the rest of the disk. A decade-long program to monitor the thermal emission of the grains located in this region demonstrates that large changes in emitted flux occur in many systems. Changes in the thermal emission between 3 and 13.5 microns were observed in HD 31648 (MWC 480), HD 163296 (MWC 275), and DG Tau. This emission is consistent with it being produced at the DSZ, where the transition from a disk of gas to one of gas+dust occurs. In the case of DG Tau, the outbursts were accompanied by increased emission on the 10 micron silicate band on one occasion, while on another occasion it went into absorption. This requires lofting of the material above the disk into the line of sight. Such changes will affect the determination of the inner disk structure obtained through interferometry measurements, and this has been confirmed in the case of HD 163296. Cyclic variations in the heating of the DSZ will lead to the annealing of large grains, the sublimation of smaller grains, possibly followed by re-condensation as the zone enters a cooling phase. Lofting of dust above the disk plane, and outward acceleration by stellar winds and radiation pressure, can re-distribute the processed material to cooler regions of the disk, where cometesimals form. This processing is consistent with the detection of the preferential concentration of large crystalline grains in the inner few AU of PMS disks using interferometric spectroscopy with the VLTI.

  14. Hydrothermal carbonization (HTC) of wheat straw: influence of feedwater pH prepared by acetic acid and potassium hydroxide.

    Science.gov (United States)

    Reza, M Toufiq; Rottler, Erwin; Herklotz, Laureen; Wirth, Benjamin

    2015-04-01

    In this study, influence of feedwater pH (2-12) was studied for hydrothermal carbonization (HTC) of wheat straw at 200 and 260°C. Acetic acid and KOH were used as acidic and basic medium, respectively. Hydrochars were characterized by elemental and fiber analyses, SEM, surface area, pore volume and size, and ATR-FTIR, while HTC process liquids were analyzed by HPLC and GC. Both hydrochar and HTC process liquid qualities vary with feedwater pH. At acidic pH, cellulose and elemental carbon increase in hydrochar, while hemicellulose and pseudo-lignin decrease. Hydrochars produced at pH 2 feedwater has 2.7 times larger surface area than that produced at pH 12. It also has the largest pore volume (1.1 × 10(-1) ml g(-1)) and pore size (20.2 nm). Organic acids were increasing, while sugars were decreasing in case of basic feedwater, however, phenolic compounds were present only at 260°C and their concentrations were increasing in basic feedwater. Copyright © 2015 Elsevier Ltd. All rights reserved.

  15. An estimation of reactor thermal power uncertainty using UFM-based feedwater flow rate in nuclear power plants

    International Nuclear Information System (INIS)

    Byung Ryul Jung; Ho Cheol Jang; Byung Jin Lee; Se Jin Baik; Woo Hyun Jang

    2005-01-01

    Most of Pressurized Water Reactors (PWRs) utilize the venturi meters (VMs) to measure the feedwater (FW) flow rate to the steam generator in the calorimetric measurement, which is used in the reactor thermal power (RTP) estimation. However, measurement drifts have been experienced due to some anomalies on the venturi meter (generally called the venturi meter fouling). The VM's fouling tends to increase the measured pressure drop across the meter, which results in indication of increased feedwater flow rate. Finally, the reactor thermal power is overestimated and the actual reactor power is to be reduced to remain within the regulatory limits. To overcome this VM's fouling problem, the Ultrasonic Flow Meter (UFM) has recently been gaining attention in the measurement of the feedwater flow rate. This paper presents the applicability of a UFM based feedwater flow rate in the estimation of reactor thermal power uncertainty. The FW and RTP uncertainties are compared in terms of sensitivities between the VM- and UFM-based feedwater flow rates. Data from typical Optimized Power Reactor 1000 (OPR1000) plants are used to estimate the uncertainty. (authors)

  16. LOFT/LP-FW-1, Loss of Fluid Test, PWR Response to Loss-of-Feedwater Transient

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The first OECD LOFT experiment was conducted on February 20, 1983. It was designed to evaluate the generic PWR system response during a complete loss-of-feedwater transient. The objective of the experiment was to investigate the performance of primary 'feed and bleed' using a 'bleed' from the PORV and 'feed' from the HPIS to provide decay heat removal and system pressure reduction while maintaining the primary coolant inventory. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  17. CRBRP decay heat removal systems

    International Nuclear Information System (INIS)

    Hottel, R.E.; Louison, R.; Boardman, C.E.; Kiley, M.J.

    1977-01-01

    The Decay Heat Removal Systems for the Clinch River Breeder Reactor Plant (CRBRP) are designed to adequately remove sensible and decay heat from the reactor following normal shutdown, operational occurrences, and postulated accidents on both a short term and a long term basis. The Decay Heat Removal Systems are composed of the Main Heat Transport System, the Main Condenser and Feedwater System, the Steam Generator Auxiliary Heat Removal System (SGAHRS), and the Direct Heat Removal Service (DHRS). The overall design of the CRBRP Decay Heat Removal Systems and the operation under normal and off-normal conditions is examined. The redundancies of the system design, such as the four decay heat removal paths, the emergency diesel power supplies, and the auxiliary feedwater pumps, and the diversities of the design such as forced circulation/natural circulation and AC Power/DC Power are presented. In addition to overall design and system capabilities, the detailed designs for the Protected Air Cooled Condensers (PACC) and the Air Blast Heat Exchangers (ABHX) are presented

  18. Procedures for the design of the main mechanical components of a wind system; Dimensionamento dos componentes mecanicos principais de aerogeradores

    Energy Technology Data Exchange (ETDEWEB)

    Hirata, M.H.; Marco Filho, F. de [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia

    1990-12-31

    Procedures for the design of the main mechanical components of a wind system were developed. One of the main concerns was related to the possibility of its use in small micro-computers. This goal was reached and an APPLE II computer was used. The resulting algorithm permits a friendly interaction between man and machine. 5 refs., 12 figs

  19. Visual Representations of Microcosm in Textbooks of Chemistry: Constructing a Systemic Network for Their Main Conceptual Framework

    Science.gov (United States)

    Papageorgiou, George; Amariotakis, Vasilios; Spiliotopoulou, Vasiliki

    2017-01-01

    The main objective of this work is to analyse the visual representations (VRs) of the microcosm depicted in nine Greek secondary chemistry school textbooks of the last three decades in order to construct a systemic network for their main conceptual framework and to evaluate the contribution of each one of the resulting categories to the network.…

  20. AUTOMATIC CONTROL SYSTEM FOR REGULATED HIGH TEMPERATURE MAIN COMBUSTION CHAMBER OF MANEUVERABLE AIRCRAFT MULTIMODE GAS TURBINE ENGINE

    Directory of Open Access Journals (Sweden)

    T. V. Gras’Ko

    2014-01-01

    Full Text Available The paper describes choosing and substantiating the control laws, forming the appearance the automatic control system for regulated high temperature main combustion chamber of maneuverable aircraft multimode gas turbine engine aimed at sustainable and effective functioning of main combustion chamber within a broad operation range.

  1. Modeling soil organic carbon dynamics and their driving factors in the main global cereal cropping systems

    Science.gov (United States)

    Wang, Guocheng; Zhang, Wen; Sun, Wenjuan; Li, Tingting; Han, Pengfei

    2017-10-01

    Changes in the soil organic carbon (SOC) stock are determined by the balance between the carbon input from organic materials and the output from the decomposition of soil C. The fate of SOC in cropland soils plays a significant role in both sustainable agricultural production and climate change mitigation. The spatiotemporal changes of soil organic carbon in croplands in response to different carbon (C) input management and environmental conditions across the main global cereal systems were studied using a modeling approach. We also identified the key variables that drive SOC changes at a high spatial resolution (0.1° × 0.1°) and over a long timescale (54 years from 1961 to 2014). A widely used soil C turnover model (RothC) and state-of-the-art databases of soil and climate variables were used in the present study. The model simulations suggested that, on a global average, the cropland SOC density increased at annual rates of 0.22, 0.45 and 0.69 Mg C ha-1 yr-1 under crop residue retention rates of 30, 60 and 90 %, respectively. Increasing the quantity of C input could enhance soil C sequestration or reduce the rate of soil C loss, depending largely on the local soil and climate conditions. Spatially, under a specific crop residue retention rate, relatively higher soil C sinks were found across the central parts of the USA, western Europe, and the northern regions of China. Relatively smaller soil C sinks occurred in the high-latitude regions of both the Northern and Southern hemispheres, and SOC decreased across the equatorial zones of Asia, Africa and America. We found that SOC change was significantly influenced by the crop residue retention rate (linearly positive) and the edaphic variable of initial SOC content (linearly negative). Temperature had weak negative effects, and precipitation had significantly negative impacts on SOC changes. The results can help guide carbon input management practices to effectively mitigate climate change through soil C

  2. Modeling soil organic carbon dynamics and their driving factors in the main global cereal cropping systems

    Directory of Open Access Journals (Sweden)

    G. Wang

    2017-10-01

    Full Text Available Changes in the soil organic carbon (SOC stock are determined by the balance between the carbon input from organic materials and the output from the decomposition of soil C. The fate of SOC in cropland soils plays a significant role in both sustainable agricultural production and climate change mitigation. The spatiotemporal changes of soil organic carbon in croplands in response to different carbon (C input management and environmental conditions across the main global cereal systems were studied using a modeling approach. We also identified the key variables that drive SOC changes at a high spatial resolution (0.1°  ×  0.1° and over a long timescale (54 years from 1961 to 2014. A widely used soil C turnover model (RothC and state-of-the-art databases of soil and climate variables were used in the present study. The model simulations suggested that, on a global average, the cropland SOC density increased at annual rates of 0.22, 0.45 and 0.69 Mg C ha−1 yr−1 under crop residue retention rates of 30, 60 and 90 %, respectively. Increasing the quantity of C input could enhance soil C sequestration or reduce the rate of soil C loss, depending largely on the local soil and climate conditions. Spatially, under a specific crop residue retention rate, relatively higher soil C sinks were found across the central parts of the USA, western Europe, and the northern regions of China. Relatively smaller soil C sinks occurred in the high-latitude regions of both the Northern and Southern hemispheres, and SOC decreased across the equatorial zones of Asia, Africa and America. We found that SOC change was significantly influenced by the crop residue retention rate (linearly positive and the edaphic variable of initial SOC content (linearly negative. Temperature had weak negative effects, and precipitation had significantly negative impacts on SOC changes. The results can help guide carbon input management practices to

  3. Interim Corrective Measures Work Plan for the Expanded Bioventing System Eglin Main Base Old Fire Training Area

    National Research Council Canada - National Science Library

    1997-01-01

    This interim corrective measures work plan (ICM work plan) presents the scope for an expanded bioventing system for in situ treatment of fuel-contaminated soils at the Eglin Main Base Old Fire Training Area (old Eglin FTA...

  4. Analysis of man-machine interaction for control and display system in main control room of light water reactor

    International Nuclear Information System (INIS)

    Santosa, Kussigit; Supriatna, Piping; Karlina, Itjeu; Widagdo, Suharyo; Darlis; Sudiono, Bambang

    1998-01-01

    One of potential hazard in Nuclear Power Plant is the failure of its operation. The accident or operation failure in the reactor must be concerned event its probability is low. The important thing should be concerned is 'Analysis of Man-Machine Interaction (MMI) for Control and Display System in Main Control Room (MCR) of Nuclear Power Reactor', especially LWR type. Control and Display System in MCR of Reactor is the main part of MMI link process in Reactor MCR work system. Signal from display system showed performance process in reactor, while this signal will be received by operator. This signal will be described through central nerve for making decision what kind must be done. Then the operator manage the next process of reactor operation through control system. So by knowing Analysis of Man-Machine Interaction for Control and Display System in Main Control Room of Power Reactor, we can understand human error probability of the operator in reactor operation

  5. Characterization of Pump-Induced Acoustics in Space Launch System Main Propulsion System Liquid Hydrogen Feedline Using Airflow Test Data

    Science.gov (United States)

    Eberhart, C. J.; Snellgrove, L. M.; Zoladz, T. F.

    2015-01-01

    High intensity acoustic edgetones located upstream of the RS-25 Low Pressure Fuel Turbo Pump (LPFTP) were previously observed during Space Launch System (STS) airflow testing of a model Main Propulsion System (MPS) liquid hydrogen (LH2) feedline mated to a modified LPFTP. MPS hardware has been adapted to mitigate the problematic edgetones as part of the Space Launch System (SLS) program. A follow-on airflow test campaign has subjected the adapted hardware to tests mimicking STS-era airflow conditions, and this manuscript describes acoustic environment identification and characterization born from the latest test results. Fluid dynamics responsible for driving discrete excitations were well reproduced using legacy hardware. The modified design was found insensitive to high intensity edgetone-like discretes over the bandwidth of interest to SLS MPS unsteady environments. Rather, the natural acoustics of the test article were observed to respond in a narrowband-random/mixed discrete manner to broadband noise thought generated by the flow field. The intensity of these responses were several orders of magnitude reduced from those driven by edgetones.

  6. Ocean observing systems support operational forecasts for the timing of Maine's lobster fishery

    Science.gov (United States)

    Mills, K.; Hernandez, C.; Pershing, A. J.

    2016-02-01

    American lobster supports one of the most valuable fisheries in the United States, with a landed value in 2013 exceeding $460M. Although US lobstermen are free to fish throughout the year, the New England climate, lobster biology and fleet dynamics lead to a strong annual cycle with catch rates rising rapidly in early summer and landings peaking in late summer. When this annual cycle is disrupted, it can impact the supply and ultimately the price of lobsters. During the record warm conditions in 2012, the rise in catch rates occurred three weeks ahead of normal. Combined with higher than normal landings from the spring Canadian fishery, the early and high volume landings in 2012 led to a collapse in price that severely stressed the U. S. fishery, especially in Maine where over 85% of the landings occur. Based on this experience, we have been developing seasonal forecasts of the phenology of Maine lobster landings. Using temperatures at 50m from four NERACOOS buoys in the Gulf of Maine, we can reliably forecast the date when the Maine lobster fishery will `turn on' for the year, with prediction accuracy peaking in April. The high-landings period normally starts in July, and the 2-3 month lead-time provides some advance warning to dealers and processors of when their capacity needs to be ready and to fishermen of potential supply chain and market impacts such as we observed in 2012. We are currently working towards finer-scale regional forecasts along the Maine coast that may include other features that will provide information to help the lobster industry adapt to the rapid changes that are underway in the Gulf of Maine.

  7. Life cycle management, design review, and condition assessment of feedwater heaters

    Energy Technology Data Exchange (ETDEWEB)

    Gammage, D.; Idvorian, N. [Babcock & Wilcox Canada Ltd., Cambridge, Ontario (Canada)

    2012-07-01

    OPEX from both the Nuclear and Fossil Power Generation Industries shows that Feedwater Heaters (FWHs) are subject to several degradation mechanisms and that this degradation commonly leads to replacement of these vessels in order to ensure reliable, efficient operation of the plants. Loss of feedwater heating will impact plant thermal performance. In response to inspection results showing on-going degradation as well as other factors, B&W Canada completed a project in conjunction with a US PWR utility to review the design, condition, and Life Cycle Management of their FWHs. This project involved a multi-disciplinary approach in order to consider all aspects of the FWHs in order to provide insight into the Life Cycle Management Plan (LCMP) so that the FWHs can be operated reliably into the future and so that adequate inspections can be conducted in order to produce a detailed condition assessment. The utility was interested in evaluating their FWH LCMP to determine if it was adequate in its requirements to enable reliable, leak-free operation of their FWH equipment. As inputs to this evaluation, it was required that B&W Canada evaluate both confirmed and plausible degradation mechanisms. They also required that the thermal hydraulic and functional design be evaluated for their particular FWHs. It was important to also incorporate industry OPEX in order to provide proper trending information for tube plugging. Out of this evaluation there were several findings and recommendations that could be used to update the utilities’ LCMP as it was apparent that the current version may not be truly reflective of the current condition of the equipment or of current industry OPEX of such FWHs. Several recommendations came from this evaluation, the most significant were: • Performing thermal/hydraulic, FIV (flow-induced vibration), and tube/shell interaction calculations to determine how the FWHs operate and how their performance can change over time as a function of tube

  8. Power-feedwater enthalpy operating domain for SBWR applying Monte Carlo simulation

    International Nuclear Information System (INIS)

    Quezada-Garcia, S.; Espinosa-Martinez, E.-G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G.

    2014-01-01

    In this work the analyses of the feedwater enthalpy effects on reactor power in a simplified boiling water reactor (SBWR) applying a methodology based on Monte Carlo's simulation (MCS), is presented. The MCS methodology was applied systematically to establish operating domain, due that the SBWR are not yet in operation, the analysis of the nuclear and thermalhydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. (author)

  9. Device for detecting the water leak within a feedwater nozzle in water cooled reactors

    International Nuclear Information System (INIS)

    Hattori, Tsunekazu.

    1984-01-01

    Purpose: To enable exact recognition and detection for the state of water leak. Constitution: The detection device comprises a thermocouple disposed to the outer surface of a feedwater nozzle, a distortion meter for detecting the change in the outer diameter of a nozzle and an acoustic emission generator disposed to the inside of the nozzle for generating a signal upon temperature change. These sensors previously monitor the states during normal operation, and thus detect the change in each of the states upon occurrence of water leakage to issue an alarm. (Kamimura, M.)

  10. Aging and service wear of auxiliary feedwater pumps for PWR nuclear power plants

    International Nuclear Information System (INIS)

    Greenstreet, W.L.

    1989-01-01

    This paper describes investigations on auxiliary feedwater pumps being done under the Nuclear Plant Aging Research (NPAR) Program. Objectives of these studies are: to identify and evaluate practical, cost-effective methods for detecting, monitoring, and assessing the severity of time-dependent degradation (aging and service wear); recommend inspection and maintenance practices; establish acceptance criteria; and help facilitate use of the results. Emphasis is given to identifying and assessing methods for detecting failure in the incipient stage and to developing degradation trends to allow timely maintenance, repair or replacement actions. 3 refs

  11. Remote visual testing (RVT) for the diagnostic inspection of feedwater heaters

    International Nuclear Information System (INIS)

    Nugent, M.J.; Pellegrino, B.A.

    1991-01-01

    In this paper the benefits and limitations of Non-Destructive Testing (NDT) on feedwater heaters will be briefly reviewed. All Remote Visual Testing (RVT) devices including borescopes, fiberscopes, videoborescopes and Closed Circuit Television (CCTV) cameras will be discussed along with currently accepted formats for documentation. The benefits of a comprehensive in-place inspection involving Remote Visual Testing will be discussed in relationship to its diagnostic capabilities. The results of eight post-service heater inspections will be discussed along with the root cause of failure of seven unique failure mechanisms. These inspections, including FWH access, RVT tool and data analysis, will be detailed

  12. Life cycle management, design review, and condition assessment of feedwater heaters

    International Nuclear Information System (INIS)

    Gammage, D.; Idvorian, N.

    2012-01-01

    OPEX from both the Nuclear and Fossil Power Generation Industries shows that Feedwater Heaters (FWHs) are subject to several degradation mechanisms and that this degradation commonly leads to replacement of these vessels in order to ensure reliable, efficient operation of the plants. Loss of feedwater heating will impact plant thermal performance. In response to inspection results showing on-going degradation as well as other factors, B&W Canada completed a project in conjunction with a US PWR utility to review the design, condition, and Life Cycle Management of their FWHs. This project involved a multi-disciplinary approach in order to consider all aspects of the FWHs in order to provide insight into the Life Cycle Management Plan (LCMP) so that the FWHs can be operated reliably into the future and so that adequate inspections can be conducted in order to produce a detailed condition assessment. The utility was interested in evaluating their FWH LCMP to determine if it was adequate in its requirements to enable reliable, leak-free operation of their FWH equipment. As inputs to this evaluation, it was required that B&W Canada evaluate both confirmed and plausible degradation mechanisms. They also required that the thermal hydraulic and functional design be evaluated for their particular FWHs. It was important to also incorporate industry OPEX in order to provide proper trending information for tube plugging. Out of this evaluation there were several findings and recommendations that could be used to update the utilities’ LCMP as it was apparent that the current version may not be truly reflective of the current condition of the equipment or of current industry OPEX of such FWHs. Several recommendations came from this evaluation, the most significant were: • Performing thermal/hydraulic, FIV (flow-induced vibration), and tube/shell interaction calculations to determine how the FWHs operate and how their performance can change over time as a function of tube

  13. Feed-water heaters alternative design comparison; Comparacion de disenos alternativos de calentadores

    Energy Technology Data Exchange (ETDEWEB)

    Torres Toledano, Gerardo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1988-12-31

    A procedure is presented for the alternative design comparison of feed water heaters, based in the failure records of damaged tubes during operation. The procedure is used for cases in which non-continuous or random inspections are made to the feed-water heaters. [Espanol] Se presenta un procedimiento para comparar disenos alternativos de calentadores, basandose en los registros de fallas de los tubos rotos acumuladas durante su operacion. El procedimiento se emplea para casos en los que se realizan inspecciones a los calentadores no continuas, ya sea periodicas o al azar.

  14. Regulatory analysis for the resolution of Generic Issue 125.II.7 ''Reevaluate Provision to Automatically Isolate Feedwater from Steam Generator During a Line Break''

    International Nuclear Information System (INIS)

    Basdekas, D.L.

    1988-09-01

    Generic Issue 125.II.7 addresses the concern related to the automatic isolation of auxiliary feedwater (AFW) to a steam generator with a broken steam or feedwater line. This regulatory analysis provides a quantitative assessment of the costs and benefits associated with the removal of the AFW automatic isolation and concludes that no new regulatory requirements are warranted. 21 refs., 7 tabs

  15. Main trends in modernization of I and C systems at NPPs in the Russian Federation

    International Nuclear Information System (INIS)

    Pobedonostsev, A.B.; Chudin, A.G.

    1998-01-01

    There are more than 20 nuclear power reactors in operation in Russia, which have been operated over 10 years. Operational time of I and C systems of those NPPs is about 30 years, though the lifetime of individual parts of I and C systems is limited by 10-15 years. I and C systems were designed in 60-70-th in accordance with the existing regulations and available technical solutions. Obsolescence of those I and C systems require the reconstruction of existing systems. There are considerations that don't allow to perform equipment replacement one-to-one but require to modernize I and C systems at NPP. (author)

  16. The influence of the political, economic and main force risk factors on the power system

    International Nuclear Information System (INIS)

    Barbu, C.; Ruxanda, G.

    1996-01-01

    This paper deals mainly with the economic aspects of the Romanian power development policy in the context of European Union integration. There are presented the gradual steps of regional interconnection which are currently undertaken under PHARE program. Special attention is given to the legislation, power production technology, costs, and environmental protection problems. (C.M.) 1 fig

  17. Main control system verification and validation of NPP digital I and C system based on engineering simulator

    International Nuclear Information System (INIS)

    Lin Meng; Hou Dong; Liu Pengfei; Yang Zongwei; Yang Yanhua

    2010-01-01

    Full-scope digital instrumentation and controls system (I and C) technique is being introduced in Chinese new constructed Nuclear Power Plant (NPP), which mainly includes three parts: control system, reactor protection system and engineered safety feature actuation system. For example, SIEMENS TELEPERM XP and XS distributed control system (DCS) have been used in Ling Ao Phase II NPP, which is located in Guangdong province, China. This is the first NPP project in China that Chinese engineers are fully responsible for all the configuration of actual analog and logic diagram, although experience in NPP full-scope digital I and C is very limited. For the safety, it has to be made sure that configuration is right and control functions can be accomplished before the phase of real plant testing on reactor. Therefore, primary verification and validation (V and V) of I and C needs to be carried out. Except the common and basic way, i.e. checking the diagram configuration one by one according to original design, NPP engineering simulator is applied as another effective approach of V and V. For this purpose, a virtual NPP thermal-hydraulic model is established as a basis according to Ling Ao Phase II NPP design, and the NPP simulation tools can provide plant operation parameters to DCS, accept control signal from I and C and give response. During the test, one set of data acquisition equipments are used to build a connection between the engineering simulator (software) and SIEMENS DCS I/O cabinet (hardware). In this emulation, original diagram configuration in DCS and field hardware structures are kept unchanged. In this way, firstly judging whether there are some problems by observing the input and output of DCS without knowing the internal configuration. Then secondly, problems can be found and corrected by understanding and checking the exact and complex configuration in detail. At last, the correctness and functionality of the control system are verified. This method is

  18. Development of monitoring and control system for a mine main fan based on frequency converter

    Science.gov (United States)

    Zhang, Y. C.; Zhang, R. W.; Kong, X. Z.; Y Gong, J.; Chen, Q. G.

    2013-12-01

    In the process of mine exploitation, the requirement of air flow rate often changes. The procedure of traditional control mode of the fan is complex and it is hard to meet the worksite requirement for air. This system is based on Principal Computer (PC) monitoring system and high performance PLC control system. In this system, the frequency converter is adapted to adjust the fan speed and the air of worksite can be regulated steplessly. The function of the monitoring and control system contains on-line monitoring and centralized control. The system can monitor the parameters of fan in real-time, control the operation of frequency converter, as well as, control the fan and its accessory equipments. At the same time, the automation level of the system is highly, the field equipments can be monitored and controlled automatically. So, the system is an important safeguard for mine production.

  19. Development of monitoring and control system for a mine main fan based on frequency converter

    International Nuclear Information System (INIS)

    Zhang, Y C; Kong, X Z; Chen, Q G; Zhang, R W; Gong, J Y

    2013-01-01

    In the process of mine exploitation, the requirement of air flow rate often changes. The procedure of traditional control mode of the fan is complex and it is hard to meet the worksite requirement for air. This system is based on Principal Computer (PC) monitoring system and high performance PLC control system. In this system, the frequency converter is adapted to adjust the fan speed and the air of worksite can be regulated steplessly. The function of the monitoring and control system contains on-line monitoring and centralized control. The system can monitor the parameters of fan in real-time, control the operation of frequency converter, as well as, control the fan and its accessory equipments. At the same time, the automation level of the system is highly, the field equipments can be monitored and controlled automatically. So, the system is an important safeguard for mine production

  20. The modern instrumentation used for monitoring and controlling the main parameters of the regenerative electro-mechano-hydraulic drive systems

    Science.gov (United States)

    Cristescu, Corneliu; Drumea, Petrin; Krevey, Petrica

    2009-01-01

    In this work is presented the modern instrumentation used for monitoring and controlling the main parameters for one regenerative drive system, used to recovering the kinetic energy of motor vehicles, lost in the braking phase, storing and using this energy in the starting or accelerating phases. Is presented a Romanian technical solution for a regenerative driving system, based on a hybrid solution containing a hydro-mechanic module and an existing thermal motor drive, all conceived as a mechatronics system. In order to monitoring and controlling the evolution of the main parameters, the system contains a series of sensors and transducers that provide the moment, rotation, temperature, flow and pressure values. The main sensors and transducers of the regenerative drive system, their principal features and tehnical conecting solutions are presented in this paper, both with the menaging electronic and informational subsystems.

  1. Analysis of the Pactel loss-of-feedwater experiments

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V.; Virtanen, E.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Kouhia, J. [VTT Energy, Espoo (Finland). Nuclear Energy

    1995-12-31

    The western thermal-hydraulic system codes, like RELAP5 and CATHARE2, are developed and validated to model the phenomena in the vertical steam generators. The differences between vertical and horizontal steam generators are so significant that the calculational models developed for the vertical steam generators are not directly applicable for the horizontal steam generators. The PACTEL loss of feed water experiments have been performed to study the overall behaviour of the horizontal steam generators and especially to produce experimental data for code assessment. In this paper the results from APROS, CATHARE2 and RELAP5 calculations of a series of experiments are presented. 6 refs.

  2. Analysis of the Pactel loss-of-feedwater experiments

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V; Virtanen, E; Haapalehto, T [Lappeenranta Univ. of Technology (Finland); Kouhia, J [VTT Energy, Espoo (Finland). Nuclear Energy

    1996-12-31

    The western thermal-hydraulic system codes, like RELAP5 and CATHARE2, are developed and validated to model the phenomena in the vertical steam generators. The differences between vertical and horizontal steam generators are so significant that the calculational models developed for the vertical steam generators are not directly applicable for the horizontal steam generators. The PACTEL loss of feed water experiments have been performed to study the overall behaviour of the horizontal steam generators and especially to produce experimental data for code assessment. In this paper the results from APROS, CATHARE2 and RELAP5 calculations of a series of experiments are presented. 6 refs.

  3. Analysis of the Pactel loss-of-feedwater experiments

    International Nuclear Information System (INIS)

    Korteniemi, V.; Virtanen, E.; Haapalehto, T.; Kouhia, J.

    1995-01-01

    The western thermal-hydraulic system codes, like RELAP5 and CATHARE2, are developed and validated to model the phenomena in the vertical steam generators. The differences between vertical and horizontal steam generators are so significant that the calculational models developed for the vertical steam generators are not directly applicable for the horizontal steam generators. The PACTEL loss of feed water experiments have been performed to study the overall behaviour of the horizontal steam generators and especially to produce experimental data for code assessment. In this paper the results from APROS, CATHARE2 and RELAP5 calculations of a series of experiments are presented

  4. The active filter voltage ripple correction system of the Brookhaven AGS main magnet power supply

    International Nuclear Information System (INIS)

    Marneris, I.; Bonati, R.; Geller, J.; Sandberg, J.N.; Soukas, A.

    1995-01-01

    This paper, and a companion paper, describe the improvements to the Main Magnet Power Supply (MMPS) so that it enables a more flexible operation of the AGS, enhances its reliability, and also improves the MMPS's ultimate performance specifications. One of the major areas for the latter is the fixed target program operating off the AGS slow extracted beam lines. The active filter, by improving the MMPS output ripple, is instrumental in the improvement of the ultimate duty factor of the extraction beam spill

  5. The main microelements and phosphorus content of sediments formed in a drinking water supply system

    OpenAIRE

    Marina Valentukeviciene; Ramune Zurauskiene; Jonas Satkunas

    2016-01-01

    Groundwater is the only source for drinking water supply in Lithuania. Twenty water intakes exploiting Quaternary aquifers are operating in Vilnius City. The main aim of this study was to characterize the heavy metal content of internal pipeline sediments in the water supply network. It also provides a new insight into the accumulation of phosphorus and its variation in pipeline sediments in the study area. The results of this research reflect the level of heavy metals that accumulated during...

  6. Heating water. Specifications for feedwater; Heizungswasser. Eine Standortbestimmung

    Energy Technology Data Exchange (ETDEWEB)

    Hannemann, M. [Hannemann Wassertechnik (Germany)

    2004-08-01

    Water is indispensable for life and for engineering. It is so universal that we seem to have lost respect for it. In heating systems, it is recommended to know about its technical properties and to account for them by appropriate measures. (orig.) [German] Wasser ist fuer Leben und Technik unverzichtbar. Vielleicht ist es auf die enormen Vorkommen und die Selbstverstaendlichkeit der Nutzung zurueckzufuehren, dass wir ein Stueck Achtung vor diesem Gut verloren haben. Wer bei der technischen Verwendung als Waermetraeger in Heizungsanlagen die speziellen Eigenschaften und die Wechselwirkungen mit Werkstoffen missachtet, kann schnell in unruhiges Fahrwasser geraten, mithin bis zum Schiffbruch. Wer die Wechselwirkungen kennt, beachtet und mit speziellen Behandlungsverfahren gegensteuert, ist daher gut beraten. (orig.)

  7. Heterogeneity of critical systems as the main factor determining their radiostability

    International Nuclear Information System (INIS)

    Gudkov, I.N.

    1984-01-01

    It is shown that meristem can be considered as a convenient object for investigating regularities in the propagation kinetics of cells of critical systems and mechanisms of its regulation. Critical organs are rather perfect heterogeneous systems of automatic control where processes of their propagation and functional activity occur under strict control of spectral regulation systems to provide permanent composition and amount of cells. These systems influence purposeful cell alterations when they pass separate stages of the cell cycle. They provide the formation of funds of reserve cells and reliability of functioning of these highely sensitive organism tissues

  8. The MainSTREAM Component Platform: A Holistic Approach to Microfluidic System Design

    DEFF Research Database (Denmark)

    Sabourin, David; Skafte-Pedersen, Peder; Søe, Martin Jensen

    2013-01-01

    A microfluidic component library for building systems driving parallel or serial microfluidic-based assays is presented. The components are a miniaturized eight-channel peristaltic pump, an eight-channel valve, sample-to-waste liquid management, and interconnections. The library of components...... of reaction chips; (2) highly parallel pumping and routing/valving capability; (3) methods to interface pumps and chip-to-liquid management systems; (4) means to construct a portable system; (5) reconfigurability/flexibility in system design; (6) means to interface to microscopes; and (7) compatibility...

  9. Telemetry system for monitoring the ECG for patients with high cardiovascular risk. Main design requirements and technical solutions

    International Nuclear Information System (INIS)

    Rodriguez, J; Meissimilly, G; Berovides, JD

    2005-01-01

    In this paper the main design requirements concerning the setting up of a telemetry ECG monitoring system are presented. The design's most important technical solutions as well as some details are also discussed. This system is intended to provide skilled medical assistance during the cardiac rehabilitation of both asymptomatic and high risk coronary patients

  10. Operational experience on reduction of feedwater iron and liquid radwaste input for Kuosheng Nuclear Power Plant

    International Nuclear Information System (INIS)

    Wen, T.J.; Huang, Theresa Chen; Liu, Wen Tsung; Liu, T.C.; Shyur, Tzu Sheng; Shen, S.C.

    1998-01-01

    Other than cobalt alloys, or low cobalt materials, feedwater iron content plays an important role in crud activation and transport causing the growth of out-of-core radiation fields and associated with radwaste generation. Before installing prefilter in the upstream of condensate deep-bed demineralizer, increasing demand for suspended solid removal required new backwash and regeneration technique in Kuosheng Nuclear Power Plant. At steady state full power operation, the average iron concentration in condensate demineralizer influent was 8-15 ppb. Considering both the necessity of backwash and reduction of liquid radwaste input, several actions had been taken to promote the crud removal capabilities without using ultrasonic resin cleaner and controlled feedwater iron content between 0.5 and 2.0 ppb. This modified resin backwash technique would also generate minimum liquid radwaste. Meanwhile, significant efforts have been made to promote the quality of waste water by carefully control input streams as well as backwash modification to reduce liquid radwaste generation. The daily quantity of liquid radwaste has decreased dramatically in the past two years and is effectively controlled under the expected average daily input of design basis. (author)

  11. Evaluation method for two-phase flow and heat transfer in a feed-water heater

    International Nuclear Information System (INIS)

    Takamori, Kazuhide; Minato, Akihiko

    1993-01-01

    A multidimensional analysis code for two-phase flow using a two-fluid model was improved by taking into consideration the condensation heat transfer, film thickness, and film velocity, in order to develop an evaluation method for two-phase flow and heat transfer in a feed-water heater. The following results were obtained by a two-dimensional analysis of a feed-water heater for a power plant. (1) In the model, the film flowed downward in laminar flow due to gravity, with droplet entrainment and deposition. For evaluation of the film thickness, Fujii's equation was used in order to account for forced convection of steam flow. (2) Based on the former experimental data, the droplet deposition coefficient and droplet entrainment rate of liquid film were determined. When the ratio at which the liquid film directly flowed from an upper heat transfer tube to a lower heat transfer tube was 0.7, the calculated total heat transfer rate agreed with the measured value of 130 MW. (3) At the upper region of a heat transfer tube bundle where film thickness was thin, and at the outer region of a heat transfer tube bundle where steam velocity was high, the heat transfer rate was large. (author)

  12. Simulation of loss of feedwater transient of MASLWR test facility by MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Juyeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is being current developed domestically also adopts helical coil steam generator, KINS has joined this ICSP to evaluate performance of domestic regulatory audit thermal-hydraulic code (MARS-KS code) in various respects including wall-to-fluid heat transfer model modification implemented in the code by independent international experiment database. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3). In the present study, KINS simulation results by the MARS-KS code (KS-002 version) for the SP-2 experiment are presented in detail and conclusions on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the loss of feedwater transient of the MASLWR test facility. Steady state run shows helical coil specific heat transfer models implemented in the code is reasonable. However, through the transient run, it is also found that three-dimensional effect within the HPC and axial conduction effect through the HTP are not well reproduced by the code.

  13. Condensation heat transfer of a feed-water heater and improvement of its performance

    International Nuclear Information System (INIS)

    Takamori, Kazuhide; Murase, Michio; Baba, Yoshikazu; Aihara, Tsuyoshi

    1995-01-01

    In this study, a condensation heat transfer model, coupled with a three-dimensional two-phase flow analysis, was developed. In the heat transfer model, the liquid film flow rate on the heat transfer tubes was calculated by a mass balance equation and the liquid film thickness was calculated from the liquid film flow rate using Nusselt's laminar flow model and Fujii's equation for the steam velocity effect. The model was verified by condensation heat transfer experiments. In the experiments, 112 horizontal, staggered tubes with an outer diameter of 16mm and length of 0.55m were used. The calculated over-all heat transfer coefficients agreed with the data within ±5% under the inlet quality conditions of 13-100%. Based on a three-dimensional two-phase flow analysis, an improved feed-water heater with support plates, which have flow holes between the upper and lower tube bundles, was designed. The total heat exchange capacity of the improved feed-water heater increased about 6%. (author)

  14. Dynamics of Mycoplasma hyopneumoniae seroconversion and infection in pigs in the three main production systems.

    Science.gov (United States)

    Giacomini, Enrico; Ferrari, Nicola; Pitozzi, Alessandra; Remistani, Michela; Giardiello, Daniele; Maes, Dominiek; Alborali, Giovanni Loris

    2016-06-01

    In this study, we investigated the dynamics of Mycoplasma hyopneumoniae infections in 66 pig farms, with different production systems (one-, two-, and three-site systems), and considered different risk factors. Serological assay was used to detect serum antibodies against M. hyopneumoniae and real time polymerase chain reaction (RT-PCR) was performed to detect M. hyopneumoniae DNA in tracheobronchial swabs. Results demonstrated that M. hyopneumoniae infection status was predominantly influenced by the age of the animals and the type of production system. Infection rates were higher in older animals and the prevalence was higher in the one- and two-site systems than in the three-site systems. Dynamics of infection by RT-PCR showed that earlier M. hyopneumoniae infection on one-site farms occurs earlier, while on two- and three-site farms occurs later but spreads faster, suggesting that contact between animals of different age favors the transmission.

  15. A legacy endures. A Maine system emphasizes its sponsor's mission in all aspects of its work.

    Science.gov (United States)

    Stapleton, Marguerite

    2005-01-01

    The Sisters of Charity Health System, Lewiston, ME, a member of Covenant Health Systems, Lexington, MA, remains deeply committed to the mission of service begun by its foundress, St. Marguerite d'Youville. Although St. Marguerite experienced a hard life, her resilience and her commitment to the poor and disadvantaged serve as an inspiration to those who continue her legacy of compassionate care. The founding work of St. Marguerite and the sisters has helped to foster a culture in which the mission of service thrives among the system's 2,000 employees. This culture can be attributed to two things: the system's organizational values of compassion, stewardship, respect, and excellence; and the recognition of those employees whose work embodies these values. From the boardroom to the patient room, mission is integrated into each decision and action. Every two years, each of Covenant Health System's member facilities engages in a mission assessment process that examines various aspects of mission, including Catholic identity, holistic care, care for the poor, mission values integration, ethics and employee relations. In addition, the Sisters of Charity Health System's board has its own standing Mission and Community Committee, which looks strategically at how creatively and faithfully the system is continuing to live its mission.

  16. The Main Subsystems Involved in Defining the Quality Management System in a Hospital

    Directory of Open Access Journals (Sweden)

    Dobrea Valentina Alina

    2010-06-01

    Full Text Available The hospital is the most important organization in health field, so they have to improve the quality in all the activities deployed. A very suitable way to show the hospital’s preoccupation for quality of health services is the quality management system certificate according ISO 9001/2000. In understanding the architecture of the hospital quality management system is necessary to decompose this system in subsystems and analyze each separately: the managerial subsystem, the human subsystem, the social subsystem, thetechnical subsystem, the informative subsystem. The relationship between those subsystems leads to the continuous improvement of quality in health services.

  17. Mexico: An Evaluation of the Main Features of the Tax System

    OpenAIRE

    Jorge Martinez-Vazquez

    2001-01-01

    Mexico's tax system is a paradox. The tax policy and tax administration reforms of the late 1980s and early 1990s delivered a tax structure that is in many ways comparable, if not superior, to that in many OECD countries. However, Mexico's tax system continues to perform in some fundamental ways, in particular in its ability to raise adequate revenues, worse than the tax system of many developing countries. The basic objective of this evaluation is to try to explain this paradox. In doing tha...

  18. Columbia River System Operation Review : Final Environmental Impact Statement, Main Report Exhibits.

    Energy Technology Data Exchange (ETDEWEB)

    Columbia River System Operation Review (U.S.)

    1995-11-01

    This Volume is a part of the Final Environmental Impact Statement (EIS) for the Columbia River System. This volume contains technical exhibits of cultural resources and commentary on the (System Operation Review) SOR process. The Confederated Tribes of the Umatilla Indian Reservation comment is the majority of the material in the volume, in the Consultation Plan, Identification of trust resources; Criteria for the selection of a System Operating Strategy; comment on rights protection and implementation of Federal Trust responsibility; analysis of the draft EIS. Comment by other Native American Tribes and groups is also included: Confederated Tribes of the Colville Reservation; Kootenai Tribe of Idaho; Spokane Tribe of Indians; Coeur d` Alene tribe.

  19. Columbia River system operation review: Final environmental impact statement. Main report exhibits

    International Nuclear Information System (INIS)

    1995-11-01

    This Volume is a part of the Final Environmental Impact Statement (EIS) for the Columbia River System. This volume contains technical exhibits of cultural resources and commentary on the (System Operation Review) SOR process. The Confederated Tribes of the Umatilla Indian Reservation comment is the majority of the material in the volume, in the Consultation Plan, Identification of trust resources; Criteria for the selection of a System Operating Strategy; comment on rights protection and implementation of Federal Trust responsibility; analysis of the draft EIS. Comment by other Native American Tribes and groups is also included: Confederated Tribes of the Colville Reservation; Kootenai Tribe of Idaho; Spokane Tribe of Indians; Coeur d' Alene tribe

  20. Grapevine fruiting cuttings: validation of an experimental system to study grapevine physiology. I. Main vegetative characteristics

    Directory of Open Access Journals (Sweden)

    Nathalie Ollat

    1998-03-01

    Full Text Available A method for producing fruiting cuttings of grapevine cv. Cabernet Sauvignon is described. The main developmental features of these cuttings are presented. The clusters reach maturity after 5 months and the period of cluster development is not shortened. Before flowering, a careful control of stem and leaf growth improves the partitioning of stored carbon towards the roots and the cluster. Vegetative growth occurs mainly between fruit set and veraison. During ripening, the leaf to fruit ratio is between 20 and 30 cm2 of leaf area per gram of fruit. In the cuttings as well as in grapevines from the vineyard, free polyamine content in the leafblade does not change during development. Conjugate polyamine content does not evolve in the same way. It peaks at veraison for vineyard leaves but decreases continuously for cuttings. A proper nutrient solution provides a minerai composition in accordance with viticultural requirements. The cuttings also behave like vineyard plants in response to different levels of potassium in the nutrient solution.

  1. Highway Safety Information System guidebook for the Maine state data files. Volume 2 : single variable tabulations

    Science.gov (United States)

    2012-10-01

    The United States and European Union (EU) share many of the same transportation research issues, challenges, and goals. They also share a belief that cooperative vehicle (also termed connected vehicle) systems, based on vehicle-to-vehicle and vehicle...

  2. Optical design of the National Ignition Facility main laser and switchyard/target area beam transport systems

    Science.gov (United States)

    Miller, John L.; English, R. Edward, Jr.; Korniski, Ronald J.; Rodgers, J. Michael

    1999-07-01

    The optical design of the main laser and transport mirror sections of the National Ignition Facility are described. For the main laser the configuration, layout constraints, multiple beam arrangement, pinhole layout and beam paths, clear aperture budget, ray trace models, alignment constraints, lens designs, wavefront performance, and pupil aberrations are discussed. For the transport mirror system the layout, alignment controls and clear aperture budget are described.

  3. The Main Directions of Improving the Management of the Higher Education System

    Directory of Open Access Journals (Sweden)

    Terovanesov Mykhajlo R.

    2017-04-01

    Full Text Available The article is aimed at determining the mechanisms of improving the management of the system of higher education in the context of the current regulatory support, which will be the means to solve the problem of inefficient functioning, lack of financing, and low quality of higher education. The factors, causing negative impact on development of the higher education system were analyzed, the most important of which are the lack of financing and the shortcomings of the legal support of the education sector. It has been determined that a shortage of funds and lack of communication between the education sphere and employers reduces the efficiency of management of the system of higher education. The legal support of the educational activities requires improvement due to the declarative nature of individual provisions on the management of the system of higher education. Consideration of the legal factor is necessary for regulating the decentralization of management in the education sphere, improvement of financing, perfection of the system of control of the education quality. Ways for decentralizing the management of the education sphere, increasing its effectiveness, and improving the economic status of higher education institutions, have been suggested. The scientific-practical recommendations have been formulated to attract potential consumers of educational services to reforming the higher education system and increasing its effectiveness.

  4. A Study of Japanese Consumption Tax System : Mainly on Multiple Tax Rates and Input Tax Credit Methods

    OpenAIRE

    栗原, 克文

    2007-01-01

    One of the most important discussions on Japanese tax system reform includes how consumption tax (Value-added tax) system ought to be. Facing issues like depopulation, aging society and large budget deficit, consumption tax can be an effective source of revenue to secure social security. This article mainly focuses on multiple tax rates and input tax credit methods of Japanese consumption tax system. Because of regressive nature of consumption tax, tax rate reduction, exemption on foodstuffs ...

  5. Control and reversal of biofilm episode in a nuclear power plant main condenser system

    International Nuclear Information System (INIS)

    Terry, J.P.; Schultz, G.; Ewens, S.

    1996-01-01

    This paper documents a biofilm episode in the main condenser of the Callaway Generating Station. The problem was reversed by the use of oxidizing biocide and a wetting/penetrating agent, along with pH adjustments and other measures. All of the factors leading up to the problem are discussed in detail as are the measures taken to reverse the problem. The most difficult aspect of the problem was identifying the onset of biofilm buildup. The existing background noise in the back pressure and fouling factor data obscured the early stages of the problem and delayed detection and treatment. Alternative monitoring techniques are discussed and the treatment and reversal of the problem are completely documented. The rationale for differentiating between biofouling and deposition caused by inorganic precipitation are discussed

  6. EDF (Electricite de France) feedback shot-peening on feedwater plants working to 360 0 C: prediction correlation and follow-up of thermal stresses relaxation

    International Nuclear Information System (INIS)

    Gauchet, J.P.

    1995-01-01

    This study predicts life duration of shot-peening effect and finally to allow the plant operator to prepare routine stopping, considering the following four steps have been: the shot-peening parameters must been carefully chosen and implementation must be reliable and perfectly reproducible; the residual stresses and cold working state checked by X-ray diffraction; the EDF feedback on different steam-water system components working at around-300 0 C and repaired by shot-peening, like feed heater water boxes, water tanks and vessels, steam pipes; a program, carried out on a feedwater tank repaired by welding and hot-peening and working at 360 0 C, on the correlation between expected and effective results. (author). 7 refs., 3 figs., 1 tab

  7. Main tasks of studying strong regulation of excitation of complex electrical system generators

    Energy Technology Data Exchange (ETDEWEB)

    Gruzdev, I.A.; Yekimova, M.M.

    1982-01-01

    A survey is made of the current state of studies of the damping properties of complex electricity systems. The calculation programs of stability are based on frequency methods using the method of D-division. Now, when ARV of strong effect dominates at the SG, the task of coordinating their adjustments develops. Consequently, the following questions are discussed: study of the properties of quality functional with several points of regulation in the circuits of different structure; development of the efficient procedures for coordinating the ARV adjustment of the related energy systems; and creation of resources for solving these tasks. Results are presented of coordinating the ARV adjustments of the generators of the 3-machine electricity system. As an example, nonlinear relationships are shown between the obtained degree of stability and the coefficient of stabilization.

  8. Direct ultimate disposal of spent fuel DEAB. Systems analysis. Ultimate disposal concepts. Final report. Main volume

    International Nuclear Information System (INIS)

    Wahl, A.

    1995-10-01

    The results elaborated under the project, systems analysis of mixed radwaste disposal concepts and systems analysis of ultimate disposal concepts, provide a comprehensive description and assessment of a radwaste repository, for heat generating wastes and for wastes with negligible heat generation, and thus represent the knowledge basis for forthcoming planning work for a repository in an abandoned salt mine. A fact to be considered is that temperature field calculations have shown that there is room for further optimization with regard to the mine layout. The following aspects have been analysed: (1) safety of operation; (2) technical feasibility and realisation and licensability of the concepts; (3) operational aspects; (4) varieties of utilization of the salt dome for the intended purpose (boreholes for waste emplacement, emplacement in galleries, multi-horizon systems); (5) long-term structural stability of the mine; (6) economic efficiency; (7) nuclear materials safeguards. (orig./HP) [de

  9. SWR 1000: the main design features of the advanced boiling water reactor with passive safety systems

    International Nuclear Information System (INIS)

    Carsten, Pasler

    2007-01-01

    The SWR-1000 (1000 MW) is a boiling water reactor whose economic efficiency in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies have been enlarged from a 10*10 rod array to a 12*12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving the control rods, as well as allowing spectral-shift and stretch-out operation. The plant safety concept is based on a combination of passive safety systems and a reduced number of active safety systems. All postulated accidents can be controlled using passive systems alone. Control of a postulated core melt accident is assured with considerable safety margins thanks to passive flooding of the containment for in-vessel melt retention. The SWR-1000 is compliant with international nuclear codes and standards, and is also designed to withstand

  10. The history of the construction of main state irrigation systems in the postwar period

    Directory of Open Access Journals (Sweden)

    A. N. Chepurda

    2017-03-01

    It was established that the construction of large public irrigation systems in Ukraine started in 1951 stated that hydro construction that unfolded in the south of Ukraine, opened prospects to final elimination of the impact of drought and dry winds, the prospects for an unprecedented economic prosperity of the area.

  11. Children's perspectives on crime and the criminal justice system: main findings

    CSIR Research Space (South Africa)

    Badenhorst, C

    2008-01-01

    Full Text Available This research paper is an exploratory pilot study aimed at accessing the views of children on the criminal justice system, their perceptions of how children in conflict with the law are treated, the impact that crime has on them, their schools...

  12. Acoustic Modeling and Analysis for the Space Shuttle Main Propulsion System Liner Crack Investigation

    Science.gov (United States)

    Casiano, Matthew J.; Zoladz, Tom F.

    2004-01-01

    Cracks were found on bellows flow liners in the liquid hydrogen feedlines of several space shuttle orbiters in 2002. An effort to characterize the fluid environment upstream of the space shuttle main engine low-pressure fuel pump was undertaken to help identify the cause of the cracks and also provide quantitative environments and loads of the region. Part of this effort was to determine the duct acoustics several inches upstream of the low-pressure fuel pump in the region of a bellows joint. A finite element model of the complicated geometry was made using three-dimensional fluid elements. The model was used to describe acoustics in the complex geometry and played an important role in the investigation. Acoustic mode shapes and natural frequencies of the liquid hydrogen in the duct and in the cavity behind the flow liner were determined. Forced response results were generated also by applying an edgetone-like forcing to the liner slots. Studies were conducted for state conditions and also conditions assuming two-phase entrapment in the backing cavity. Highly instrumented single-engine hot fire data confirms the presence of some of the predicted acoustic modes.

  13. Spectrophotometry of zirconocene-polymethyl alumoxan catalytic systems: analysis of main components and parametric simulation

    International Nuclear Information System (INIS)

    Ryabenko, A.G.; Fajngol'd, E.E.; Ushakov, E.N.; Bravaya, N.M.

    2005-01-01

    Transformation of electronic spectra of absorption of zirconocene catalytic systems Ph 2 CCpFluZrCl 2 -poly methyl-alumoxan (MAO) and rac-Me 2 Si(2-Me,4-PhInd) 2 ZrCl 2 -MAO (Flu-fluorenyl, Ind-indenyl) in toluene as ratio of reagents Al MAO /Zr changes from 0 to 3000 mol mol -1 . Analysis of spectroscopic data with using the statistical methods made possible recognition of number of products of reaction in every system. Reaction model involving three equilibrium is suggested. Effective values of equilibrium constants and absorption spectra of separate products of reactions were evaluated by means of parametric self-simulation of experimental spectra [ru

  14. Columbia River system operation review: Final environmental impact statement. Main report

    International Nuclear Information System (INIS)

    1995-11-01

    The System Operation Review (SOR) Final EIS addresses four actions: (a) need to develop coordinated strategy for managing the multiple uses of the Federal Columbia River system (System Operating Strategy [SOS]); (b) need to provide interested parties other than management agencies with a long-term role in system planning (Forum); (c) need to renew or change current Canadian Entitlement Allocation Agreements (CEAA); and (d) need to renegotiate and renew the Pacific Northwest Coordination Agreement (PNCA). SOS alternatives analyzed are: (1) operation prior to Endangered Species Act listings of salmon stocks; (2) current operations (no action); (3) stable storage project operation; (4) natural river operation; (5) fixed drawdown; (6) operating strategies proposed by the U.S. Fish and Wildlife Service, National Marine Fisheries Service, State fisheries agencies, Native American tribes, and Federal operating agencies; and (7) Preferred Alternative. The seven Forum alternatives analyzed are: (1) decisionmaking by the SOR lead agencies (preferred alternative); (2) decisionmaking by SOR lead agencies and recommendations by an existing regional entity; (3) decisionmaking by SOR lead agencies and recommendations by a new regional entity; (4) decisionmaking by a Federal consultation forum; (5) decisionmaking by a new entity; (6) decisionmaking by one Federal operating agency; (7) decisionmaking by a Federal agency other than an operating agency. PNCA alternatives analyzed are: (1) no replacement contract; (2) contract to maximize regional power benefits; (3) roll over existing PNCA; (4) current PNCA with modified operating procedures (preferred alternative); (5) current PNCA with nonpower modifications. CEAA alternatives include: (1) no action (no replacement of current allocation agreements); (2) entitlement allocation: 55 percent Federal; 45 percent non-Federal; (3) entitlement allocation: 70 percent Federal, 30 percent non-Federal (preferred alternative); (4) no agreement

  15. ELECTROMAGNETIC SAFETY OF ELECTRIC TRANSPORT SYSTEMS: MAIN SOURCES AND PARAMETERS OF MAGNETIC FIELDS

    Directory of Open Access Journals (Sweden)

    N. G. Ptitsyna

    2013-03-01

    Full Text Available Magnetic fields produced by electric drive vehicles may break electromagnetic safety. For electromagnetic safety and electromagnetic compatibility knowledge about characteristics and sources of magnetic fields in the electric transport is necessary. The article deals with analysis of available data about magnetic fields in electric cars and comparison with results of our measurements carried out in the other types of electrified transport systems.

  16. Main lines of reorganization of the system of postgraduate training of radiologists

    International Nuclear Information System (INIS)

    Pavlov, A.S.; Kostromina, K.N.; Datsenko, V.S.

    1988-01-01

    The authors have summed up the experience accumulated over 20 years at the Chair of Clinical Radiology of the Central Order of Lenin Institute of Advanced Medical Training. Doctors majoring in this field, usually radiologists at the age over 35 when their creative initiative is on a decrease. Therfore the authors have proposed numerous measures aimed at the improvement of the system of training in radiology and an increase in the number of young specialists

  17. Columbia River System Operation Review : Final Environmental Impact Statement, Main Report.

    Energy Technology Data Exchange (ETDEWEB)

    Columbia River System Operation Review (U.S.); United States. Bonneville Power Administration; United States. Army. Corps of Engineers. North Pacific Division; United States. Bureau of Reclamation. Pacific Northwest Region.

    1995-11-01

    The System Operation Review (SOR) Final EIS addresses four actions: (a) need to develop coordinated strategy for managing the multiple uses of the Federal Columbia River system (System Operating Strategy [SOS]); (b) need to provide interested parties other than management agencies with a long-term role in system planning (Forum); (c) need to renew or change current Canadian Entitlement Allocation Agreements (CEAA); and (d) need to renegotiate and renew the Pacific Northwest Coordination Agreement (PNCA). SOS alternatives analyzed are: (1) operation prior to Endangered Species Act listings of salmon stocks; (2) current operations (no action); (3) stable storage project operation; (4) natural river operation; (5) fixed drawdown; (6) operating strategies proposed by the U.S. Fish and Wildlife Service, National Marine Fisheries Service, State fisheries agencies, Native American tribes, and Federal operating agencies; and (7) Preferred Alternative. The seven Forum alternatives analyzed are: (1) decisionmaking by the SOR lead agencies (preferred alternative); (2) decisionmaking by SOR lead agencies and recommendations by an existing regional entity; (3) decisionmaking by SOR lead agencies and recommendations by a new regional entity; (4) decisionmaking by a Federal consultation forum; (5) decisionmaking by a new entity; (6) decisionmaking by one Federal operating agency; (7) decisionmaking by a Federal agency other than an operating agency. PNCA alternatives analyzed are: (1) no replacement contract; (2) contract to maximize regional power benefits; (3) roll over existing PNCA; (4) current PNCA with modified operating procedures (preferred alternative); (5) current PNCA with nonpower modifications. CEAA alternatives include: (1) no action (no replacement of current allocation agreements); (2) entitlement allocation: 55 percent Federal; 45 percent non-Federal; (3) entitlement allocation: 70 percent Federal, 30 percent non-Federal (preferred alternative); (4) no agreement.

  18. Detailed Pictures of Multiple Asteroid Systems in the Main-Belt

    Science.gov (United States)

    Marchis, F.; Emery, J. P.; Enriquez, J. E.; Descamps, P.; Berthier, J.; Vachier, F.; Durech, J.

    2011-12-01

    Since their discovery less than 10 years ago, ~200 known multiple asteroid systems have been studied with a combination of observing techniques, including adaptive optics, lightcurve photometry, and mid-infrared spectrophotometry. Those observations show that ~15 large (D>100km) asteroids that are known to possess km-sized satellite(s) (22 Kalliope, 45 Eugenia, 87 Sylvia, 93 Minerva, 216 Kleopatra, ...) share common orbital characteristics, implying a common formation scenario: e.g. catastrophic disruption or ejection after an oblique impact. More than 70 smaller (10-15km) binary asteroid systems have been detected through anomalies in their lightcurves and are believed to have formed by fission due to the YORP effect. By comparison with meteorite analog densities, mid-IR data reveal that these systems have a significant porosity (larger than 30%) implying a rubble-pile interior. We will review these key results and discuss their implications for the interior of asteroids in the light of recent space mission results. Future explorations using new ground-based facilities and space mission concepts will be also discussed. This work is supported by the NSF grant AAG-0807468 and NASA grant NNX11AD62G

  19. Geothermal Heat Pump System for New Student Housing Project at the University at Albany Main Campus

    Energy Technology Data Exchange (ETDEWEB)

    Lnu, Indumathi [Univ. of Albany, NY (United States)

    2015-08-27

    University at Albany successfully designed, constructed and is operating a new student housing building that utilizes ground source heat pump (GSHP) for heating and cooling the entire 191,500SF building. The installed system consists of a well field with 150 bores, 450 feet deep and (189) terminal heat pump units for a total capacity of 358 Tons cooling and 4,300 MBtu/h heating. The building opened in Fall 2012. The annual energy use and cost intensity of the building, after the changes made during the first 2 years’ of operation is 57kBtu/SF/Year and $1.30/SF/Year respectively. This is approximately 50% lower than the other residential quads on campus, despite the fact that the quads are not air-conditioned. The total project cost from design through 3-years of operations is approximately $6 Million, out of which $5.7 Million is for construction of the GSHP system including the well field. The University received a $2.78 Million grant from the Department of Energy. The estimated utility cost savings, compared to a baseline building with conventional HVAC system, is approximately $185,000. The estimated simple payback, after grant incentives, is 15 years. Additionally, the project has created 8.5FTE equivalent jobs.

  20. Conceptual design main progress of EAST Articulated Maintenance Arm (EAMA) system

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Shanshuang, E-mail: shiss@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Song, Yuntao; Cheng, Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Villedieu, Eric; Bruno, Vincent [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Feng, Hansheng [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wu, Huapeng [Laboratory of Intelligent Machines, Lappeenranta University of Technology (Finland); Wang, Peng [State Key Laboratory of Solid Lubrication, Lanzhou Institute of Chemical Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Hao, Zhiwei; Li, Yang; Wang, Kun; Pan, Hongtao [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-03-15

    Highlights: • EAST Articulated Maintenance Arm (EAMA) system is being collaboratively developed by ASIPP and CEA-IRFM. • Conceptual design for a 3-DOF wrist end effector with gripper has been finished. • Kinematic design can reach 90% of the workspace inside EAST tokamak vessel. • A prototype of EAMA arm segment has been built to validate the design. - Abstract: EAST articulated maintenance arm (EAMA) system is being collaboratively developed by ASIPP and CEA-IRFM for the purpose of remote inspection and simple maintenance operations in EAST vacuum vessel during physical experiments without breaking the ultra-high vacuum condition. The EAMA system design is based on a similar articulated inspection arm robot successfully demonstrated in Tore Supra in 2008. In order to better meet EAST configurations and maintenance requirements, optimized mechanisms and dimensions are considered for EAMA robot as upgrades. Besides, the segmented arm is equipped with a 3-DOF wrist end effector and gripper for gripping operation as well as inspection. Some calculations and simulations on statics, kinematics and workspace of EAMA have been presented to validate the feasibility. This paper introduces the overall design of the EAMA robot and presents implementation progress.

  1. Conceptual design main progress of EAST Articulated Maintenance Arm (EAMA) system

    International Nuclear Information System (INIS)

    Shi, Shanshuang; Song, Yuntao; Cheng, Yong; Villedieu, Eric; Bruno, Vincent; Feng, Hansheng; Wu, Huapeng; Wang, Peng; Hao, Zhiwei; Li, Yang; Wang, Kun; Pan, Hongtao

    2016-01-01

    Highlights: • EAST Articulated Maintenance Arm (EAMA) system is being collaboratively developed by ASIPP and CEA-IRFM. • Conceptual design for a 3-DOF wrist end effector with gripper has been finished. • Kinematic design can reach 90% of the workspace inside EAST tokamak vessel. • A prototype of EAMA arm segment has been built to validate the design. - Abstract: EAST articulated maintenance arm (EAMA) system is being collaboratively developed by ASIPP and CEA-IRFM for the purpose of remote inspection and simple maintenance operations in EAST vacuum vessel during physical experiments without breaking the ultra-high vacuum condition. The EAMA system design is based on a similar articulated inspection arm robot successfully demonstrated in Tore Supra in 2008. In order to better meet EAST configurations and maintenance requirements, optimized mechanisms and dimensions are considered for EAMA robot as upgrades. Besides, the segmented arm is equipped with a 3-DOF wrist end effector and gripper for gripping operation as well as inspection. Some calculations and simulations on statics, kinematics and workspace of EAMA have been presented to validate the feasibility. This paper introduces the overall design of the EAMA robot and presents implementation progress.

  2. A high fidelity model and code generator for the simulation of BOP systems

    International Nuclear Information System (INIS)

    Galen, S.; Vinay, M.

    1993-01-01

    TOPMERET represents a significant advance in the modelling fidelity of Balance of Plant systems (BOP). It is extremely flexible and can accommodate a variety of systems, including main steam, feedwater, turbine, condenser, offgas, large volumes, such as the containment, and water systems such as service water. It handles both normal and abnormal operating scenarios, including pipe break accidents. It was tested successfully on various simulators, and meets the fidelity required of BOP system models so as to successfully integrate with the high level of control automation of European designs. (Z.S.) 1 ref

  3. Extensive feedwater quality control and monitoring concept for preventing chemistry-related failures of boiler tubes in a subcritical thermal power plant

    International Nuclear Information System (INIS)

    Vidojkovic, Sonja; Onjia, Antonije; Matovic, Branko; Grahovac, Nebojsa; Maksimovic, Vesna; Nastasovic, Aleksandra

    2013-01-01

    Prevention and minimizing corrosion processes on steam generating equipment is highly important in the thermal power industry. The maintenance of feedwater quality at a level corresponding to the standards of technological designing, followed by timely respond to the fluctuation of measured parameters, has a decisive role in corrosion prevention. In this study, the comprehensive chemical control of feedwater quality in 210 MW Thermal Power Plant (TPP) was carried out in order to evaluate its potentiality to assure reliable function of the boiler and discover possible irregularity that might be responsible for frequent boiler tube failures. Sensitive on-line and off-line analytical instruments were used for measuring key and diagnostic parameters considered to be crucial for boiler safety and performances. Obtained results provided evidences for exceeded levels of oxygen, silica, sodium, chloride, sulfate, copper, and conductivity what distinctly demonstrated necessity of feedwater control improvement. Consequently, more effective feedwater quality monitoring concept was recommended. In this paper, the explanation of presumable root causes of corrosive contaminants was given including basic directions for their maintenance in proscribed limits. -- Highlights: • Feedwater quality monitoring practice in a thermal power plant has been evaluated. • The more efficient feedwater quality control have been applied. • Analysis of feedwater quality parameters has been performed. • Exceeded levels of corrosive contaminants were found. • Recommendations for their maintenance at proscribed values were given

  4. Main Consolidations and Improvements of the Control System and Instrumentation for the LHC Cryogenics

    CERN Document Server

    Fluder, C; Bremer, J; Bremer, K; Ivens, B; Casas-Cubillos, J; Claudet, S; Gomes, P; Ivens, B; Perin, A; Pezzetti, M; Tovar-Gonzalez, A; Vauthier, N

    2013-01-01

    Operation of the LHC during 2010 and 2011 with 3.5 TeV beam energy and luminosity up to 3.65x1033 cm-2 s-1, led to radiation-induced failures of micro-electronic devices used in the cryogenic control system. Mitigating actions addressed equipment relocation and corrective patches on electronics and software. Driven by the technical requirements and by feedback from the cryogenic operation team, numerous consolidations and improvements were implemented on-the-fly, enhancing availability and operability of the LHC cryogenics. Furthermore, additional diagnostic tools, test benches, technical procedures and trainings have been provided to strengthen first line support services.

  5. Development of water chemistry diagnosis system for BWR primary loop

    International Nuclear Information System (INIS)

    Nagase, Makoto; Asakura, Yamato; Sakagami, Masaharu; Uchida, Shunsuke; Ohsumi, Katsumi.

    1988-01-01

    The prototype of a water chemistry diagnosis system for BWR primary loop has been developed. Its purposes are improvement of water chemistry control and reduction of the work burden on plant chemistry personnel. It has three main features as follows. (1) Intensifying the observation of water chemistry conditions by variable sampling intervals based on the on-line measured data. (2) Early detection of water chemistry data trends using a second order regression curve which is calculated from the measured data, and then searching the cause of anomaly if anything (3) Diagnosis of Fe concentration in feedwater using model simulations, in order to lower the radiation level in the primary system. (author)

  6. Main findings

    International Nuclear Information System (INIS)

    2014-01-01

    Licensing regimes vary from country to country. When the license regime involves several regulators and several licenses, this may lead to complex situations. Identifying a leading organisation in charge of overall coordination including preparation of the licensing decision is a useful practice. Also, if a stepwise licensing process is implemented, it is important to fix in legislation decisions and/or time points and to identify the relevant actors. There is considerable experience in civil and mining engineering that can be applied when constructing a deep geological disposal facility. Specific challenges are, however, the minimization of disturbances to the host rock and the understanding of its long-term behavior. Construction activities may affect the geo-hydraulic and geochemical properties of the various system components which are important safety features of the repository system. Clearly defined technical specifications and an effective quality management plan are important in ensuring successful repository implementation which is consistent with safety requirements. Monitoring plan should also be defined in advance. The regulatory organization should prepare itself to the licensing review before construction by allocating sufficient resources. It should increase its competence, e.g., by interacting early with the implementer and through its own R and D. This will allow the regulator to define appropriate technical conditions associated to the construction license and to elaborate a relevant inspection plan of the construction work. After construction, obtaining the operational license is the most important and crucial step. Main challenges include (a) establishing sufficient confidence so that the methods for closing the individual disposal units comply with the safety objectives and (b) addressing the issue of ageing of materials during a 50-100 years operational period. This latter challenge is amplified when reversibility/retrievability is required

  7. Design and performance of main vacuum pumping system of SST-1 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Khan, Ziauddin, E-mail: ziauddin@ipr.res.in; Pathan, Firozkhan; George, Siju; Dhanani, Kalpesh; Paravastu, Yuvakiran; Semwal, Pratibha; Pradhan, Subrata

    2014-01-15

    Highlights: •SST-1 Tokamak was successfully commissioned. •Vacuum vessel and cryostat were pumped down to 6.3 × 10{sup −7} mbar and 1.3 × 10{sup −5} mbar. •Leaks developed during baking were detected in-situ by RGA and confirmed later on. •Cryo-pumping effect was observed when LN2 thermal shields reached below 273 K. •Non-standard aluminum wire-seals have shown leak tightness < 1.0 × 10{sup −9} mbar l/s. -- Abstract: Steady-state Superconducting Tokamak (SST-1) was installed and it is commissioning for overall vacuum integrity, magnet systems functionality in terms of successful cool down to 4.5 K and charging up to 10 kA current was started from August 2012. Plasma operation of 100 kA current for more than 100 ms was also envisaged. It is comprised of vacuum vessel (VV) and cryostat (CST). Vacuum vessel, an ultra-high (UHV) vacuum chamber with net volume of 23 m{sup 3} was maintained at the base pressure of 6.3 × 10{sup −7} mbar for plasma confinement. Cryostat, a high-vacuum (HV) chamber with empty volume 39 m{sup 3} housing superconducting magnet system, bubble thermal shields and hydraulics for these circuits, maintained at 1.3 × 10{sup −5} mbar in order to provide suitable environment for these components. In order to achieve these ultimate vacuums, two numbers of turbo-molecular pumps (TMP) are installed in vacuum vessel while three numbers of turbo-molecular pumps are installed in cryostat. Initial pumping of both the chambers was carried out by using suitable Roots pumps. PXI based real time controlled system is used for remote operation of the complete pumping operation. In order to achieve UHV inside the vacuum vessel, it was baked at 150 °C for longer duration. Aluminum wire-seals were used for all non-circular demountable ports and a leak tightness < 1.0 × 10{sup −9} mbar l/s were achieved.

  8. Main Memory DBMS

    NARCIS (Netherlands)

    P.A. Boncz (Peter); L. Liu (Lei); M. Tamer Özsu

    2008-01-01

    htmlabstractA main memory database system is a DBMS that primarily relies on main memory for computer data storage. In contrast, normal database management systems employ hard disk based persisntent storage.

  9. Study (Prediction of Main Pipes Break Rates in Water Distribution Systems Using Intelligent and Regression Methods

    Directory of Open Access Journals (Sweden)

    Massoud Tabesh

    2011-07-01

    Full Text Available Optimum operation of water distribution networks is one of the priorities of sustainable development of water resources, considering the issues of increasing efficiency and decreasing the water losses. One of the key subjects in optimum operational management of water distribution systems is preparing rehabilitation and replacement schemes, prediction of pipes break rate and evaluation of their reliability. Several approaches have been presented in recent years regarding prediction of pipe failure rates which each one requires especial data sets. Deterministic models based on age and deterministic multi variables and stochastic group modeling are examples of the solutions which relate pipe break rates to parameters like age, material and diameters. In this paper besides the mentioned parameters, more factors such as pipe depth and hydraulic pressures are considered as well. Then using multi variable regression method, intelligent approaches (Artificial neural network and neuro fuzzy models and Evolutionary polynomial Regression method (EPR pipe burst rate are predicted. To evaluate the results of different approaches, a case study is carried out in a part ofMashhadwater distribution network. The results show the capability and advantages of ANN and EPR methods to predict pipe break rates, in comparison with neuro fuzzy and multi-variable regression methods.

  10. Considerations on the question of applying ion exchange or reverse osmosis methods in boiler feedwater processing

    International Nuclear Information System (INIS)

    Marquardt, K.; Dengler, H.

    1976-01-01

    This consideration is to show that the method of reverse osmosis presents in many cases an interesting and economical alternative to part and total desolination plants using ion exchangers. The essential advantages of the reverse osmosis are a higher degree of automization, no additional salting of the removed waste water, small constructional volume of the plant as well as favourable operational costs with increasing salt content of the crude water to be processed. As there is a relatively high salt breakthrough compared to the ion exchange method, the future tendency in boiler feedwater processing will be more towards a combination of methods of reverse osmosis and post-purification through continuous ion exchange methods. (orig./LH) [de

  11. Evaluation of total loss of feedwater accident/recovery phase and investigation of the associated EOP

    International Nuclear Information System (INIS)

    Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung

    1993-01-01

    To evaluate the sequence of event and the thermohydraulic behavior during total loss of feedwater accident and recovery procedure, a RELAP5/MOD3 calculation is performed and compared with the LOFT L9-1/L3-3 experiment. Also, the predictability of the code for the major thermohydraulic phenomena following the accident is assessed. As a result, it is found that a pressure control using the spray until the time the water level reaches the top of the pressurizer, an overpressure protection by pressurizer PORV, a recovery of the secondary heat removal capability by refilling steam generator, and an effective cooldown by the continued natural circulation can be perfomed without core uncovery. It is also found that the plantspecific evaluation is necessary to confirm the effectiveness of the current symptom-oriented emergency operating procedure, especially in an overpressure protection performance and steam generator recovery performance. (Author)

  12. Evaluation of cracking in steam generator feedwater piping in pressurized water reactor plants

    International Nuclear Information System (INIS)

    Goldberg, A.; Streit, R.D.

    1981-05-01

    Cracking in feedwater piping was detected near the inlet to steam generators in 15 pressurized water reactor plants. Sections with cracks from nine plants are examined with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Using transmission electron microscopy, fatigue striations are observed on replicas of cleaned crack surfaces. Calculations based on the observed striation spacings gave a cyclic stress value of 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses and it is concluded that the overriding factor in the cracking problem was the presence of such undocumented cyclic loads

  13. How engineering data management and system support the main process[-oriented] functions of a large-scale project

    CERN Document Server

    Hameri, A P

    1999-01-01

    By dividing the development process into successive functional operations, this paper studies the benefits of establishing configuration management procedures and of using an engineering data management systems (EDMS) in order to execute the tasks. The underlying environment is that of CERN and the ongoing, a decade long, Large Hadron Collider (LHC)-project. By identifying the main functional groups who will use the EDMS the paper outlines the basic motivations and services provided by such a system to each process function. The implications of strict configuration management on the daily operation of each functional user group are also discussed. The main argument of the paper is that each and every user of the EDMS must act in compliance with the configuration management procedures to guarantee the overall benefits from the system. The pilot EDMS being developed at CERN, which serves as a test-bed to discover the real functional needs of the organisation of an EDMS supports the conclusions. The preliminary ...

  14. Diagnosis of the Main Busbar II Panel Components Ageing of RSG-GAS Electrical System by Using Infrared Thermography

    International Nuclear Information System (INIS)

    Teguh Sulistyo; Kiswanto; Roziq Himawan; Ari Satmoko

    2007-01-01

    To support the operation of RSG-GAS safely, the diagnosis of the ageing of main busbar II BHD/BHE/BHF panel components of RSG-GAS electrical system have been done. By using infrared thermography type Thermo Tracer TH9100PM VI/PW VI. The results of the diagnosis showed that some of the components under degradation with various rate. It can cause the system failure. By understanding the components ageing degradation mechanism and performing the preventive and predictive maintenance and safety of RSG-GAS electrical system earlier, the possibility of accident can be avoided. (author)

  15. Main Memory

    NARCIS (Netherlands)

    P.A. Boncz (Peter); L. Liu (Lei); M. Tamer Özsu

    2008-01-01

    htmlabstractPrimary storage, presently known as main memory, is the largest memory directly accessible to the CPU in the prevalent Von Neumann model and stores both data and instructions (program code). The CPU continuously reads instructions stored there and executes them. It is also called Random

  16. Techniques for remote maintenance of in-cell material-handling system in the HFEF/N main cell

    International Nuclear Information System (INIS)

    Tobias, D.A.; Frickey, C.A.

    1975-01-01

    Operations in the main cell of HFEF/N have required development of remote handling equipment and unique techniques for maintaining the in-cell material-handling system. Specially designed equipment is used to remove a disabled crane or electromechanical manipulator bridge from its support rails and place it on floor stands for repair or maintenance. Support areas for the main cell, such as the spray chamber and hot repair area, provide essential decontamination, repair, and staging areas for the in-cell material-handling-system equipment and tools. A combined engineering and technical effort in upgrading existing master-slave manipulators has definitely reduced the requirements for their maintenance. The cell is primarily for postirradiation examination of LMFBR materials and fuel elements

  17. Measurements of impedances for determinating the minimum short-circuit current in main systems 500 V of underground mining establishments

    Energy Technology Data Exchange (ETDEWEB)

    Rittinghaus, D

    1981-09-01

    The complex short-circuit impedances of energized low-voltage main systems were measured with a double-bridge in underground mining operation. The magnitude of the short-circuit currents depends on these impedances. Customary calculations of such currents depend on empirical approximations. To verify the accuracy of these approximations, the measured impedances of 61 nodes in three different main systems were compared with the results of the calculations. The comparison made between the short-circuit currents determined by measurable quantities and the values calculated according to VDE 0118 shows that the stipulated coefficients for calculating the minimum short-circuit currents lie very far on the safe side. An amendment for calculating the short-circuit in accordance with VDE 0118 is therefore suggested.

  18. Trends in Nonfatal Agricultural Injury in Maine and New Hampshire: Results From a Low-Cost Passive Surveillance System.

    Science.gov (United States)

    Scott, Erika; Bell, Erin; Hirabayashi, Liane; Krupa, Nicole; Jenkins, Paul

    2017-01-01

    Agriculture is a dangerous industry, and although data on fatal injuries exist, less is known about nonfatal injuries. The purpose of this study is to describe trends in agricultural morbidity in Maine and New Hampshire from 2008 to 2010 using a newly established passive surveillance system. This passive system is supplied by injury cases gathered from prehospital care reports and hospital data. Demographics and specifics of the event were recorded for each incident case. The average age of injured people in Maine and New Hampshire was 41.7. Women constituted 43.8% of all agricultural injuries. Machinery- (n = 303) and animal- (n = 523) related injuries accounted for most agricultural incidents. Of all injured women, over 60% sustained injuries due to animal-related causes. Agricultural injuries were spread across the two states, with clustering in southern New Hampshire and south central Maine, with additional injuries in the Aroostook County area, which is located in the northeast part of the state. Seasonal variation in agricultural injuries was evident with peaks in the summer months. There was some overlap between the agricultural and logging industry for tree-related work. Our methods are able to capture traumatic injury in agriculture in sufficient detail to prioritize interventions and to evaluate outcomes. The system is low-cost and has the potential to be sustained over a long period. Differences in rates of animal- and machinery-related injuries suggest the need for state-specific safety prioritization.

  19. Main Memory

    OpenAIRE

    Boncz, Peter; Liu, Lei; Özsu, M.

    2008-01-01

    htmlabstractPrimary storage, presently known as main memory, is the largest memory directly accessible to the CPU in the prevalent Von Neumann model and stores both data and instructions (program code). The CPU continuously reads instructions stored there and executes them. It is also called Random Access Memory (RAM), to indicate that load/store instructions can access data at any location at the same cost, is usually implemented using DRAM chips, which are connected to the CPU and other per...

  20. The main but not the accessory olfactory system is involved in the processing of socially relevant chemosignals in ungulates.

    Directory of Open Access Journals (Sweden)

    Matthieu eKELLER

    2012-09-01

    Full Text Available Ungulates like sheep and goats have, like many other mammalian species, two complementary olfactory systems. The relative role played by these two systems has long been of interest regarding the sensory control of social behavior. The study of ungulate social behavior could represent a complimentary alternative to rodent studies because they live in a more natural environment and their social behaviors depend heavily on olfaction. In addition, the relative size of the main olfactory bulb (in comparison to the accessory olfactory bulb is more developped than in many other lissencephalic species like rodents. In this review, we present data showing a clear involvement of the main olfactory system in two well-characterized social situations under olfactory control in ungulates, namely maternal behavior and offspring recognition at birth and the reactivation of the gonadotropic axis of females exposed to males during the anestrous season. In conclusion, we discuss the apparent discrepancy between the absence of evidence for a role of the vomeronasal system in ungulate social behavior and the existence of a developed accessory olfactory system in these species.

  1. IWR-MAIN Water Use Forecasting System. Version 5.1. User’s Manual and System Description

    Science.gov (United States)

    1987-12-01

    fflr Fr opv U.S. Army Corps of Engineers Water Resources Support Center JUNE 1988 Institute for Water Resources IWR REPORT 88-R-6 IWR-MAIN WATER USE... Indoor Use) 1-40 Screen 16.3 - Reduction Factors (Outdoor Use) 141 Screen 16.4 - Reduction Factors (Maximum-Day Use) 1-42 Screen 17.1 - Determination...in the Library of Conservation Coefficients: Indoor Dimension 11-74 1-20 Interaction Factor Values in the Library of Conservation Coefficients

  2. [Disinfectants and main sanitary and preventive measures for protection of ventilation and air-conditioning systems from Legionella contamination].

    Science.gov (United States)

    Gerasimov, V N; Golov, E A; Khramov, M V; Diatlov, I A

    2008-01-01

    The study was devoted to selection and assessment of disinfecting preparations for prevention of contamination by Legionella. Using system of criteria for quality assessment of disinfectants, seven newdomestic ones belonging to quaternary ammonium compounds class or to oxygen-containing preparations and designed for disinfecting of air-conditioning and ventilation systems were selected. Antibacterial and disinfecting activities of working solutions of disinfectants were tested in laboratory on the test-surfaces and test-objects of premises' air-conditioning and ventilation systems contaminated with Legionella. High antimicrobial and disinfecting activity of new preparations "Dezactiv-M", "ExtraDez", "Emital-Garant", "Aquasept Plus", "Samarovka", "Freesept", and "Ecobreeze Oxy" during their exposure on objects and materials contaminated with Legionella was shown. Main sanitary and preventive measures for defending of air-conditioning and ventilation systems from contamination by Legionella species were presented.

  3. MAIN ABSTRACTS

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    Reflection on Some Issues Regarding the System of Socialism with Chinese Characteristics Zhang Xingmao The establishment of the system of socialism with Chinese characteristics, as the symbol of China's entry into the socialist society with Chinese characteristics, is the significant development of Marist theory of social formation. The Chinese model is framed and defined by the socialist system with Chinese characteristics, therefore the study of different levels and aspects of the Chinese model should be related to the relevant Chinese system to guarantee a scientific interpretation. Under the fundamental system of socialism, the historical and logical starting point of the formation of socialism with Chinese characteristics lies in eliminating the private ownership first and then allowing the existence and rapid development of the non-public sectors of the economy. With the gradual establishment and on the basis of the basic economic system in the preliminary stage of Socialism, and with the adaptive adjustments in the economic, political, cultural, and social systems, the socialist system with Chinese characteristics is gradually formed.

  4. NVL-C: Static Analysis Techniques for Efficient, Correct Programming of Non-Volatile Main Memory Systems

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seyong [ORNL; Vetter, Jeffrey S [ORNL

    2016-01-01

    Computer architecture experts expect that non-volatile memory (NVM) hierarchies will play a more significant role in future systems including mobile, enterprise, and HPC architectures. With this expectation in mind, we present NVL-C: a novel programming system that facilitates the efficient and correct programming of NVM main memory systems. The NVL-C programming abstraction extends C with a small set of intuitive language features that target NVM main memory, and can be combined directly with traditional C memory model features for DRAM. We have designed these new features to enable compiler analyses and run-time checks that can improve performance and guard against a number of subtle programming errors, which, when left uncorrected, can corrupt NVM-stored data. Moreover, to enable recovery of data across application or system failures, these NVL-C features include a flexible directive for specifying NVM transactions. So that our implementation might be extended to other compiler front ends and languages, the majority of our compiler analyses are implemented in an extended version of LLVM's intermediate representation (LLVM IR). We evaluate NVL-C on a number of applications to show its flexibility, performance, and correctness.

  5. Simulation of the fault transitory of the feedwater controller in a Boiling water reactor with the Ramona-3B code

    International Nuclear Information System (INIS)

    Hernandez M, J.L.; Ortiz V, J.

    2005-01-01

    The obtained results when carrying out the simulation of the fault transitory of the feedwater controller (FCAA) with the Ramona-3B code, happened in the Unit 2 of the Laguna Verde power plant (CNLV), in September of the year 2000 are presented. The transitory originates as consequence of the controller's fault of speed of a turbo pump of feedwater. The work includes a short description of the event, the suppositions considered for the simulation and the obtained results. Also, a discussion of the impact of the transitory event is presented on aspects of reactor safety. Although the carried out simulation is limited by the capacities of the code and for the lack of available information, it was found that even in a conservative situation, the power was incremented only in 12% above the nominal value, while that the thermal limit determined by the minimum reason of the critical power, MCPR, always stayed above the limit values of operation and safety. (Author)

  6. Investigation into sensitivity of Darlington boiler 2 feedwater flow calibration factor to boiler level control valve configuration

    Energy Technology Data Exchange (ETDEWEB)

    Coppens, D. [Darlington Nuclear Generating Station, Ontario Power Generation, Bowmanville, Ontario (Canada); Gurevich, Y. [Daystar Technologies Inc., Toronto, Ontario (Canada); Ton, V. [Inspection and Maintenance Services Div., Ontario Power Generation, Ajax, Ontario (Canada); Zobin, D. [AMEC NSS Ltd., Toronto, Ontario (Canada)

    2009-07-01

    The Ultrasonic Cross-Correlation Flow Meter (USCCFM) has been used for regular feedwater flow calibration at Darlington NGS since the early nineties. Typical measurement repeatability over the duration of a calibration run (normally several weeks long) is within {+-}0.2%. However, it was recently noticed that BO2 calibration factor experienced sudden changes of close to 1%. The paper will describe several different approaches used for identifying the reason for the observed effect. The investigation has revealed that changes in USCCFM readings are due to the complicated geometry of BO2 feedwater piping and that its accuracy can be as high as a fraction of percent if several readings are averaged around the pipe. (author)

  7. Optimation of Operation System Integration between Main and Feeder Public Transport (Case Study: Trans Jakarta-Kopaja Bus Services)

    Science.gov (United States)

    Miharja, M.; Priadi, Y. N.

    2018-05-01

    Promoting a better public transport is a key strategy to cope with urban transport problems which are mostly caused by a huge private vehicle usage. A better public transport service quality not only focuses on one type of public transport mode, but also concerns on inter modes service integration. Fragmented inter mode public transport service leads to a longer trip chain as well as average travel time which would result in its failure to compete with a private vehicle. This paper examines the optimation process of operation system integration between Trans Jakarta Bus as the main public transport mode and Kopaja Bus as feeder public transport service in Jakarta. Using scoring-interview method combined with standard parameters in operation system integration, this paper identifies the key factors that determine the success of the two public transport operation system integrations. The study found that some key integration parameters, such as the cancellation of “system setoran”, passenger get in-get out at official stop points, and systematic payment, positively contribute to a better service integration. However, some parameters such as fine system, time and changing point reliability, and information system reliability are among those which need improvement. These findings are very useful for the authority to set the right strategy to improve operation system integration between Trans Jakarta and Kopaja Bus services.

  8. Improvement on Main/backup Controller Switching Device of the Nozzle Throat Area Control System for a Turbofan Aero Engine

    Science.gov (United States)

    Li, Jie; Duan, Minghu; Yan, Maode; Li, Gang; Li, Xiaohui

    2014-06-01

    A full authority digital electronic controller (FADEC) equipped with a full authority hydro-mechanical backup controller (FAHMBC) is adopted as the nozzle throat area control system (NTACS) of a turbofan aero engine. In order to ensure the switching reliability of the main/backup controller, the nozzle throat area control switching valve was improved from three-way convex desktop slide valve to six-way convex desktop slide valve. Simulation results show that, if malfunctions of FAEDC occur and abnormal signals are outputted from FADEC, NTACS will be seriously influenced by the main/backup controller switching in several working states, while NTACS will not be influenced by using the improved nozzle throat area control switching valve, thus the controller switching process will become safer and smoother and the working reliability of this turbofan aero engine is improved by the controller switching device improvement.

  9. Review of the Main Security Problems with Multi-Agent Systems used in E-commerce Applications

    Directory of Open Access Journals (Sweden)

    Alfonso GONZÁLEZ BRIONES

    2016-12-01

    Full Text Available The ability to connect to the Internet from a wide variety of devices such as smart phones, IoT devices and desktops at anytime and anywhere, produces a large number of e-commerce transactions, such as purchases of clothes, ticket entrances for performances, or banking operations. The increasing number of these transactions has also created an increase in the number of threats and attacks by third parties to access user data banks. It is important to control the access procedure to user data so that the number of threats does not continue to grow. To do so, it is necessary to prevent unauthorized access, theft and fraud in electronic commerce, which is required to ensure the safety of these transactions. Many e-commerce platforms are developed through multi-agent-systems because they include certain advantages to control the product, resource management, task distribution, etc. However, there are a number of threats that can jeopardize the safety of the agents that make up the system. These issues must be taken into account in the development of these multi-agent systems. However, existing methods of development do not cover in depth the issue of security. It is necessary to present and classify the potential security flaws of multi-agent systems. Therefore, the present research presents a review of the main vulnerabilities that occur in multi-agent systems responsible for managing e-commerce applications, as well as the proposed solutions to the major security problems on these platform systems. The main conclusions provided by this research is the need to optimize security measures and enhance the different security solutions applied in e-commerce applications in order to prevent identity theft, access to private data, access control, etc. It is therefore essential to continue to develop the security methods employed in applications such as e-commerce as different types of attacks and threats continue to evolve.

  10. Specialized data analysis for the Space Shuttle Main Engine and diagnostic evaluation of advanced propulsion system components

    Science.gov (United States)

    1993-01-01

    The Marshall Space Flight Center is responsible for the development and management of advanced launch vehicle propulsion systems, including the Space Shuttle Main Engine (SSME), which is presently operational, and the Space Transportation Main Engine (STME) under development. The SSME's provide high performance within stringent constraints on size, weight, and reliability. Based on operational experience, continuous design improvement is in progress to enhance system durability and reliability. Specialized data analysis and interpretation is required in support of SSME and advanced propulsion system diagnostic evaluations. Comprehensive evaluation of the dynamic measurements obtained from test and flight operations is necessary to provide timely assessment of the vibrational characteristics indicating the operational status of turbomachinery and other critical engine components. Efficient performance of this effort is critical due to the significant impact of dynamic evaluation results on ground test and launch schedules, and requires direct familiarity with SSME and derivative systems, test data acquisition, and diagnostic software. Detailed analysis and evaluation of dynamic measurements obtained during SSME and advanced system ground test and flight operations was performed including analytical/statistical assessment of component dynamic behavior, and the development and implementation of analytical/statistical models to efficiently define nominal component dynamic characteristics, detect anomalous behavior, and assess machinery operational condition. In addition, the SSME and J-2 data will be applied to develop vibroacoustic environments for advanced propulsion system components, as required. This study will provide timely assessment of engine component operational status, identify probable causes of malfunction, and indicate feasible engineering solutions. This contract will be performed through accomplishment of negotiated task orders.

  11. Generic evaluation of feedwater transients and small break loss-of-coolant accidents in combustion engineering designed operating plants

    International Nuclear Information System (INIS)

    1980-01-01

    The purpose of this report is to summarize the results of a generic evaluation of feedwater transients, small break loss-of-coolant accidents (LOCAs), and other TMI-2-related events in the Combustion Engineering (CE)-designed operating plants and to establish or confirm the bases for their continued operation. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas

  12. An evaluation of the Davis-Besse loss of feedwater event (June 1985) from an accident management perspective

    International Nuclear Information System (INIS)

    Di Salvo, R.; Leonard, M.T.; Wreathall, J.

    1986-01-01

    An accident management perspective is used to analyze events associated with a total loss-of-feedwater at the Davis-Besse nuclear power plant in June 1985. The relationships of accident management to the closely associated concepts of risk management and emergency management are delineated. The analysis shows that the principal contributors to the event's occurrence were shortcomings in risk management. Successful performance by the operators in accident management was principally responsible for terminating the event without consequence to public health

  13. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

    Energy Technology Data Exchange (ETDEWEB)

    Shah, V.N.; Ware, A.G.; Porter, A.M.

    1997-03-01

    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

  14. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

    International Nuclear Information System (INIS)

    Shah, V.N.; Ware, A.G.; Porter, A.M.

    1997-03-01

    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage

  15. Development of a multi-path ultrasonic flow meter for the application to feedwater flow measurement in nuclear power plants

    International Nuclear Information System (INIS)

    Jong, J. C.; Ha, J. H.; Kim, Y. H.; Jang, W. H.; Park, K. S.; Park, M. S.; Park, M. H.

    2002-01-01

    In this work, we propose a method to measure the feedwater flow using multi-path ultrasonic flow meter (UFM). Since the UFM measures a path velocity at which the ultrasonic wave is propagated, the flow profile may be important to convey the path velocity to the velocity averaged over the entire cross section of the flowing medium. The conventional UFM has used the smooth-wall circular pipe model presented by Nikurades. However, this model covers a lower range which is less than 3.2 million while the Reynolds number of the feedwater flow in operating nuclear power plants (NPPs) is about 20 million. Therefore, we feedwater flow in operating nuclear power plants (NPPs) is about 20 million. Therefore, we proposed the non-linear correlation model that combines the ratio between the DP output and proposed the non-linear correlation model that combines the ratio between the DP output and UFM output. Experiments were performed using both computer simulation and newly constructed NPPs' test data. The uncertainty analysis result shows that the proposed method has reasonably lower uncertainty than conventional UFM

  16. Integration of main directions of development of preschool children in innovative successive educational system “World of Music”

    Directory of Open Access Journals (Sweden)

    Baklanova Tatiana I.

    2016-01-01

    Full Text Available This article describes an innovative education system of pre-school music education “Music world” byT. I. Baklanova, and G. P. Novikova included in a new system of Russian preschool education “Paths”.The education system “Music world” consists of an author’s concept, an integrated programme of musical education, training, development and improvement of health of children of preschool age (3-7 years, two grants for children, methodical recommendations for tutors and musical directors of preschool educational organizations. This education system is developed on a successive basis with a set of textbooks “Music” of T.I. Baklanova (“Planet of Knowledge” series.The integrated approach to five main directions of development and education of children included in the new Federal state educational standards of preschool education is applied in the education system “Music world” for the first time. It is social and communicative, informative, speech, art and esthetic and physical development. These directions are realized in “Music world” in several interconnected interdisciplinary contexts: axiological, cultural and historical, ethnocultural, etc. Polycontextual approach to development of maintenance of preschool music education in combination with integration in him all directions of development and education of preschool children causes scientific novelty, practical importance and efficiency of the education system “Music world”.

  17. Improvement on reliability of control system in power plant

    International Nuclear Information System (INIS)

    Taguchi, S.; Mizumoto, T.; Hirose, Y.; Kashiwai, J.; Takami, I.; Shono, M.; Roji, Y.; Kizaki, S.

    1985-01-01

    Studies made of Japanese PWR operating experiences have revealed that failures in the control system are the primary causes of unscheduled shutdowns. An attempt has, therefore, been made to improve the reliability of the control system in order to raise the plant reliability. The following are the procedures applied to solve the issue; study of operating experiences, fault tree analysis and failure mode and effects analysis. Improvement measures are developed for the control system whose failure threatens to cause the plant trip during the plant life. These systems are the main feedwater control system, rod control system, pressurizer control system and main steam control system in the primary control system. As a result, the plant unavailability is expected to be reduced significantly by applying the improvements. The improvements are applied to the plants under construction and the operating plants in co-operation with utilities and vendors. (author)

  18. The CEGB programme on comparative assessment of alternatives to hydrazine for oxygen removal from aqueous systems

    International Nuclear Information System (INIS)

    Case, B.; Wall, K.H.; Wates, R.W.

    1994-01-01

    The principal conclusion from the programme of work carried out by the CEGB was that Carbohydrazide appeared to be the most promising alternative to hydrazine for boiler feedwater conditioning. It is an effective oxygen scavenger at higher feedwater temperatures, its breakdown products are not a serious threat to circuit materials and it is safer to handle than hydrazine. The main disadvantages, at the time of the study, appeared to be the lack of a simple reliable method for measuring residual carbohydrazide in feedwater and a cost for the alternative chemical of up to 20 times that of hydrazine. (orig.)

  19. Data report of ROSA/LSTF experiment TR-LF-07. Loss-of-feedwater transient with primary feed-and-bleed operation

    International Nuclear Information System (INIS)

    Takeda, Takeshi

    2016-07-01

    An experiment TR-LF-07 was conducted on June 23, 1992 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-IV (ROSA-IV) Program. The ROSA/LSTF experiment TR-LF-07 simulated a loss-of-feedwater transient in a pressurized water reactor (PWR) under assumptions of primary feed-and-bleed operation and total failure of auxiliary feedwater system. A safety injection (SI) signal was generated when steam generator (SG) secondary-side collapsed liquid level decreased to 3 m. Primary depressurization was initiated by fully opening a power-operated relief valve (PORV) of pressurizer (PZR) 30 min after the SI signal. High pressure injection (HPI) system was started in loop with PZR 12 s after the SI signal, while it was initiated in loop without PZR when the primary pressure decreased to 10.7 MPa. The primary and SG secondary pressures were kept almost constant because of cycle opening of the PZR PORV and SG relief valves. The PZR liquid level began to drop steeply following the PORV full opening, which caused liquid level formation at the hot leg. The PZR and hot leg liquid levels recovered due to the HPI actuation in both loops. The primary pressure became lower than the SG secondary pressure, which resulted in the actuation of accumulator (ACC) system in both loops. The PZR and hot legs became full of liquid again after the ACC actuation. The primary feed-and-bleed operation by use of the PORV, HPI and ACC systems was effective to core cooling because of no core uncovery. The experiment was terminated when the continuous core cooling was confirmed due to the successive coolant injection from the HPI system even after the ACC termination. The obtained data would be useful to study operator actions and procedures in the PWR multiple fault events which behaviors in the PZR affect. This report summarizes the test procedures, conditions and major observation in the ROSA/LSTF experiment TR-LF-07. (author)

  20. Practical use of valve seating machine with remote control system for main steam isolation valve at N.P.S

    International Nuclear Information System (INIS)

    Ito, Sadao; Noda, Hiroshi; Sadamura, Morito; Utsunomiya, Yasushi.

    1975-01-01

    The main steam isolation valves in BWR power stations are installed at the boundary of reactor containment vessels, and 2 valves in each main steam system total 8 valves in a plant. They are pneumatically operated Y type globe valves for preventing the release of radioactive substances in the atmosphere in case of the breaking of main steam pipes and also preventing the loss of coolant in case of the breaking of recirculating equipments. Therefore careful leak test, inspection, and seat-fitting are carried out to the valves at each regular maintenance. The manual maintenance work is difficult because of narrow space and the reduction of exposure, and the seat-fitting work requires the skill of high degree, therefore Okano Valve Manufacturing Co. and Tokyo Electric Power Co. jointly started the research and development of an automatic valve seating machine, and successfully put it to practical use in Fukushima No.1 Nuclear Power Station in Nov. 1974. First, the problems in the manual seat-fitting work were investigated, and the means to mechanically solve them were materialized with a prototype machine. After its mock-up test, an actual machine was designed and manufactured. The test result showed remarkable reduction of exposure and labor-saving, and the leak evaluation was sufficiently below the allowable value. (Kako, I.)