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Sample records for feedwater system seismic

  1. 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

  2. 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

  3. 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)

  4. 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.)

  5. 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.)

  6. 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.)

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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.)

  12. 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

  13. 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.

  14. 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.)

  15. 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

  16. 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

  17. 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)

  18. 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)

  19. 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

  20. 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)

  1. 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)

  2. 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

  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. 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)

  5. 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)

  6. 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)

  7. 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)

  8. 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)

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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)

  14. 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.

  15. 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

  16. 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

  17. 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...

  18. 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)

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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)

  5. Seismic intrusion detector system

    Science.gov (United States)

    Hawk, Hervey L.; Hawley, James G.; Portlock, John M.; Scheibner, James E.

    1976-01-01

    A system for monitoring man-associated seismic movements within a control area including a geophone for generating an electrical signal in response to seismic movement, a bandpass amplifier and threshold detector for eliminating unwanted signals, pulse counting system for counting and storing the number of seismic movements within the area, and a monitoring system operable on command having a variable frequency oscillator generating an audio frequency signal proportional to the number of said seismic movements.

  6. 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

  7. 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.)

  8. 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

  9. 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)

  10. 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.)

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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)

  16. 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.)

  17. Seismic data acquisition systems

    International Nuclear Information System (INIS)

    Kolvankar, V.G.; Nadre, V.N.; Rao, D.S.

    1989-01-01

    Details of seismic data acquisition systems developed at the Bhabha Atomic Research Centre, Bombay are reported. The seismic signals acquired belong to different signal bandwidths in the band from 0.02 Hz to 250 Hz. All these acquisition systems are built around a unique technique of recording multichannel data on to a single track of an audio tape and in digital form. Techniques of how these signals in different bands of frequencies were acquired and recorded are described. Method of detecting seismic signals and its performance is also discussed. Seismic signals acquired in different set-ups are illustrated. Time indexing systems for different set-ups and multichannel waveform display systems which form essential part of the data acquisition systems are also discussed. (author). 13 refs., 6 figs., 1 tab

  18. 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

  19. 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.)

  20. 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

  1. 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.)

  2. 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.

  3. 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

  4. 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.)

  5. 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)

  6. Induced Seismicity Monitoring System

    Science.gov (United States)

    Taylor, S. R.; Jarpe, S.; Harben, P.

    2014-12-01

    There are many seismological aspects associated with monitoring of permanent storage of carbon dioxide (CO2) in geologic formations. Many of these include monitoring underground gas migration through detailed tomographic studies of rock properties, integrity of the cap rock and micro seismicity with time. These types of studies require expensive deployments of surface and borehole sensors in the vicinity of the CO2 injection wells. Another problem that may exist in CO2 sequestration fields is the potential for damaging induced seismicity associated with fluid injection into the geologic reservoir. Seismic hazard monitoring in CO2 sequestration fields requires a seismic network over a spatially larger region possibly having stations in remote settings. Expensive observatory-grade seismic systems are not necessary for seismic hazard deployments or small-scale tomographic studies. Hazard monitoring requires accurate location of induced seismicity to magnitude levels only slightly less than that which can be felt at the surface (e.g. magnitude 1), and the frequencies of interest for tomographic analysis are ~1 Hz and greater. We have developed a seismo/acoustic smart sensor system that can achieve the goals necessary for induced seismicity monitoring in CO2 sequestration fields. The unit is inexpensive, lightweight, easy to deploy, can operate remotely under harsh conditions and features 9 channels of recording (currently 3C 4.5 Hz geophone, MEMS accelerometer and microphone). An on-board processor allows for satellite transmission of parameter data to a processing center. Continuous or event-detected data is kept on two removable flash SD cards of up to 64+ Gbytes each. If available, data can be transmitted via cell phone modem or picked up via site visits. Low-power consumption allows for autonomous operation using only a 10 watt solar panel and a gel-cell battery. The system has been successfully tested for long-term (> 6 months) remote operations over a wide range

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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.)

  17. 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

  18. 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

  19. 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)

  20. 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.)

  1. 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)

  2. Integrated system for seismic evaluations

    International Nuclear Information System (INIS)

    Xu, J.; Philippacopoulos, A.J.; Miller, C.A.; Costantino, C.J.; Graves, H.

    1989-01-01

    This paper describes the various features of the seismic module of the CARES system (computer analysis for rapid evaluation of structures). This system was developed to perform rapid evaluations of structural behavior and capability of nuclear power plant facilities. The CARES is structural in a modular format. Each module performs a specific type of analysis i.e., static or dynamic, linear or nonlinear, etc. This paper describes the features of the seismic module in particular. The development of the seismic modules of the CARES system is based on an approach which incorporates major aspects of seismic analysis currently employed by the industry into an integrated system that allows for carrying out interactively computations of structural response to seismic motions. The code operates on a PC computer system and has multi-graphics capabilities

  3. 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

  4. Integrated system for seismic evaluations

    International Nuclear Information System (INIS)

    Xu, J.; Philippacopoulos, A.J.; Miller, C.A.; Costantino, C.J.; Graves, H.

    1989-01-01

    This paper describes the various features of the Seismic Module of the CARES system (Computer Analysis for Rapid Evaluation of Structures). This system was developed by Brookhaven National Laboratory (BNL) for the US Nuclear Regulatory Commission to perform rapid evaluations of structural behavior and capability of nuclear power plant facilities. The CARES is structured in a modular format. Each module performs a specific type of analysis i.e., static or dynamic, linear or nonlinear, etc. This paper describes the features of the Seismic Module in particular. The development of the Seismic Module of the CARES system is based on an approach which incorporates all major aspects of seismic analysis currently employed by the industry into an integrated system that allows for carrying out interactively computations of structural response to seismic motions. The code operates on a PC computer system and has multi-graphics capabilities. It has been designed with user friendly features and it allows for interactive manipulation of various analysis phases during the seismic design process. The capabilities of the seismic module include (a) generation of artificial time histories compatible with given design ground response spectra, (b) development of Power Spectral Density (PSD) functions associated with the seismic input, (c) deconvolution analysis using vertically propagating shear waves through a given soil profile, and (d) development of in-structure response spectra or corresponding PSD's. It should be pointed out that these types of analyses can also be performed individually by using available computer codes such as FLUSH, SAP, etc. The uniqueness of the CARES, however, lies on its ability to perform all required phases of the seismic analysis in an integrated manner. 5 refs., 6 figs

  5. 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

  6. 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

  7. 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)

  8. 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

  9. 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)

  10. 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)

  11. 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.

  12. Seismic design of piping systems

    International Nuclear Information System (INIS)

    Anglaret, G.; Beguin, J.L.

    1986-01-01

    This paper deals with the method used in France for the PWR nuclear plants to derive locations and types of supports of auxiliary and secondary piping systems taking earthquake in account. The successive steps of design are described, then the seismic computation method and its particular conditions of applications for piping are presented. The different types of support (and especially seismic ones) are described and also their conditions of installation. The method used to compare functional tests results and computation results in order to control models is mentioned. Some experiments realised on site or in laboratory, in order to validate models and methods, are presented [fr

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. An economical educational seismic system

    Science.gov (United States)

    Lehman, J. D.

    1980-01-01

    There is a considerable interest in seismology from the nonprofessional or amateur standpoint. The operation of a seismic system can be satisfying and educational, especially when you have built and operated the system yourself. A long-period indoor-type sensor and recording system that works extremely well has been developed in the James Madison University Physics Deparment. The system can be built quite economically, and any educational institution that cannot commit themselves to a professional installation need not be without first-hand seismic information. The system design approach has been selected by college students working a project or senior thesis, several elementary and secondary science teachers, as well as the more ambitious tinkerer or hobbyist at home 

  19. Advanced Seismic While Drilling System

    Energy Technology Data Exchange (ETDEWEB)

    Robert Radtke; John Fontenot; David Glowka; Robert Stokes; Jeffery Sutherland; Ron Evans; Jim Musser

    2008-06-30

    . An APS Turbine Alternator powered the SeismicPULSER{trademark} to produce two Hz frequency peak signals repeated every 20 seconds. Since the ION Geophysical, Inc. (ION) seismic survey surface recording system was designed to detect a minimum downhole signal of three Hz, successful performance was confirmed with a 5.3 Hz recording with the pumps running. The two Hz signal generated by the sparker was modulated with the 3.3 Hz signal produced by the mud pumps to create an intense 5.3 Hz peak frequency signal. The low frequency sparker source is ultimately capable of generating selectable peak frequencies of 1 to 40 Hz with high-frequency spectra content to 10 kHz. The lower frequencies and, perhaps, low-frequency sweeps, are needed to achieve sufficient range and resolution for realtime imaging in deep (15,000 ft+), high-temperature (150 C) wells for (a) geosteering, (b) accurate seismic hole depth, (c) accurate pore pressure determinations ahead of the bit, (d) near wellbore diagnostics with a downhole receiver and wired drill pipe, and (e) reservoir model verification. Furthermore, the pressure of the sparker bubble will disintegrate rock resulting in an increased overall rates of penetration. Other applications for the SeismicPULSER{trademark} technology are to deploy a low-frequency source for greater range on a wireline for Reverse Vertical Seismic Profiling (RVSP) and Cross-Well Tomography. Commercialization of the technology is being undertaken by first contacting stakeholders to define the value proposition for rig site services utilizing SeismicPULSER{trademark} technologies. Stakeholders include national oil companies, independent oil companies, independents, service companies, and commercial investors. Service companies will introduce a new Drill Bit SWD service for deep HTHP wells. Collaboration will be encouraged between stakeholders in the form of joint industry projects to develop prototype tools and initial field trials. No barriers have been identified

  20. 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)

  1. 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

  2. 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

  3. 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

  4. Seismic design practices for power systems

    International Nuclear Information System (INIS)

    Schiff, A.J.

    1991-01-01

    In this paper, the evolution of seismic design practices in electric power systems is reviewed. In California the evolution had led to many installation practices that are directed at improving the seismic ruggedness of power system facilities, particularly high voltage substation equipment. The primary means for substantiating the seismic ruggedness of important, hard to analyze substation equipment is through vibration testing. Current activities include system evaluations, development of emergency response plans and their exercise, and review elements that impact the entire system, such as energy control centers and communication systems. From a national perspective there is a need to standardize seismic specifications, identify a seismic specialist within each utility and enhance communications among these specialists. There is a general need to incorporate good seismic design practices on a national basis emphasizing new construction

  5. 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)

  6. 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

  7. 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.

  8. 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.)

  9. 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.

  10. 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.)

  11. 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

  12. 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

  13. Improvement of seismic observation systems in JOYO

    International Nuclear Information System (INIS)

    Sumino, Kozo; Suto, Masayoshi; Tanaka, Akihiro

    2013-01-01

    In the experimental fast reactor 'Joyo' in order to perform the seismic observation in and around the building block and ground, SMAC type seismographs had continuously been used for about 38 years. However, this equipment aged, and the 2011 off the Pacific Coast of Tohoku Earthquake on Mach 11, 2011 increased the importance of seismic data of the reactor facilities from the viewpoint of earthquake-proof safety. For these reasons, Joyo updated the system to the seismic observation system reflecting the latest technology/information, while keeping consistency with the observation data of the former seismographs (SMAC type seismograph). This updating improved various problems on the former observation seismographs. In addition, the installation of now observation points in the locations that are important in seismic safety evaluation expanded the data, and further improved the reliability of the seismic observation and evaluation on 'Joyo'. (A.O.)

  14. Seismic proving test of process computer systems with a seismic floor isolation system

    International Nuclear Information System (INIS)

    Fujimoto, S.; Niwa, H.; Kondo, H.

    1995-01-01

    The authors have carried out seismic proving tests for process computer systems as a Nuclear Power Engineering Corporation (NUPEC) project sponsored by the Ministry of International Trade and Industry (MITI). This paper presents the seismic test results for evaluating functional capabilities of process computer systems with a seismic floor isolation system. The seismic floor isolation system to isolate the horizontal motion was composed of a floor frame (13 m x 13 m), ball bearing units, and spring-damper units. A series of seismic excitation tests was carried out using a large-scale shaking table of NUPEC. From the test results, the functional capabilities during large earthquakes of computer systems with a seismic floor isolation system were verified

  15. 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

  16. Multichannel long period seismic data acquisition system

    International Nuclear Information System (INIS)

    Kolvankar, V.G.; Rao, D.S.

    1990-01-01

    This paper discusses the specifications and performance of an eight channel long period seismic digital data acquisition system, which is developed and installed at Seismic Array Station, Gauribidanur, Karnataka State. The paper describes how these data in an unedited form are recorded on a single track of magnetic tape inter-mittantly, which has resulted in recording of 50 days data on a single tapespool. A time indexing technique which enables quick access to any desired portion of a recorded tape is also discussed. Typical examples of long period seismic event signals recorded by this system are also illustrated. Various advantages, the system provides over the analog multichannel instrumentation tape recording system, operating at Seismic Array Station for th e last two decades, are also discussed. (author). 7 figs

  17. Seismic margins and calibration of piping systems

    International Nuclear Information System (INIS)

    Shieh, L.C.; Tsai, N.C.; Yang, M.S.; Wong, W.L.

    1985-01-01

    The Seismic Safety Margins Research Program (SSMRP) is a US Nuclear Regulatory Commission-funded, multiyear program conducted by Lawrence Livermore National Laboratory (LLNL). Its objective is to develop a complete, fully coupled analysis procedure for estimating the risk of earthquake-induced radioactive release from a commercial nuclear power plant and to determine major contributors to the state-of-the-art seismic and systems analysis process and explicitly includes the uncertainties in such a process. The results will be used to improve seismic licensing requirements for nuclear power plants. In Phase I of SSMRP, the overall seismic risk assessment methodology was developed and assembled. The application of this methodology to the seismic PRA (Probabilistic Risk Assessment) at the Zion Nuclear Power Plant has been documented. This report documents the method deriving response factors. The response factors, which relate design calculated responses to best estimate values, were used in the seismic response determination of piping systems for a simplified seismic probablistic risk assessment. 13 references, 31 figures, 25 tables

  18. 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

  19. Seismic analysis of a nonlinear airlock system

    International Nuclear Information System (INIS)

    Huang, S.N.

    1983-01-01

    The containment equipment airlock door of the Fast Flux Test Facility utilizes screw-type actuators as a push-pull mechanism for closing and opening operations. Special design features were used to protect these actuators from pressure differential loading. These made the door behave as a nonlinear system during a seismic event. Seismic analyses, utilizing the time history method, were conducted to determine the seismic loads on these scew-type actuators. Several sizes of actuators were examined. Procedures for determining the final optimum design are discussed in detail

  20. 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)

  1. 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)

  2. 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)

  3. 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

  4. 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.

  5. 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.

  6. Systems considerations in seismic margin evaluations

    International Nuclear Information System (INIS)

    Buttermer, D.R.

    1987-01-01

    Increasing knowledge in the geoscience field has led to the understanding that, although highly unlikely, it is possible for a nuclear power plant to be subjected to earthquake ground motion greater than that for which the plant was designed. While it is recognized that there are conservatisms inherent in current design practices, interest has developed in evaluating the seismic risk of operating plants. Several plant-specific seismic probabilistic risk assessments (SPRA) have been completed to address questions related to the seismic risk of a plant. The results from such SPRAs are quite informative, but such studies may entail a considerable amount of expensive analysis of large portions of the plant. As an alternative to an SPRA, it may be more practical to select an earthquake level above the design basis for which plant survivability is to be demonstrated. The principal question to be addressed in a seismic margin evaluation is: At what ground motion levels does one have a high confidence that the probability of seismically induced core damage is sufficiently low? In a seismic margin evaluation, an earthquake level is selected (based on site-specific geoscience considerations) for which a stable, long-term safe shutdown condition is to be demonstrated. This prespecified earthquake level is commonly referred to as the seismic margin earthquake (SME). The Electric Power Research Institute is currently supporting a research project to develop procedures for use by the utilities to allow them to perform nuclear plant seismic margin evaluations. This paper describes the systems-related aspects of these procedures

  7. Feasibility of seismic alert systems in India

    International Nuclear Information System (INIS)

    Chauhan, P.K.S.; Pandey, Y.

    2012-01-01

    Natural disasters like flood, earthquakes and cyclones are very frequent in India since historical times. As far as the casualties are concerned, globally earthquakes are second in the list after the flood. The loss of property due to these earthquakes is huge and enormous. In the light of the present knowledge base, earthquake prediction is far from being a reality. An early earthquake warning has potential to save the precious human lives. In the present day scenario seismic instrumentation and telecommunication permits the implementation of seismic alert system (SAS) based on the real-time measurement of ground motions near the source. SAS is capable of providing a warning of several seconds before the arrival of destructive seismic waves caused by a large earthquake. SAS is successfully operational in many countries of the world. In a country, like India where earthquakes are taking heavy toll on the human lives and property, seismic alert system may prove to be very important step in natural hazard mitigation strategy. In this paper, an attempt has been made to compute the available alarm time before the destructive earthquake waves reaches to the cities like Delhi, Lucknow, Patna and Kolkata taking Himalaya as the source and feasibility of seismic alert system in Indian scenario. (author)

  8. 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)

  9. Optical seismic sensor systems and methods

    Science.gov (United States)

    Beal, A. Craig; Cummings, Malcolm E.; Zavriyev, Anton; Christensen, Caleb A.; Lee, Keun

    2015-12-08

    Disclosed is an optical seismic sensor system for measuring seismic events in a geological formation, including a surface unit for generating and processing an optical signal, and a sensor device optically connected to the surface unit for receiving the optical signal over an optical conduit. The sensor device includes at least one sensor head for sensing a seismic disturbance from at least one direction during a deployment of the sensor device within a borehole of the geological formation. The sensor head includes a frame and a reference mass attached to the frame via at least one flexure, such that movement of the reference mass relative to the frame is constrained to a single predetermined path.

  10. Extreme loads seismic testing of conduit systems

    International Nuclear Information System (INIS)

    Howard, G.E.; Ibanez, P.; Harrison, S.; Shi, Z.T.

    1991-01-01

    Rigid steel conduit (thin-wall tubes with threaded connections) containing electrical cabling are a common feature in nuclear power plants. Conduit systems are in many cases classified in U.S.A. practice as Seismic Category I structures. this paper summarizes results and others aspects of a dynamic test program conducted to investigate conduit systems seismic performance under three-axis excitation for designs representative at a nuclear power plant sited near Ft. Worth, Texas (a moderate seismic zone), with a Safe Shutdown Earthquake (SSE) of 0.12 g. Test specimens where subjected to postulated seismic events, including excitation well in excess of Safe Shutdown Earthquake events typical for U.S.A. nuclear power stations. A total of 18 conduit systems of 9-meter nominal lengths were shake table mounted and subjected to a variety of tests. None of the specimens suffered loss of load capacity when subjected to a site-enveloping Safe Shutdown Earthquake (SSE). Clamp/attachment hardware failures only began to occur when earthquake input motion was scaled upward to minimum values of 2.3-4.6 times site enveloping SSE response spectra. Tensile and/or shear failure of clamp attachment bolts or studs was the failure mode in all case in which failure was induced. (author)

  11. Seismic analysis of nuclear piping system

    International Nuclear Information System (INIS)

    Shrivastava, S.K.; Pillai, K.R.V.; Nandakumar, S.

    1975-01-01

    To illustrate seismic analysis of nuclear power plant piping, a simple piping system running between two floors of the reactor building is assumed. Reactor building floor response is derived from time-history method. El Centre earthquake (1940) accelerogram is used for time-history analysis. The piping system is analysed as multimass lumped system. Behaviour of the pipe during the said earthquake is discussed. (author)

  12. 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)

  13. Development of Vertical Cable Seismic System

    Science.gov (United States)

    Asakawa, E.; Murakami, F.; Sekino, Y.; Okamoto, T.; Ishikawa, K.; Tsukahara, H.; Shimura, T.

    2011-12-01

    In 2009, Ministry of Education, Culture, Sports, Science and Technology(MEXT) started the survey system development for Hydrothermal deposit. We proposed the Vertical Cable Seismic (VCS), the reflection seismic survey with vertical cable above seabottom. VCS has the following advantages for hydrothermal deposit survey. (1) VCS is an efficient high-resolution 3D seismic survey in limited area. (2) It achieves high-resolution image because the sensors are closely located to the target. (3) It avoids the coupling problems between sensor and seabottom that cause serious damage of seismic data quality. (4) Because of autonomous recording system on sea floor, various types of marine source are applicable with VCS such as sea-surface source (GI gun etc.) , deep-towed or ocean bottom source. Our first experiment of 2D/3D VCS surveys has been carried out in Lake Biwa, JAPAN, in November 2009. The 2D VCS data processing follows the walk-away VSP, including wave field separation and depth migration. Seismic Interferometry technique is also applied. The results give much clearer image than the conventional surface seismic. Prestack depth migration is applied to 3D data to obtain good quality 3D depth volume. Seismic Interferometry technique is applied to obtain the high resolution image in the very shallow zone. Based on the feasibility study, we have developed the autonomous recording VCS system and carried out the trial experiment in actual ocean at the water depth of about 400m to establish the procedures of deployment/recovery and to examine the VC position or fluctuation at seabottom. The result shows that the VC position is estimated with sufficient accuracy and very little fluctuation is observed. Institute of Industrial Science, the University of Tokyo took the research cruise NT11-02 on JAMSTEC R/V Natsushima in February, 2011. In the cruise NT11-02, JGI carried out the second VCS survey using the autonomous VCS recording system with the deep towed source provided by

  14. 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

  15. Induced seismicity associated with enhanced geothermal system

    Energy Technology Data Exchange (ETDEWEB)

    Majer, Ernest; Majer, Ernest L.; Baria, Roy; Stark, Mitch; Oates, Stephen; Bommer, Julian; Smith, Bill; Asanuma, Hiroshi

    2006-09-26

    Enhanced Geothermal Systems (EGS) offer the potential to significantly add to the world energy inventory. As with any development of new technology, some aspects of the technology has been accepted by the general public, but some have not yet been accepted and await further clarification before such acceptance is possible. One of the issues associated with EGS is the role of microseismicity during the creation of the underground reservoir and the subsequent extraction of the energy. The primary objectives of this white paper are to present an up-to-date review of the state of knowledge about induced seismicity during the creation and operation of enhanced geothermal systems, and to point out the gaps in knowledge that if addressed will allow an improved understanding of the mechanisms generating the events as well as serve as a basis to develop successful protocols for monitoring and addressing community issues associated with such induced seismicity. The information was collected though literature searches as well as convening three workshops to gather information from a wide audience. Although microseismicity has been associated with the development of production and injection operations in a variety of geothermal regions, there have been no or few adverse physical effects on the operations or on surrounding communities. Still, there is public concern over the possible amount and magnitude of the seismicity associated with current and future EGS operations. It is pointed out that microseismicity has been successfully dealt with in a variety of non-geothermal as well as geothermal environments. Several case histories are also presented to illustrate a variety of technical and public acceptance issues. It is concluded that EGS Induced seismicity need not pose any threat to the development of geothermal resources if community issues are properly handled. In fact, induced seismicity provides benefits because it can be used as a monitoring tool to understand the

  16. Seismic monitoring: a unified system for research and verifications

    International Nuclear Information System (INIS)

    Thigpen, L.

    1979-01-01

    A system for characterizing either a seismic source or geologic media from observational data was developed. This resulted from an examination of the forward and inverse problems of seismology. The system integrates many seismic monitoring research efforts into a single computational capability. Its main advantage is that it unifies computational and research efforts in seismic monitoring. 173 references, 9 figures, 3 tables

  17. 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

  18. Seismic attenuation system for a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liszkai, Tamas; Cadell, Seth

    2018-01-30

    A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

  19. 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)

  20. 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

  1. Effects of applying three-dimensional seismic isolation system on the seismic design of FBR

    International Nuclear Information System (INIS)

    Hirata, Kazuta; Yabana, Shuichi; Kanazawa, Kenji; Matsuda, Akihiro

    1997-01-01

    In this study conceptional three-dimensional seismic isolation system for fast breeder reactor (FBR) is proposed. Effects of applying three-dimensional seismic isolation system on the seismic design for the FBR equipment are evaluated quantitatively. From the evaluation, it is concluded following effects are expected by applying the three-dimensional seismic isolation system to the FBR and the effects are evaluated quantitatively. (1) Reduction of membrane thickness of the reactor vessel (2) Suppression of uplift of fuels by reducing vertical seismic response of the core (3) Reduction of the supports for the piping system (4) Three-dimensional base isolation system for the whole reactor building is advantageous to the combined isolation system of horizontal base isolation for the reactor building and vertical isolation for the equipment. (author)

  2. 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)

  3. 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)

  4. 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

  5. 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)

  6. Development and seismic evaluation of the seismic monitoring analysis system for HANARO

    International Nuclear Information System (INIS)

    Ryu, J. S.; Youn, D. B.; Kim, H. G.; Woo, J. S.

    2003-01-01

    Since the start of operation, the seismic monitoring system has been utilized for monitoring an earthquake at the HANARO site. The existing seismic monitoring system consists of field sensors and monitoring panel. The analog-type monitoring system with magnetic tape recorder is out-of-date model. In addition, the disadvantage of the existing system is that it does not include signal-analyzing equipment. Therefore, we have improved the analog seismic monitoring system except the field sensors into a new digital Seismic Monitoring Analysis System(SMAS) that can monitor and analyze earthquake signals. To achieve this objective for HANARO, the digital type hardware of the SMAS has been developed. The seismic monitoring and analysis programs that can provide rapid and precise information for an earthquake were developed. After the installation of the SMAS, we carried out the Site Acceptance Test (SAT) to confirm the functional capability of the newly developed system. The results of the SAT satisfy the requirements of the fabrication technical specifications. In addition, the seismic characteristics and structural integrity of the SMAS were evaluated. The results show that the cabinet of SMAS can withstand the effects of seismic loads and remain functional. This new SMAS is operating in the HANARO instrument room to acquire and analyze the signal of an earthquake

  7. Parameters of the Seismic system in Armenia

    Energy Technology Data Exchange (ETDEWEB)

    Karapetyan, N.K.

    1976-01-01

    An examination is made of the seismic system parameters in Armenia and the adjoining regions of Azerbaidzhan, Georgia, Iran, and Turkey. Data are given on correlations between the energy class, magnitude and intensity scale of earthquakes, and values for the level of activity and angular coefficients as a function of the region under examination, the time of observation, and method of determination; and diagrams are presented which illustrate the pattern of earthquake recurrence for the period 1679 to 1968, and observation times essential for determining earthquake recurrence with a given accuracy of 10% for the Armenian Highlands. 3 references, 2 figures, 2 tables.

  8. Study on three dimensional seismic isolation system

    International Nuclear Information System (INIS)

    Morishita, Masaki; Kitamura, Seiji

    2003-01-01

    Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) launched joint research programs on structural design and three-dimensional seismic isolation technologies, as part of the supporting R and D activities for the feasibility studies on commercialized fast breeder reactor cycle systems. A research project by JAPC under the auspices of the Ministry of Economy, Trade, and Industry (METI) with technical support by JNC is included in this joint study. This report contains the results of the research on the three-dimensional seismic isolation technologies, and the results of this year's study are summarized in the following five aspects. (1) Study on Earthquake Condition for Developing 3-dimensional Base Isolation System. The case study S2 is one of the maximum ground motions, of which the records were investigated up to this time. But a few observed near the fault exceed the case study S2 in the long period domain, depending on the fault length and conditions. Generally it is appropriate that the response spectra ratio (vertical/horizontal) is 0.6. (2) Performance Requirement for 3-dimensional Base Isolation System and Devices. Although the integrity map of main equipment/piping dominate the design criteria for the 3-dimensional base isolation system, the combined integrity map is the same as those of FY 2000, which are under fv=1Hz and over hv=20%. (3) Developing Targets and Schedule for 3-dimensional Isolation Technology. The target items for 3-dimensional base isolation system were rearranged into a table, and developing items to be examined concerning the device were also adjusted. A development plan until FY 2009 was made from the viewpoint of realization and establishment of a design guideline on 3-dimensional base isolation system. (4) Study on 3-dimensional Entire Building Base Isolation System. Three ideas among six ideas that had been proposed in FY2001, i.e., '3-dimensional base isolation system incorporating hydraulic

  9. 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

  10. 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

  11. 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)

  12. 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.)

  13. Development of Vertical Cable Seismic System (2)

    Science.gov (United States)

    Asakawa, E.; Murakami, F.; Tsukahara, H.; Ishikawa, K.

    2012-12-01

    The vertical cable seismic is one of the reflection seismic methods. It uses hydrophone arrays vertically moored from the seafloor to record acoustic waves generated by surface, deep-towed or ocean bottom sources. Analyzing the reflections from the sub-seabed, we could look into the subsurface structure. This type of survey is generally called VCS (Vertical Cable Seismic). Because VCS is an efficient high-resolution 3D seismic survey method for a spatially-bounded area, we proposed the method for the hydrothermal deposit survey tool development program that the Ministry of Education, Culture, Sports, Science and Technology (MEXT) started in 2009. We are now developing a VCS system, including not only data acquisition hardware but data processing and analysis technique. Our first experiment of VCS surveys has been carried out in Lake Biwa, JAPAN in November 2009 for a feasibility study. Prestack depth migration is applied to the 3D VCS data to obtain a high quality 3D depth volume. Based on the results from the feasibility study, we have developed two autonomous recording VCS systems. After we carried out a trial experiment in the actual ocean at a water depth of about 400m and we carried out the second VCS survey at Iheya Knoll with a deep-towed source. In this survey, we could establish the procedures for the deployment/recovery of the system and could examine the locations and the fluctuations of the vertical cables at a water depth of around 1000m. The acquired VCS data clearly shows the reflections from the sub-seafloor. Through the experiment, we could confirm that our VCS system works well even in the severe circumstances around the locations of seafloor hydrothermal deposits. We have carried out two field surveys in 2011. One is a 3D survey with a boomer for a high-resolution surface source and the other one for an actual field survey in the Izena Cauldron an active hydrothermal area in the Okinawa Trough. Through these surveys, we have confirmed that the

  14. GSETT-3: testing the experimental international seismic monitoring system

    International Nuclear Information System (INIS)

    Ringdal, Frode

    1995-01-01

    Global seismic monitoring system has been developed by the Conference on Disarmaments (CDs) ad hoc group of scientific experts to consider international cooperative measures to detect and identify seismic events (the GSE), based in Geneva. In the course of its work, the GSE has conducted two large-scale global technical tests, Global Seismic Events Technical Test-1 (GSETT-1) in 1984 and GSETT-2 in 1991. The GSE has now embarked upon its third and most ambitious technical test, GSETT-3, which will encompass the development, testing and evaluation of a working prototype of the eventual Comprehensive Test Ban Treaty (CTBT) seismic monitoring system

  15. Development of Vertical Cable Seismic System (3)

    Science.gov (United States)

    Asakawa, E.; Murakami, F.; Tsukahara, H.; Mizohata, S.; Ishikawa, K.

    2013-12-01

    The VCS (Vertical Cable Seismic) is one of the reflection seismic methods. It uses hydrophone arrays vertically moored from the seafloor to record acoustic waves generated by surface, deep-towed or ocean bottom sources. Analyzing the reflections from the sub-seabed, we could look into the subsurface structure. Because VCS is an efficient high-resolution 3D seismic survey method for a spatially-bounded area, we proposed the method for the hydrothermal deposit survey tool development program that the Ministry of Education, Culture, Sports, Science and Technology (MEXT) started in 2009. We are now developing a VCS system, including not only data acquisition hardware but data processing and analysis technique. We carried out several VCS surveys combining with surface towed source, deep towed source and ocean bottom source. The water depths of the survey are from 100m up to 2100m. The target of the survey includes not only hydrothermal deposit but oil and gas exploration. Through these experiments, our VCS data acquisition system has been completed. But the data processing techniques are still on the way. One of the most critical issues is the positioning in the water. The uncertainty in the positions of the source and of the hydrophones in water degraded the quality of subsurface image. GPS navigation system are available on sea surface, but in case of deep-towed source or ocean bottom source, the accuracy of shot position with SSBL/USBL is not sufficient for the very high-resolution imaging. We have developed another approach to determine the positions in water using the travel time data from the source to VCS hydrophones. In the data acquisition stage, we estimate the position of VCS location with slant ranging method from the sea surface. The deep-towed source or ocean bottom source is estimated by SSBL/USBL. The water velocity profile is measured by XCTD. After the data acquisition, we pick the first break times of the VCS recorded data. The estimated positions of

  16. An assessment of seismic monitoring in the United States; requirement for an Advanced National Seismic System

    Science.gov (United States)

    ,

    1999-01-01

    This report assesses the status, needs, and associated costs of seismic monitoring in the United States. It sets down the requirement for an effective, national seismic monitoring strategy and an advanced system linking national, regional, and urban monitoring networks. Modernized seismic monitoring can provide alerts of imminent strong earthquake shaking; rapid assessment of distribution and severity of earthquake shaking (for use in emergency response); warnings of a possible tsunami from an offshore earthquake; warnings of volcanic eruptions; information for correctly characterizing earthquake hazards and for improving building codes; and data on response of buildings and structures during earthquakes, for safe, cost-effective design, engineering, and construction practices in earthquake-prone regions.

  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. Seismic qualification of a commercial grade emergency diesel generator system in high seismic zones

    International Nuclear Information System (INIS)

    Khan, Mohsin R.; Chen, Wayne W.H.; Chu, Winnie S.

    2004-01-01

    The paper presents the seismic qualification of a commercially procured emergency diesel generator (EDG) system for use in a nuclear power plant. Response spectrum analyses of finite element models, validated using in situ vibration test data, were performed to qualify the skid and floor mounted mechanical components whose functional capacity and structural integrity can be analyzed. Time history analyses of these models were also performed to obtain the amplified response spectra for seismic testing of small valves, electrical and electro-mechanical components whose functional capacity can not be analyzed to establish the seismic qualification. The operational loads were obtained by in-plant vibration monitoring. Full scale shake table testing was performed for auxiliary electrical cabinets. It is concluded that with some minor structural modifications, a commercial grade EDG system can be qualified for safety-related applications in nuclear power plants located in high seismic zones. (author)

  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. IDEF method for designing seismic information system in CTBT verification

    International Nuclear Information System (INIS)

    Zheng Xuefeng; Shen Junyi; Jin Ping; Zhang Huimin; Zheng Jiangling; Sun Peng

    2004-01-01

    Seismic information system is of great importance for improving the capability of CTBT verification. A large amount of money has been appropriated for the research in this field in the U.S. and some other countries in recent years. However, designing and developing a seismic information system involves various technologies about complex system design. This paper discusses the IDEF0 method to construct function models and the IDEF1x method to make information models systemically, as well as how they are used in designing seismic information system in CTBT verification. (authors)

  1. Letter report seismic shutdown system failure mode and effect analysis

    International Nuclear Information System (INIS)

    KECK, R.D.

    1999-01-01

    The Supply Ventilation System Seismic Shutdown ensures that the 234-52 building supply fans, the dry air process fans and vertical development calciner are shutdown following a seismic event. This evaluates the failure modes and determines the effects of the failure modes

  2. Seismic fragility test of a 6-inch diameter pipe system

    International Nuclear Information System (INIS)

    Chen, W.P.; Onesto, A.T.; DeVita, V.

    1987-02-01

    This report contains the test results and assessments of seismic fragility tests performed on a 6-inch diameter piping system. The test was funded by the US Nuclear Regulatory Commission (NRC) and conducted by ETEC. The objective of the test was to investigate the ability of a representative nuclear piping system to withstand high level dynamic seismic and other loadings. Levels of loadings achieved during seismic testing were 20 to 30 times larger than normal elastic design evaluations to ASME Level D limits would permit. Based on failure data obtained during seismic and other dynamic testing, it was concluded that nuclear piping systems are inherently able to withstand much larger dynamic seismic loadings than permitted by current design practice criteria or predicted by the probabilistic risk assessment (PRA) methods and several proposed nonlinear methods of failure analysis

  3. 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

  4. 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.

  5. 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

  6. 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

  7. Development of an evaluation method for seismic isolation systems of nuclear power facilities. Seismic design analysis methods for crossover piping system

    International Nuclear Information System (INIS)

    Tai, Koichi; Sasajima, Keisuke; Fukushima, Shunsuke; Takamura, Noriyuki; Onishi, Shigenobu

    2014-01-01

    This paper provides seismic design analysis methods suitable for crossover piping system, which connects between seismic isolated building and non-isolated building in the seismic isolated nuclear power plant. Through the numerical study focused on the main steam crossover piping system, seismic response spectrum analysis applying ISM (Independent Support Motion) method with SRSS combination or CCFS (Cross-oscillator, Cross-Floor response Spectrum) method has found to be quite effective for the seismic design of multiply supported crossover piping system. (author)

  8. 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)

  9. Origins of a national seismic system in the United States

    Science.gov (United States)

    Filson, John R.; Arabasz, Walter J.

    2016-01-01

    This historical review traces the origins of the current national seismic system in the United States, a cooperative effort that unifies national, regional, and local‐scale seismic monitoring within the structure of the Advanced National Seismic System (ANSS). The review covers (1) the history and technological evolution of U.S. seismic networks leading up to the 1990s, (2) factors that made the 1960s and 1970s a watershed period for national attention to seismology, earthquake hazards, and seismic monitoring, (3) genesis of the vision of a national seismic system during 1980–1983, (4) obstacles and breakthroughs during 1984–1989, (5) consensus building and convergence during 1990–1992, and finally (6) the two‐step realization of a national system during 1993–2000. Particular importance is placed on developments during the period between 1980 and 1993 that culminated in the adoption of a charter for the Council of the National Seismic System (CNSS)—the foundation for the later ANSS. Central to this story is how many individuals worked together toward a common goal of a more rational and sustainable approach to national earthquake monitoring in the United States. The review ends with the emergence of ANSS during 1999 and 2000 and its statutory authorization by Congress in November 2000.

  10. Evaluation of seismic characteristics and structural integrity for the cabinet of HANARO seismic monitoring analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Jeong Soo; Yoon, Doo Byung

    2003-06-01

    The HANARO SMAS(Seismic Monitoring Analysis System) is classified as Non-Nuclear Safety(NNS), seismic category I, and quality class T. It is required that this system can perform required functions, which are to preserve its structural integrity during and after an OBE or SSE. In this work, the structural integrity and seismic characteristics of the cabinet of the newly developed SMAS have been estimated. The most parts of the cabinet are identically designed with those of Yonggwhang and Gori Nuclear Power Plants(NPPs), unit 1 that successfully completed the required seismic qualification tests. The structure of the cabinet of the SMAS is manufactured by the manufacturer of the cabinet of Yonggwhang and Gori NPPs. To evaluate the seismic characteristics of the SMAS, the RRS(Required Response Spectra) of the newly developed cabinet are compared with those of Yonggwhang and Gori NPPs, unit 1. In addition, natural frequencies of the cabinet of HANARO, Yonggwhang, and Gori NPPs were measured for the comparison of the seismic characteristics of the installed cabinets. In case of HANARO, the bottom of the cabinet is welded to the base plate. The base plate is fixed to the concrete foundation by using anchor bolts. For the evaluation of the structural integrity of the welding parts and the anchor bolts, the maximum stresses and forces of the welding parts and the anchor bolts due to seismic loading are estimated. The analysis results show that maximum stresses and forces are less than the allowable limits. This new SMAS is operating at HANARO instrument room to acquire and analyze the signal of earthquake.

  11. Seismic attenuation system for the AEI 10 meter Prototype

    International Nuclear Information System (INIS)

    Wanner, A; Bergmann, G; Fricke, T; Lück, H; Mow-Lowry, C M; Strain, K A; Goßler, S; Danzmann, K; Bertolini, A

    2012-01-01

    Isolation from seismic motion is vital for vibration sensitive experiments. The seismic attenuation system (SAS) is a passive mechanical isolation system for optics suspensions. The low natural frequency of a SAS allows seismic isolation starting below 0.2 Hz. The desired isolation at frequencies above a few hertz is 70–80 dB in both horizontal and vertical degrees of freedom. An introduction to the SAS for the AEI 10 m Prototype, an overview of the mechanical design and a description of the major components are given. (paper)

  12. Seismic proving test of BWR primary loop recirculation system

    International Nuclear Information System (INIS)

    Sato, H.; Shigeta, M.; Karasawa, Y.

    1987-01-01

    The seismic proving test of BWR Primary Loop Recirculation system is the second test to use the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory. The purpose of this test is to prove the seismic reliability of the primary loop recirculation system (PLR), one of the most important safety components in the BWR nuclear plants, and also to confirm the adequacy of seismic analysis method used in the current seismic design. To achieve the purpose, the test was conducted under conditions and scale as near as possible to actual systems. The strength proving test was carried out with the test model mounted on the vibration table in consideration of basic design earthquake ground motions and other conditions to confirm the soundness of structure and the strength against earthquakes. Detailed analysis and analytic evaluation of the data obtained from the test was conducted to confirm the adequacy of the seismic analysis method and earthquake response analysis method used in the current seismic design. Then, on the basis of the results obtained, the seismic safety and reliability of BWR primary loop recirculation of the actual plants was fully evaluated

  13. 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)

  14. 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)

  15. Seismic alarm system for Ignalina nuclear power plant

    International Nuclear Information System (INIS)

    Wieland, M.; Griesser, L.; Austin, G.E.; Tiurin, S.; Kuendig, C.

    2001-01-01

    A seismic alarm system will be installed at the Ignalina Nuclear Power Plant (INPP) in Lithuania. There are two reactors, both RMBK 1500 MW units. Each reactor is a water cooled, graphite moderated, channel type reactor. INPP has the most advanced version of the RMBK reactor design series. The first and second units of INPP went into service at the end of 1983 and in August 1987 respectively. Their design lifetime is approx. 30 years. The various buildings and plant have been designed for two earthquake levels, that is the design earthquake and the maximum possible earthquake with peak ground accelerations ranging from 1.2% to 10% of the acceleration due to gravity. Certain parts of the buildings and some of the equipment of the first and second units do not comply with Western seismic standards. As seismic strengthening of the existing buildings and equipment is not feasible economically, a reactor protection system based on an earthquake early warning system was recommended. This system essentially consists of six seismic stations encircling INPP at a radial distance of approx. 30 km and a seventh station at INPP. Each station includes three seismic substations each 500 m apart. The ground motion at each station is measured continuously by three accelerometers and one seismometer. Data is transmitted via telemetry to the control centre at INPP. Early warning alarms are generated if a seismic threshold is exceeded. This paper discusses the characteristics of INPP, the seismic alarm system presently under construction and the experience with other early warning and seismic alarm systems. (author)

  16. 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.)

  17. Mechanical System Simulations for Seismic Signature Modeling

    National Research Council Canada - National Science Library

    Lacombe, J

    2001-01-01

    .... Results for an M1A1 and T72 are discussed. By analyzing the simulated seismic signature data in conjunction with the spectral features associated with the vibrations of specific vehicle sprung and un-sprung components we are able to make...

  18. Control of mixed seismic isolation systems

    International Nuclear Information System (INIS)

    Teodorescu, Catalin-Stefan

    2013-01-01

    Vibration attenuation control designs are proposed for reduced plant models consisting of n-degree-of-freedom base seismically-isolated structures (i.e., a specific type of earthquake-resistant design), modeled by uncertain nonlinear systems and subjected to one-dimensional horizontal ground acceleration (i.e. the earthquake signal), treated as unknown disturbance but assumed to be bounded. In control systems literature, this is a perturbation attenuation problem. The main result of this PhD is the development of a modified version of Leitmann and co-authors' classical result on the stabilization of uncertain nonlinear systems. The proposed theorem consists of a bounded nonlinear feedback control law that is capable of ensuring uniform boundedness and uniform ultimate boundedness in closed-loop. In particular, it can be applied to solving semi-active control design problems, which are currently dealt with in earthquake engineering. The control objective is to improve the behavior (i.e. response) of mixed base-isolated structures to external disturbance, namely earthquakes. What differentiates our problem from the majority to be found in the literature is that: (i) attention is being paid to the protection of equipment placed inside the structure an not only to the structure itself; (ii) instead of using regular performance indicators expressed in terms of relative base displacement versus floors accelerations, we use solely the pseudo-acceleration floor response spectra, as it was proposed in previous recent works by Politopoulos and Pham. Actually, this work is an attempt to explicitly use floor response spectra as performance criterion. Concerning the application procedure, some of the topics that were detailed are: (i) modeling of earthquake signals; (ii) tuning of control law parameters based on vibration theory; (iii) validation and testing of the closed-loop behavior using numerical simulations: for simplicity reasons, we take n=2. This procedure can be

  19. Seismic qualification of SPX1 shutdown systems - tests and calculations

    International Nuclear Information System (INIS)

    Brochard, D.; Buland, P.

    1988-01-01

    The SUPERPHENIX 1 shutdown system is composed of two main systems: the Complementary Shutdown System SAC (Systeme d'Arret Complementaire) and the Primary Shutdown System (SCP) (Systeme de Commande Principal). In case of a seismic event, the insertability of the different shutdown systems has to be demonstrated. Tests have been performed on the SAC and have shown that this system was not sensitive to the seismic excitation (the drop time increases of 10% at SSE level). For the SCP, as an analytical demonstration was felt difficult to achieve, it was decided to perform a full scale testing program. These tests have been performed for the two types of SCP which are present in Superphenix: SCP 1 (Creusot Loire design), SCP 2 (Novatome design). As there was no existing facility in France to test this kind of slender structure (21 metres high) a new facility named VESUBIE was designed and installed in an existing pit located at the Saclay nuclear research center. The objectives of the tests were the following: to demonstrate insertability of control rod; to demonstrate absence of seismic induced damage to the SCP; to measure increase of scram time; to measure seismic induced stresses; to obtain data for code correlation. After completion of the tests, measurements have been correlated with results obtained from a non-linear finite element model. Time history correlations were achieved for SCP 1. Afterwards a calculation was performed in hot condition to find if there was some effect of temperature on SCP seismic response. 2 refs, 8 figs

  20. Seismic analysis of mechanical systems at Pickering NGS

    International Nuclear Information System (INIS)

    Ghobarah, A.

    1995-11-01

    The objective of this study is to assess the seismic withstand capacity of selected safety-related mechanical systems associated with the Pressure Relief Duct (PRD) at the Pickering A Nuclear Generating Station. These systems are attached to the PRD and include the Emergency Coolant Injection System piping, the Vacuum Ducts, the Emergency Water Storage System, the PRD expansion joint seals and the PRD to Reactor Building joint seals. The input support motion to the mechanical systems is taken to be the seismic response of the PRD determined in an earlier study using various levels of predetermined ground response spectrum envelope. (author). 12 refs., 13 tabs., 48 figs

  1. A report on seismic re-evaluation of Cirus systems

    International Nuclear Information System (INIS)

    Varma, Veto; Reddy, G.R.; Vaze, K.K.; Kushwaha, H.S.

    2003-06-01

    Cirus was initiated way back in 1955 and its design was made with the methods prevailing at that time. The design codes and safety standards have changed since then, particularly with respect to seismic design criteria. As the structure is an important safety related structure it is mandatory to meet the present statutory requirement. This report contains the seismic qualification for some of the Cirus systems. The report has four parts. Part I gives the analytical studies performed in the containment building, Part II describes of experimental studies carried out to validate the analytical studies for containment builaing, Part III explains the seismic retrofitting of Battery bank, and Part IV summarizes the seismic qualification of inlet and exhaust damper of Cirus. (author)

  2. Seismic evaluation of piping systems using screening criteria

    International Nuclear Information System (INIS)

    Campbell, R.D.; Landers, D.F.; Minichiello, J.C.; Slagis, G.C.; Antaki, G.A.

    1994-01-01

    This document may be used by a qualified review team to identify potential sources of seismically induced failure in a piping system. Failure refers to the inability of a piping system to perform its expected function following an earthquake, as defined in Table 1. The screens may be used alone or with the Seismic Qualification Utility Group -- Generic Implementation Procedure (SQUG-GIP), depending on the piping system's required function, listed in Table 1. Features of a piping system which do not the screening criteria are called outliers. Outliers must either be resolved through further evaluations, or be considered a potential source of seismically induced failure. Outlier evaluations, which do not necessarily require the qualification of a complete piping system by stress analysis, may be based on one or more of the following: simple calculations of pipe spans, search of the test or experience data, vendor data, industry practice, etc

  3. 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.

  4. 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

  5. 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

  6. Seismic analysis response factors and design margins of piping systems

    International Nuclear Information System (INIS)

    Shieh, L.C.; Tsai, N.C.; Yang, M.S.; Wong, W.L.

    1985-01-01

    The objective of the simplified methods project of the Seismic Safety Margins Research Program is to develop a simplified seismic risk methodology for general use. The goal is to reduce seismic PRA costs to roughly 60 man-months over a 6 to 8 month period, without compromising the quality of the product. To achieve the goal, it is necessary to simplify the calculational procedure of the seismic response. The response factor approach serves this purpose. The response factor relates the median level response to the design data. Through a literature survey, we identified the various seismic analysis methods adopted in the U.S. nuclear industry for the piping system. A series of seismic response calculations was performed. The response factors and their variabilities for each method of analysis were computed. A sensitivity study of the effect of piping damping, in-structure response spectra envelop method, and analysis method was conducted. In addition, design margins, which relate the best-estimate response to the design data, are also presented

  7. A SEISMIC DESIGN OF NUCLEAR REACTOR BUILDING STRUCTURES APPLYING SEISMIC ISOLATION SYSTEM IN A HIGH SEISMICITY REGION –A FEASIBILITY CASE STUDY IN JAPAN-

    Directory of Open Access Journals (Sweden)

    TETSUO KUBO

    2014-10-01

    Full Text Available A feasibility study on the seismic design of nuclear reactor buildings with application of a seismic isolation system is introduced. After the Hyogo-ken Nanbu earthquake in Japan of 1995, seismic isolation technologies have been widely employed for commercial buildings. Having become a mature technology, seismic isolation systems can be applied to NPP facilities in areas of high seismicity. Two reactor buildings are discussed, representing the PWR and BWR buildings in Japan, and the application of seismic isolation systems is discussed. The isolation system employing rubber bearings with a lead plug positioned (LRB is examined. Through a series of seismic response analyses using the so-named standard design earthquake motions covering the design basis earthquake motions obtained for NPP sites in Japan, the responses of the seismic isolated reactor buildings are evaluated. It is revealed that for the building structures examined herein: (1 the responses of both isolated buildings and isolating LRBs fulfill the specified design criteria; (2 the responses obtained for the isolating LRBs first reach the ultimate condition when intensity of motion is 2.0 to 2.5 times as large as that of the design-basis; and (3 the responses of isolated reactor building fall below the range of the prescribed criteria.

  8. A seismic design of nuclear reactor building structures applying seismic isolation system in a seismicity region-a feasibility case study in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kubo, Tetsuo [The University of Tokyo, Tokyo (Japan); Yamamoto, Tomofumi; Sato, Kunihiko [Mitsubishi Heavy Industries, Ltd., Kobe (Japan); Jimbo, Masakazu [Toshiba Corporation, Yokohama (Japan); Imaoka, Tetsuo [Hitachi-GE Nuclear Energy, Ltd., Hitachi (Japan); Umeki, Yoshito [Chubu Electric Power Co. Inc., Nagoya (Japan)

    2014-10-15

    A feasibility study on the seismic design of nuclear reactor buildings with application of a seismic isolation system is introduced. After the Hyogo-ken Nanbu earthquake in Japan of 1995, seismic isolation technologies have been widely employed for commercial buildings. Having become a mature technology, seismic isolation systems can be applied to NPP facilities in areas of high seismicity. Two reactor buildings are discussed, representing the PWR and BWR buildings in Japan, and the application of seismic isolation systems is discussed. The isolation system employing rubber bearings with a lead plug positioned (LRB) is examined. Through a series of seismic response analyses using the so-named standard design earthquake motions covering the design basis earthquake motions obtained for NPP sites in Japan, the responses of the seismic isolated reactor buildings are evaluated. It is revealed that for the building structures examined herein: (1) the responses of both isolated buildings and isolating LRBs fulfill the specified design criteria; (2) the responses obtained for the isolating LRBs first reach the ultimate condition when intensity of motion is 2.0 to 2.5 times as large as that of the design-basis; and (3) the responses of isolated reactor building fall below the range of the prescribed criteria.

  9. Seismic Fracture Characterization Methodologies for Enhanced Geothermal Systems

    Energy Technology Data Exchange (ETDEWEB)

    Queen, John H. [Hi-Geophysical, Inc., Ponca, OK (United States)

    2016-05-09

    Executive Summary The overall objective of this work was the development of surface and borehole seismic methodologies using both compressional and shear waves for characterizing faults and fractures in Enhanced Geothermal Systems. We used both surface seismic and vertical seismic profile (VSP) methods. We adapted these methods to the unique conditions encountered in Enhanced Geothermal Systems (EGS) creation. These conditions include geological environments with volcanic cover, highly altered rocks, severe structure, extreme near surface velocity contrasts and lack of distinct velocity contrasts at depth. One of the objectives was the development of methods for identifying more appropriate seismic acquisition parameters for overcoming problems associated with these geological factors. Because temperatures up to 300º C are often encountered in these systems, another objective was the testing of VSP borehole tools capable of operating at depths in excess of 1,000 m and at temperatures in excess of 200º C. A final objective was the development of new processing and interpretation techniques based on scattering and time-frequency analysis, as well as the application of modern seismic migration imaging algorithms to seismic data acquired over geothermal areas. The use of surface seismic reflection data at Brady's Hot Springs was found useful in building a geological model, but only when combined with other extensive geological and geophysical data. The use of fine source and geophone spacing was critical in producing useful images. The surface seismic reflection data gave no information about the internal structure (extent, thickness and filling) of faults and fractures, and modeling suggests that they are unlikely to do so. Time-frequency analysis was applied to these data, but was not found to be significantly useful in their interpretation. Modeling does indicate that VSP and other seismic methods with sensors located at depth in wells will be the most

  10. Improved Seismic Acquisition System and Data Processing for the Italian National Seismic Network

    Science.gov (United States)

    Badiali, L.; Marcocci, C.; Mele, F.; Piscini, A.

    2001-12-01

    A new system for acquiring and processing digital signals has been developed in the last few years at the Istituto Nazionale di Geofisica e Vulcanologia (INGV). The system makes extensive use of the internet communication protocol standards such as TCP and UDP which are used as the transport highway inside the Italian network, and possibly in a near future outside, to share or redirect data among processes. The Italian National Seismic Network has been working for about 18 years equipped with vertical short period seismometers and transmitting through analog lines, to the computer center in Rome. We are now concentrating our efforts on speeding the migration towards a fully digital network based on about 150 stations equipped with either broad band or 5 seconds sensors connected to the data center partly through wired digital communication and partly through satellite digital communication. The overall process is layered through intranet and/or internet. Every layer gathers data in a simple format and provides data in a processed format, ready to be distributed towards the next layer. The lowest level acquires seismic data (raw waveforms) coming from the remote stations. It handshakes, checks and sends data in LAN or WAN according to a distribution list where other machines with their programs are waiting for. At the next level there are the picking procedures, or "pickers", on a per instrument basis, looking for phases. A picker spreads phases, again through the LAN or WAN and according to a distribution list, to one or more waiting locating machines tuned to generate a seismic event. The event locating procedure itself, the higher level in this stack, can exchange information with other similar procedures. Such a layered and distributed structure with nearby targets allows other seismic networks to join the processing and data collection of the same ongoing event, creating a virtual network larger than the original one. At present we plan to cooperate with other

  11. A seismic monitoring system for response and failure of structures with intentionally reduced seismic strength

    International Nuclear Information System (INIS)

    Takanashi, Koichi; Ohi, Kenichi

    1988-01-01

    A group of steel and reinforced concrete scaled structures with intentionally reduced seismic strength to 1/3 to 1/2 were constructed in 1983 for long term observation in order to collect precise data of earthquake response and grasp failure mechanisms during earthquakes. A monitoring system was installed in the structures as well as in the surrounding soil. Some reliable data have been successfully recorded since then, which can be available for verification of analytical models. (author)

  12. 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.)

  13. Seismic evaluation of safety systems at the Savannah River reactors

    International Nuclear Information System (INIS)

    Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.M.; Ketcham, D.R.

    1989-01-01

    A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table testing which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its ''Generic Safety Evaluation Report'' approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the United States and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluating program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

  14. Seismic evaluation of safety systems at the Savannah River reactors

    International Nuclear Information System (INIS)

    Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.; Ketcham, D.

    1989-01-01

    A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table tested which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its Generic Safety Evaluation Report approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the US and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective approach developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluation program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

  15. Alternate seismic support for pipeline systems in nuclear power plants

    International Nuclear Information System (INIS)

    Muthumani, K.; Gopalakrishnan, N.; Sathish Kumar, K.; Sreekala, R.; Rama Rao, G.V.; Reddy, G.R.; Parulekar, Y.M.

    2008-01-01

    Failure free design of supporting systems for pipe lines carrying highly toxic or radioactive liquids at very high temperature is an important issue in the safety aspect for a nuclear power plant installation which is a key topic for researchers all around the world. Generally, these pipeline systems are designed to be held rigid by conventional snubber supports for protection from earthquakes. The piping design must balance seismic deformations and other deformations due to thermal effect. A rigid pipeline system using conventional snubber supports always leads to an increase in thermal stresses; hence a rational seismic design for pipeline supporting systems becomes essential. Contrary to this rigid design, it is possible to design a flexible pipeline system and to decrease the seismic response by increasing the damping through the use of passive energy absorbing elements, which dissipate vibration energy. This paper presents the experimental and analytical studies carried out on modeling yielding type elasto-plastic passive energy-absorbing elements to be used in a passive energy-dissipating device for the control of large seismic deformations of pipelines subjected to earthquake loading. (author)

  16. Simulation-based seismic loss estimation of seaport transportation system

    International Nuclear Information System (INIS)

    Ung Jin Na; Shinozuka, Masanobu

    2009-01-01

    Seaport transportation system is one of the major lifeline systems in modern society and its reliable operation is crucial for the well-being of the public. However, past experiences showed that earthquake damage to port components can severely disrupt terminal operation, and thus negatively impact on the regional economy. The main purpose of this study is to provide a methodology for estimating the effects of the earthquake on the performance of the operation system of a container terminal in seaports. To evaluate the economic loss of damaged system, an analytical framework is developed by integrating simulation models for terminal operation and fragility curves of port components in the context of seismic risk analysis. For this purpose, computerized simulation model is developed and verified with actual terminal operation records. Based on the analytical procedure to assess the seismic performance of the terminal, system fragility curves are also developed. This simulation-based loss estimation methodology can be used not only for estimating the seismically induced revenue loss but also serve as a decision-making tool to select specific seismic retrofit technique on the basis of benefit-cost analysis

  17. Demonstration of NonLinear Seismic Soil Structure Interaction and Applicability to New System Fragility Seismic Curves

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology

    2014-09-01

    Risk calculations should focus on providing best estimate results, and associated insights, for evaluation and decision-making. Specifically, seismic probabilistic risk assessments (SPRAs) are intended to provide best estimates of the various combinations of structural and equipment failures that can lead to a seismic induced core damage event. However, in general this approach has been conservative, and potentially masks other important events (for instance, it was not the seismic motions that caused the Fukushima core melt events, but the tsunami ingress into the facility). SPRAs are performed by convolving the seismic hazard (the frequency of certain magnitude events) with the seismic fragility (the conditional probability of failure of a structure, system, or component given the occurrence of earthquake ground motion). In this calculation, there are three main pieces to seismic risk quantification, 1) seismic hazard and nuclear power plants (NPPs) response to the hazard, fragility or capacity of structures, systems and components (SSC), and systems analysis. Figure 1 provides a high level overview of the risk quantification process. The focus of this research is on understanding and removing conservatism (when possible) in the quantification of seismic risk at NPPs.

  18. Seismic analysis of liquid metal reactor piping systems

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1987-01-01

    This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis. (orig./GL)

  19. 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

  20. FSI analysis of piping systems under seismic excitation

    International Nuclear Information System (INIS)

    Uras, R.A.; Ma, D.C.; Chang, Yao W.; Liu, Wing Kam

    1991-01-01

    A formulation which accounts for fluid-structure interaction of piping system under seismic excitation is presented. The governing equations of the fluid and the structure to model the pipe are stated. Using the finite element method the discretized equations are obtained. A transformation procedure for proper assembly of matrices is introduced. A solution algorithm is described. 9 refs., 2 figs

  1. 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

  2. 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

  3. 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

  4. Integrated seismic design of structure and control systems

    CERN Document Server

    Castaldo, Paolo

    2014-01-01

    The structural optimization procedure presented in this book makes it possible to achieve seismic protection through integrated structural/control system design. In particular, it is explained how slender structural systems with a high seismic performance can be achieved through inclusion of viscous and viscoelastic dampers as an integral part of the system. Readers are provided with essential introductory information on passive structural control and passive energy dissipation systems. Dynamic analyses of both single and multiple degree of freedom systems are performed in order to verify the achievement of pre-assigned performance targets, and it is explained how the optimal integrated design methodology, also relevant to retrofitting of existing buildings, should be applied. The book illustrates how structural control research is opening up new possibilities in structural forms and configurations without compromising structural performance.

  5. Seismic safety margins research program. Project I SONGS 1 AFWS Project

    International Nuclear Information System (INIS)

    Chuang, T.Y.; Smith, P.D.; Dong, R.G.; Bernreuter, D.L.; Bohn, M.P.; Cummings, G.E.; Wells, J.E.

    1981-01-01

    The seismic qualification requirements of auxiliary feedwater systems (AFWS) of Pressurized Water Reactors (PWR) were developed over a number of years. These are formalized in the publication General Design Criteria (Appendix A to 10CFR50). The full recognition of the system as an engineered safety feature did not occur until publication of the Standard Review Plan (1975). Efforts to determine how to backfit seismic requirements to earlier plants has been undertaken primarily in the Systematic Evaluation Program (SEP) for a limited number of operating reactors. Nuclear Reactor Research (RES) and NRR have requested LLNL to perform a probabilistic study on the AFWS of San Onofre Nuclear Generating Station (SONGS) Unit 1 utilizing the tools developed by the Seismic Safety Margins Research Program (SSMRP). The main objectives of this project are to: identify the weak links of AFWS; compare the failure probabilities of SONGS 1 and Zion 1 AFWS: and compare the seismic responses due to different input spectra and design values

  6. A new system for seismic yield estimation of underground explosions

    International Nuclear Information System (INIS)

    Murphy, J.R.

    1991-01-01

    Research conducted over the past decade has led to the development of a number of innovative procedures for estimating the yields of underground nuclear explosions based on systematic analyses of digital seismic data recorded from these tests. In addition, a wide variety of new data regarding the geophysical environments at Soviet test locations have now become available as a result of the Joint Verification Experiment (JVE) and associated data exchanges. The system described in this paper represents an attempt to integrate all these new capabilities and data into a comprehensive operational prototype which can be used to obtain optimum seismic estimates of explosion yield together with quantitative measures of the uncertainty in those estimates. The implementation of this system has involved a wide variety of technical tasks, including the development of a comprehensive seismic database and related database access software, formulation of a graphical test site information interface for accessing available information on explosion source conditions, design of an interactive seismic analyst station for use in processing the observed data to extract the required magnitude measures and the incorporation of formal statistical analysis modules for use in yield estimation and assessment

  7. 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

  8. 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

  9. A Study of Seismic Capacity of Nuclear Equipment with Seismic Isolation System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Kyu; Choun, Young Sun; Choi, In Kil; Seo, Jeong Moon

    2004-05-15

    In this study, the base isolation systems for equipment are presented and the responses of each isolation system are investigated. As for the base isolation systems, a natural rubber bearing (NRB), a high damping rubber bearing (HDRB) and a friction pendulum system (FPS) are selected. The shaking table tests are carried out for three kinds of structural types. As input motions, artificial time histories enveloping the US NRC RG 1.60 spectrum and the probability-based scenario earthquake spectra developed for the Korean nuclear power plant site as well as a typical near-fault earthquake record are used. Uniaxial, biaxial, and triaxial excitations are conducted with PGAs of 0.05, 0.1, 0.2 and 0.25g. Acceleration responses are measured at the top of the equipment model and the floors using an accelerometer. The reduction of the seismic forces transmitted to the equipment models are determined for different isolation systems and input motions.

  10. A Study of Seismic Capacity of Nuclear Equipment with Seismic Isolation System

    International Nuclear Information System (INIS)

    Kim, Min Kyu; Choun, Young Sun; Choi, In Kil; Seo, Jeong Moon

    2004-05-01

    In this study, the base isolation systems for equipment are presented and the responses of each isolation system are investigated. As for the base isolation systems, a natural rubber bearing (NRB), a high damping rubber bearing (HDRB) and a friction pendulum system (FPS) are selected. The shaking table tests are carried out for three kinds of structural types. As input motions, artificial time histories enveloping the US NRC RG 1.60 spectrum and the probability-based scenario earthquake spectra developed for the Korean nuclear power plant site as well as a typical near-fault earthquake record are used. Uniaxial, biaxial, and triaxial excitations are conducted with PGAs of 0.05, 0.1, 0.2 and 0.25g. Acceleration responses are measured at the top of the equipment model and the floors using an accelerometer. The reduction of the seismic forces transmitted to the equipment models are determined for different isolation systems and input motions

  11. Seismic assessment of air-cooled type emergency electric power supply system

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    JNES initiated seismic assessment programs to develop seismic review criterions for the air-cooled system (diesel generator, gas turbine generator), which will be newly installed for enhancing the diversity of emergency electric power supply system. Five principal subjects are involved in the programs: two subjects for fiscal 2011 and three ones for fiscal 2012 and 2013. The summary of outcomes is as follows: 1) Past capacity test data and related technical issues (2011). Seismic capacity data obtained from past seismic shaking tests were investigated. 2) Test programs based on the investigation of system specification (2011). Design specifications for the air-cooled system were investigated. 3) Large Air Fin Cooler (AFC) one unit model seismic capacity test and quantitative seismic capacity evaluation. AFC one unit model seismic capacity tests were conducted and quantitative seismic capacities were evaluated. (author)

  12. Seismic assessment of air-cooled type emergency electric power supply system

    International Nuclear Information System (INIS)

    2013-01-01

    JNES initiated seismic assessment programs to develop seismic review criterions for the air-cooled system (diesel generator, gas turbine generator), which will be newly installed for enhancing the diversity of emergency electric power supply system. Five principal subjects are involved in the programs: two subjects for fiscal 2011 and three ones for fiscal 2012 and 2013. The summary of outcomes is as follows: 1) Past capacity test data and related technical issues (2011). Seismic capacity data obtained from past seismic shaking tests were investigated. 2) Test programs based on the investigation of system specification (2011). Design specifications for the air-cooled system were investigated. 3) Large Air Fin Cooler (AFC) one unit model seismic capacity test and quantitative seismic capacity evaluation. AFC one unit model seismic capacity tests were conducted and quantitative seismic capacities were evaluated. (author)

  13. Seismic Design of ITER Component Cooling Water System-1 Piping

    Science.gov (United States)

    Singh, Aditya P.; Jadhav, Mahesh; Sharma, Lalit K.; Gupta, Dinesh K.; Patel, Nirav; Ranjan, Rakesh; Gohil, Guman; Patel, Hiren; Dangi, Jinendra; Kumar, Mohit; Kumar, A. G. A.

    2017-04-01

    The successful performance of ITER machine very much depends upon the effective removal of heat from the in-vessel components and other auxiliary systems during Tokamak operation. This objective will be accomplished by the design of an effective Cooling Water System (CWS). The optimized piping layout design is an important element in CWS design and is one of the major design challenges owing to the factors of large thermal expansion and seismic accelerations; considering safety, accessibility and maintainability aspects. An important sub-system of ITER CWS, Component Cooling Water System-1 (CCWS-1) has very large diameter of pipes up to DN1600 with many intersections to fulfill the process flow requirements of clients for heat removal. Pipe intersection is the weakest link in the layout due to high stress intensification factor. CCWS-1 piping up to secondary confinement isolation valves as well as in-between these isolation valves need to survive a Seismic Level-2 (SL-2) earthquake during the Tokamak operation period to ensure structural stability of the system in the Safe Shutdown Earthquake (SSE) event. This paper presents the design, qualification and optimization of layout of ITER CCWS-1 loop to withstand SSE event combined with sustained and thermal loads as per the load combinations defined by ITER and allowable limits as per ASME B31.3, This paper also highlights the Modal and Response Spectrum Analyses done to find out the natural frequency and system behavior during the seismic event.

  14. Seismic isolation systems designed with distinct multiple frequencies

    International Nuclear Information System (INIS)

    Wu, Ting-shu; Seidensticker, R.W.

    1991-01-01

    Two systems for seismic base isolation are presented. The main feature of these system is that, instead of only one isolation frequency as in conventional isolation systems, they are designed to have two distinct isolation frequencies. When the responses during an earthquake exceed the design value(s), the system will automatically and passively shift to the secondly isolation frequency. Responses of these two systems to different ground motions including a harmonic motion with frequency same as the primary isolation frequency, show that no excessive amplification will occur. Adoption of these new systems certainly will greatly enhance the safety and reliability of an isolated superstructure against future strong earthquakes. 3 refs

  15. Anatomy of the TAMA SAS seismic attenuation system

    International Nuclear Information System (INIS)

    Marka, Szabolcs; Takamori, Akiteru; Ando, Masaki; Bertolini, Alessandro; Cella, Giancarlo; DeSalvo, Riccardo; Fukushima, Mitsuhiro; Iida, Yukiyoshi; Jacquier, Florian; Kawamura, Seiji; Nishi, Yuhiko; Numata, Kenji; Sannibale, Virginio; Somiya, Kentaro; Takahashi, Ryutaro; Tariq, Hareem; Tsubono, Kimio; Ugas, Jose; Viboud, Nicolas; Wang Chenyang; Yamamoto, Hiroaki; Yoda, Tatsuo

    2002-01-01

    The TAMA SAS seismic attenuation system was developed to provide the extremely high level of seismic isolation required by the next generation of interferometric gravitational wave detectors to achieve the desired sensitivity at low frequencies. Our aim was to provide good performance at frequencies above ∼10 Hz, while utilizing only passive subsystems in the sensitive frequency band of the TAMA interferometric gravitational wave detectors. The only active feedback is relegated below 6 Hz and it is used to damp the rigid body resonances of the attenuation chain. Simulations, based on subsystem performance characterizations, indicate that the system can achieve rms mirror residual motion measured in a few tens of nanometres. We will give a brief overview of the subsystems and point out some of the characterization results, supporting our claims of achieved performance. SAS is a passive, UHV compatible and low cost system. It is likely that extremely sensitive experiments in other fields will also profit from our study

  16. Alternative methods for the seismic analysis of piping systems

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    This document is a review of 12 methods and criteria for the seismic analysis of piping systems. Each of the twelve chapters in this document cover the important technical aspects of a given method. The technical aspects presented are those the Subcommittee on Dynamic Stress Criteria believe important to the application of the method, and should not be considered as a positive or negative endorsement for any of the methods. There are many variables in an analysis of a piping system that can influence the selection of the analysis method and criteria to be applied. These variable include system configuration, technical issues, precedent, licensing considerations, and regulatory acceptance. They must all be considered in selecting the appropriate seismic analysis method and criteria. This is relevant for nuclear power plants

  17. 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)

  18. 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

  19. 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

  20. 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

  1. Identifying Conventionally Sub-Seismic Faults in Polygonal Fault Systems

    Science.gov (United States)

    Fry, C.; Dix, J.

    2017-12-01

    Polygonal Fault Systems (PFS) are prevalent in hydrocarbon basins globally and represent potential fluid pathways. However the characterization of these pathways is subject to the limitations of conventional 3D seismic imaging; only capable of resolving features on a decametre scale horizontally and metres scale vertically. While outcrop and core examples can identify smaller features, they are limited by the extent of the exposures. The disparity between these scales can allow for smaller faults to be lost in a resolution gap which could mean potential pathways are left unseen. Here the focus is upon PFS from within the London Clay, a common bedrock that is tunnelled into and bears construction foundations for much of London. It is a continuation of the Ieper Clay where PFS were first identified and is found to approach the seafloor within the Outer Thames Estuary. This allows for the direct analysis of PFS surface expressions, via the use of high resolution 1m bathymetric imaging in combination with high resolution seismic imaging. Through use of these datasets surface expressions of over 1500 faults within the London Clay have been identified, with the smallest fault measuring 12m and the largest at 612m in length. The displacements over these faults established from both bathymetric and seismic imaging ranges from 30cm to a couple of metres, scales that would typically be sub-seismic for conventional basin seismic imaging. The orientations and dimensions of the faults within this network have been directly compared to 3D seismic data of the Ieper Clay from the offshore Dutch sector where it exists approximately 1km below the seafloor. These have typical PFS attributes with lengths of hundreds of metres to kilometres and throws of tens of metres, a magnitude larger than those identified in the Outer Thames Estuary. The similar orientations and polygonal patterns within both locations indicates that the smaller faults exist within typical PFS structure but are

  2. Progress of R and D on seismic emergency information system

    International Nuclear Information System (INIS)

    2000-09-01

    After the Great Hansin-Awaji Earthquake Disaster occurred in 1995, the Science and Technology Agency commenced 'Frontier Research Program on Earthquake' in FY1996. As a part of this research program, four-year program on 'Research on Real-time Earthquake Information Transmission' has been carried out at JAERI since FY1997. Through the experience of the above earthquake disaster, the importance of accurate and prompt seismic information transmission immediately after the occurrence of the earthquake has been recognized from the viewpoint of disaster mitigation. Under this circumstance, the main activity in Real-time Earthquake Information Transmission Research at JAERI has been placed on the development of a seismic emergency information system. In order to respond to the above R and D, Seismic Emergency Information System Research Team was organized in JAERI in FY1998. In the meantime a part of this R and D program is performed under the coordinated research between JAERI and NIED(National Research Institute for Science and Disaster Prevention). This report describes the recent progress of R and D until FY1999. In the R and D, estimation techniques of hypocenter, fault and earthquake motion parameters, in which the latest results in the field of earthquake engineering were involved, were developed. Until the end of FY1999, the main part of the system, in which the above estimation techniques are introduced, is completed. By this system the seismic information is being transmitted using E-mail and homepage through the inter-net. In addition the databases on the estimated earthquake motion parameter distribution under scenario earthquakes and the surface soil amplification function around JAERI-Tokai site are prepared to examine the applicability of the system. (author)

  3. A Methodology for Assessing the Seismic Vulnerability of Highway Systems

    International Nuclear Information System (INIS)

    Cirianni, Francis; Leonardi, Giovanni; Scopelliti, Francesco

    2008-01-01

    Modern society is totally dependent on a complex and articulated infrastructure network of vital importance for the existence of the urban settlements scattered on the territory. On these infrastructure systems, usually indicated with the term lifelines, are entrusted numerous services and indispensable functions of the normal urban and human activity.The systems of the lifelines represent an essential element in all the urbanised areas which are subject to seismic risk. It is important that, in these zones, they are planned according to opportune criteria based on two fundamental assumptions: a) determination of the best territorial localization, avoiding, within limits, the places of higher dangerousness; b) application of constructive technologies finalized to the reduction of the vulnerability.Therefore it is indispensable that in any modern process of seismic risk assessment the study of the networks is taken in the rightful consideration, to be integrated with the traditional analyses of the buildings.The present paper moves in this direction, dedicating particular attention to one kind of lifeline: the highway system, proposing a methodology of analysis finalized to the assessment of the seismic vulnerability of the system

  4. Real-time Seismic Alert System of NIED

    Science.gov (United States)

    Horiuchi, S.; Fujinawa, Y.; Negishi, H.; Matsumoto, T.; Fujiwara, H.; Kunugi, T.; Hayashi, Y.

    2001-12-01

    An extensive seismic network has been constructed nationwide composed of hi-sensitivity seismographic network, broadband seismographic network and strong motion seismographic network. All these data from some 3,000 sites belonging to NIED, JMA and universities are to be accumulated and distributed through NIED to any scientists and engineering through INTERNET under the coordination of the National Seismic Research Committee of MEXT. As a practical application of those data we are now developing a real-time seismic alert information system for the purpose of providing short-term warning of imminent strong grounds motions from major earthquakes from several seconds to a few days. The contents of information are seismic focal parameters (several seconds), seismic fault plane solutions (some 10 seconds), after-shock activities (several minutes-a few days ). The fundamental fault parameters are used to build specific information at sites for particular users for use of triggering automated and /or half-automated responses. The most important application is an immediate estimate of expected shaking distribution and damages in a district using synthetic database and site effects for local governments to initial proper measures of hazard mitigation. Another application is estimation of arrival time and shaking strength at any individual site for human lives to be safeguarded. The system could also start an automatic electrical isolation and protection of computer systems, protection of hazardous chronic systems, transportation systems and so on. The information are corrected successively as seismic ground motion are received at a larger number of sites in time with the result that more accurate and more sophisticated earthquake information is transmitted to any user. Besides the rapid determination of seismic parameters, one of essential items in this alert system is the data transmission means. The data transmission is chosen to assure negligibly small delay of data

  5. Comprehensive Final Report for the Marine Seismic System Program

    Science.gov (United States)

    1985-08-01

    serve as a principal reference for transitioning marine seismic system techniques and results from the research and development arena to the...vM . .’ .■ .» .%■■.•. - Viaj ^."-;/-.■■ *• -’•’■■’■ ■ ■ - ■ • ■ -. . -. • ^;-■:■:-:•:> •■•."--.--.v. ’-• V ’.■ *.- ".i • ■ - ■ ■ v V

  6. 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)

  7. 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

  8. Connection with seismic networks and construction of real time earthquake monitoring system

    International Nuclear Information System (INIS)

    Chi, Heon Cheol; Lee, H. I.; Shin, I. C.; Lim, I. S.; Park, J. H.; Lee, B. K.; Whee, K. H.; Cho, C. S.

    2000-12-01

    It is natural to use the nuclear power plant seismic network which have been operated by KEPRI(Korea Electric Power Research Institute) and local seismic network by KIGAM(Korea Institute of Geology, Mining and Material). The real time earthquake monitoring system is composed with monitoring module and data base module. Data base module plays role of seismic data storage and classification and the other, monitoring module represents the status of acceleration in the nuclear power plant area. This research placed the target on the first, networking the KIN's seismic monitoring system with KIGAM and KEPRI seismic network and the second, construction the KIN's Independent earthquake monitoring system

  9. Stochastic seismic floor response analysis method for various damping systems

    International Nuclear Information System (INIS)

    Kitada, Y.; Hattori, K.; Ogata, M.; Kanda, J.

    1991-01-01

    A study using the stochastic seismic response analysis method which is applicable for the estimation of floor response spectra is carried out. It is pointed out as a shortcoming in this stochastic seismic response analysis method, that the method tends to overestimate floor response spectra for low damping systems, e.g. 1% of the critical damping ratio. An investigation on the cause of the shortcoming is carried out and a number of improvements in this method were also made to the original method by taking correlation of successive peaks in a response time history into account. The application of the improved method to a typical BWR reactor building is carried out. The resultant floor response spectra are compared with those obtained by deterministic time history analysis. Floor response spectra estimated by the improved method consistently cover the response spectra obtained by the time history analysis for various damping ratios. (orig.)

  10. Seismic analysis of liquid metal reactor piping systems

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1987-01-01

    To safely assess the adequacy of the LMR piping, a three-dimensional piping code, SHAPS, has been developed at Argonne National Laboratory. This code was initially intended for calculating hydrodynamic-wave propagation in a complex piping network. It has salient features for treating fluid transients of fluid-structure interactions for piping with in-line components. The code also provides excellent structural capabilities of computing stresses arising from internal pressurization and 3-D flexural motion of the piping system. As part of the development effort, the SHAPS code has been further augmented recently by introducing the capabilities of calculating piping response subjected to seismic excitations. This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis

  11. Seismic testing and analysis of a prototypic nonlinear piping system

    International Nuclear Information System (INIS)

    Barta, D.A.; Anderson, M.J.; Severud, L.K.

    1982-11-01

    A series of seismic tests and analyses of a nonlinear Fast Flux Test Facility (FFTF) prototypic piping system are described, and measured responses are compared with analytical predictions. The test loop was representative of a typical LMFBR insulated small bore piping system and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers and pipe clamps. Various piping support configurations were tested and analyzed to evaluate the effects of free play and other nonlinear stiffness characteristics on the piping system response

  12. Seismic design criteria of fire protection systems for DOE facilities

    International Nuclear Information System (INIS)

    Hardy, G.; Cushing, R.; Driesen, G.

    1991-01-01

    Fire protection systems are critical to the safety of personnel and to the protection of inventory during any kind of emergency situation that involves a fire. The importance of these fire protection systems is hightened for DOE facilities which often house nuclear, chemical or scientific processes. Current research into the topic of open-quotes fires following earthquakesclose quotes has demonstrated that the risks of a fire starting as a result of a major earthquake can be significant. Thus, fire protection systems need to be designed to withstand the anticipated seismic event for the site in question

  13. Similarity of fluctuations in correlated systems: The case of seismicity

    International Nuclear Information System (INIS)

    Varotsos, P.A.; Sarlis, N.V.; Tanaka, H.K.; Skordas, E.S.

    2005-01-01

    We report a similarity of fluctuations in equilibrium critical phenomena and nonequilibrium systems, which is based on the concept of natural time. The worldwide seismicity as well as that of the San Andreas fault system and Japan are analyzed. An order parameter is chosen and its fluctuations relative to the standard deviation of the distribution are studied. We find that the scaled distributions fall on the same curve, which interestingly exhibits, over four orders of magnitude, features similar to those in several equilibrium critical phenomena (e.g., two-dimensional Ising model) as well as in nonequilibrium systems (e.g., three-dimensional turbulent flow)

  14. The development of the operational program for seismic monitoring system of Uljin Unit 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.R.; Heo, T.Y.; Cho, B.H. [Korea Electric Power Research Institute, Taejeon (Korea, Republic of); Kang, T.G.; Kim, H.M.; Kim, Y.S.; Oh, S.M.; Kang, Y.S. [Korea Electric Power Data Network Co., Seoul (Korea, Republic of)

    1997-12-31

    Due to aging of the imported seismic monitoring system of Uljin of t 1 and 2 units it is difficult for this system to provide enough functions needed for the security of seismic safety and the evaluation of the earthquake data from the seismic instrumentation. For this reason, it is necessary to replace the seismic monitoring system of Uljin 1 and 2 units with a new system which has the localized and upgraded hardware and corresponding software. In the part of standardization of existing seismic monitoring system, furthermore, it is necessary to develop the seismic wave analysis system which incorporate newly developed software and can real-timely analyze the seismic wave. This report is the finial product of research project ``The development of the operational program for seismic monitoring system of Uljin Unit 1 and 2`` which have been performed from June 1996 to June 1997 by KEPRI and KDN. Main accomplishments - Review of regulatory criteria for seismic monitoring system -Analysis and upgrade of hardware system -Analysis and upgrade of software system - Development of seismic wave analysis system. (author). 17 refs., 49 figs., 6 tabs.

  15. Seismic behavior and design of wall-EDD-frame systems

    Directory of Open Access Journals (Sweden)

    Oren eLavan

    2015-06-01

    Full Text Available Walls and frames have different deflection lines and, depending on the seismic mass they support, may often poses different natural periods. In many cases, wall-frame structures present an advantageous behavior. In these structures the walls and the frames are rigidly connected. Nevertheless, if the walls and the frames were not rigidly connected, an opportunity for an efficient passive control strategy would arise: Connecting the two systems by energy dissipation devices (EDDs to result in wall-EDD-frame systems. This, depending on the parameters of the system, is expected to lead to an efficient energy dissipation mechanism.This paper studies the seismic behavior of wall-EDD-frame systems in the context of retrofitting existing frame structures. The controlling non-dimensional parameters of such systems are first identified. This is followed by a rigorous and extensive parametric study that reveals the pros and cons of the new system versus wall-frame systems. The effect of the controlling parameters on the behavior of the new system are analyzed and discussed. Finally, tools are given for initial design of such retrofitting schemes. These enable both choosing the most appropriate retrofitting alternative and selecting initial values for its parameters.

  16. Seismic Analysis for a Crane System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Soo; Lee, Chung Young; Ryu, Jeong Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The operation bridge used for an open-pool type research reactor is a crane system with a working deck for the handling of in-pool parts such as fuels, reactor components and reactor utilization facilities. The operation bridge allows operators to access the top of the reactor in the reactor pool and the fuel storage racks in the service pool. The operation bridge contains an operating platform mounted on a truck travelling on rails. Upright members are mounted on the truck to support the upper structure and two hoist monorails. The operation bridge consists of two hoists, upper girder frames, legs, cables, saddle frames, upper deck frames, lower deck frames, and the ladder. Static and dynamic analyses are performed to evaluate the structural integrity for the operation bridge for the required design loadings. The response spectrum analysis is employed as a dynamic analysis method

  17. Seismic Analysis for a Crane System

    International Nuclear Information System (INIS)

    Kim, Kang Soo; Lee, Chung Young; Ryu, Jeong Soo

    2012-01-01

    The operation bridge used for an open-pool type research reactor is a crane system with a working deck for the handling of in-pool parts such as fuels, reactor components and reactor utilization facilities. The operation bridge allows operators to access the top of the reactor in the reactor pool and the fuel storage racks in the service pool. The operation bridge contains an operating platform mounted on a truck travelling on rails. Upright members are mounted on the truck to support the upper structure and two hoist monorails. The operation bridge consists of two hoists, upper girder frames, legs, cables, saddle frames, upper deck frames, lower deck frames, and the ladder. Static and dynamic analyses are performed to evaluate the structural integrity for the operation bridge for the required design loadings. The response spectrum analysis is employed as a dynamic analysis method

  18. Seismic simulation and functional performance evaluation of a safety related, seismic category I control room emergency air cleaning system

    International Nuclear Information System (INIS)

    Manley, D.K.; Porco, R.D.; Choi, S.H.

    1985-01-01

    Under a nuclear contract MSA was required to design, manufacture, seismically test and functionally test a complete Safety Related, Seismic Category I, Control Room Emergency Air Cleaning System before shipment to the Yankee Atomic Electric Company, Yankee Nuclear Station in Rowe, Massachusetts. The installation of this system was required to satisfy the NRC requirements of NUREG-0737, Section III, D.3.4, ''Control Room Habitability''. The filter system tested was approximately 3 ft. wide by 8 ft. high by 18 ft. long and weighed an estimated 8300 pounds. It had a design flow rate of 3000 SCFM and contained four stages of filtration - prefilters, upstream and downstream HEPA filters and Type II sideload charcoal adsorber cells. The filter train design followed the guidelines set forth by ANSI/ASME N509-1980. Seismic Category I Qualification Testing consisted of resonance search testing and triaxial random multifrequency testing. In addition to ANSI/ASME N510-1980 testing, triaxial response accelerometers were placed at specific locations on designated prefilters, HEPA filters, charcoal adsorbers and test canisters along with accelerometers at the corresponding filter seal face locations. The purpose of this test was to demonstrate the integrity of the filters, filter seals, and monitor seismic response levels which is directly related to the system's ability to function during a seismic occurrence. The Control Room Emergency Air Cleaning System demonstrated the ability to withstand the maximum postulated earthquake for the plant site by remaining structurally sound and functional

  19. 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)

  20. Development of seismic isolation system in vertical direction

    International Nuclear Information System (INIS)

    Ohoka, Makoto; Horikiri, Morito

    1999-04-01

    A structure concept of vertical seismic isolation system which uses a common deck and a set of large dish springs was created in past studies. In this report, a series of dynamic tests on a small scale model of a common deck isolation structure were performed. The model was excited by random and seismic waves in the horizontal direction and 2-D excitation, horizontal and vertical, in order to identify the characteristics of isolation effect. The tests results are summarized as below. 1) This structure has three vibration mode. The second mode is rocking. 2) Rocking frequency depends on the excitation, for this structure has dish spring which contact with cylinders. Rocking damping varies from 2 to 8%, 3) Each mode's response peak frequency to 2-D(horizontal and vertical) excitation is almost the same the some to horizontal excitation. Vertical mode damping to 2-D excitation is about three times to horizontal excitation. 4) Isolation effect depends on a characteristics of frequency of input motion. The minimum response is to the Monju design seismic wave, soil shear wave:Vs=2000 m/sec, natural frequency of horizontal isolation in vertical direction:fv=20 Hz. A relative displacement is controlled. 5) A rocking angular displacement to 2-D excitation is about 2 times to 1-D excitation(vertical). However, it is about 1.2 E-4(rad), sufficiently small for a practical plant. (author)

  1. New developments in seismic analysis of primary and secondary systems

    International Nuclear Information System (INIS)

    Gupta, A.K.

    1984-01-01

    Primary and secondary systems often must be analyzed using decoupled models. This paper presents recent advances made at NCSU in the seismic analysis of these systems. Algorithms are presented by which coupled mode shapes and frequencies can be evaluated without performing a new eigenvalue solution, given the mode shapes and frequencies of the decoupled models. Simple and accurate equations are presented to predict changes in frequencies and responses. With the coupled mode shapes and frequencies, one can obtain any primary or secondary response directly from the input spectrum. Alternatively, one can develop instructure spectra at various locations in the primary system accounting for the primary-secondary system interaction. Correlation between the support motions is also generated. Equations are presented for evaluating complex mode shapes and frequencies of coupled systems when due to unequal damping values of primary and secondary systems, the coupled system becomes nonproportionally damped. Recent progress, in case of tuned systems is also reported

  2. Seismic analysis of hydraulic control rod driving system

    International Nuclear Information System (INIS)

    Zheng, Yanhua; Bo, Hanliang; Dong, Duo

    2002-01-01

    A simplified mathematical model was developed for the Hydraulic Control Rod Driving System (HCRDS) of a 200 MW nuclear heating reactor, which incorporated the design of its chamfer-hole step cylinder, to analyze its seismic response characteristics. The control rod motion was analyzed for different sine-wave vibration loadings on platform vibrator. The vibration frequency domain and the minimum acceleration amplitude of the control rod needed to cause the control rod to step to its next setting were compared with the design acceleration amplitude spectrum. The system design was found to be safety within the calculated limits. The safety margin increased with increasing frequency. (author)

  3. Risk assessment to determine the advisability of seismic trip systems

    International Nuclear Information System (INIS)

    Cummings, G.E.; Wells, J.E.

    1977-01-01

    Seismic trip (scram) systems have been used for many years on certain research, test, and production reactors, but not on commercial power reactors. An assessment is made of the risks associated with the presence and absence of such trip systems on power reactors. An attempt was made to go beyond the reactor per se and to consider the risks to society as a whole; for example, the advantages of tripping to avoid an earthquake-caused accident were weighed against the disadvantages associated with interrupting electric power in a time when it would be needed for emergency services. The comparative risk assessment was performed by means of fault tree analysis

  4. Seismic analysis of piping systems subjected to multiple support excitations

    International Nuclear Information System (INIS)

    Sundararajan, C.; Vaish, A.K.; Slagis, G.C.

    1981-01-01

    The paper presents the results of a comparative study between the multiple response spectrum method and the time-history method for the seismic analysis of nuclear piping systems subjected to different excitation at different supports or support groups. First, the necessary equations for the above analysis procedures are derived. Then, three actual nuclear piping systems subjected to single and multiple excitations are analyzed by the different methods, and extensive comparisons of the results (stresses) are made. Based on the results, it is concluded that the multiple response spectrum analysis gives acceptable results as compared to the ''exact'', but much more costly, time-history analysis. 6 refs

  5. 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.

  6. 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.

  7. Single Point Vulnerability Analysis of Automatic Seismic Trip System

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Seo Bin; Chung, Soon Il; Lee, Yong Suk [FNC Technology Co., Yongin (Korea, Republic of); Choi, Byung Pil [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    Single Point Vulnerability (SPV) analysis is a process used to identify individual equipment whose failure alone will result in a reactor trip, turbine generator failure, or power reduction of more than 50%. Automatic Seismic Trip System (ASTS) is a newly installed system to ensure the safety of plant when earthquake occurs. Since this system directly shuts down the reactor, the failure or malfunction of its system component can cause a reactor trip more frequently than other systems. Therefore, an SPV analysis of ASTS is necessary to maintain its essential performance. To analyze SPV for ASTS, failure mode and effect analysis (FMEA) and fault tree analysis (FTA) was performed. In this study, FMEA and FTA methods were performed to select SPV equipment of ASTS. D/O, D/I, A/I card, seismic sensor, and trip relay had an effect on the reactor trip but their single failure will not cause reactor trip. In conclusion, ASTS is excluded as SPV. These results can be utilized as the basis data for ways to enhance facility reliability such as design modification and improvement of preventive maintenance procedure.

  8. Single Point Vulnerability Analysis of Automatic Seismic Trip System

    International Nuclear Information System (INIS)

    Oh, Seo Bin; Chung, Soon Il; Lee, Yong Suk; Choi, Byung Pil

    2016-01-01

    Single Point Vulnerability (SPV) analysis is a process used to identify individual equipment whose failure alone will result in a reactor trip, turbine generator failure, or power reduction of more than 50%. Automatic Seismic Trip System (ASTS) is a newly installed system to ensure the safety of plant when earthquake occurs. Since this system directly shuts down the reactor, the failure or malfunction of its system component can cause a reactor trip more frequently than other systems. Therefore, an SPV analysis of ASTS is necessary to maintain its essential performance. To analyze SPV for ASTS, failure mode and effect analysis (FMEA) and fault tree analysis (FTA) was performed. In this study, FMEA and FTA methods were performed to select SPV equipment of ASTS. D/O, D/I, A/I card, seismic sensor, and trip relay had an effect on the reactor trip but their single failure will not cause reactor trip. In conclusion, ASTS is excluded as SPV. These results can be utilized as the basis data for ways to enhance facility reliability such as design modification and improvement of preventive maintenance procedure

  9. Seismic effects on technological equipment and systems of nuclear power plants

    International Nuclear Information System (INIS)

    Masopust, R.; Pecinka, L.; Podrouzek, J.

    1983-01-01

    A survey is given of problems related to the construction of nuclear power plants with regard to seismic resistance. Sei--smic resistance of technological equipment is evaluated by experimental trials, calculation or the combination of both. Existing and future standards are given for the given field. The Czechoslovak situation is discussed as related to the construction of the Mochovce nuclear power plant. Procedures for testing seismic resistance, types of tests and methods of simulating seismic excitation are described. Antiseismic measures together with structural elements for limiting the seismic effects on technological equipment and nuclear power plant systems are summed up on the basis of foreign experience. (E.F.)

  10. '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)

  11. Seismic Fragility of the LANL Fire Water Distribution System

    Energy Technology Data Exchange (ETDEWEB)

    Greg Mertz

    2007-03-30

    The purpose of this report is to present the results of a site-wide system fragility assessment. This assessment focuses solely on the performance of the water distribution systems that supply Chemical and Metallurgy Research (CMR), Weapons Engineering and Tritium Facility (WETF), Radioactive Liquid Waste Treatment Facility (RLWTF), Waste Characterization, Reduction, Repackaging Facility (WCRRF), and Transuranic Waste Inspectable Storage Project (TWISP). The analysis methodology is based on the American Lifelines Alliance seismic fragility formulations for water systems. System fragilities are convolved with the 1995 LANL seismic hazards to develop failure frequencies. Acceptance is determined by comparing the failure frequencies to the DOE-1020 Performance Goals. This study concludes that: (1) If a significant number of existing isolation valves in the water distribution system are closed to dedicate the entire water system to fighting fires in specific nuclear facilities; (2) Then, the water distribution systems for WETF, RLWTF, WCRRF, and TWISP meet the PC-2 performance goal and the water distribution system for CMR is capable of surviving a 0.06g earthquake. A parametric study of the WETF water distribution system demonstrates that: (1) If a significant number of valves in the water distribution system are NOT closed to dedicate the entire water system to fighting fires in WETF; (2) Then, the water distribution system for WETF has an annual probability of failure on the order of 4 x 10{sup -3} that does not meet the PC-2 performance goal. Similar conclusions are expected for CMR, RLWTF, WCRRF, and TWISP. It is important to note that some of the assumptions made in deriving the results should be verified by personnel in the safety-basis office and may need to be incorporated in technical surveillance requirements in the existing authorization basis documentation if credit for availability of fire protection water is taken at the PC-2 level earthquake levels

  12. Seismic Fragility of the LANL Fire Water Distribution System

    International Nuclear Information System (INIS)

    Greg Mertz Jason Cardon Mike Salmon

    2007-01-01

    The purpose of this report is to present the results of a site-wide system fragility assessment. This assessment focuses solely on the performance of the water distribution systems that supply Chemical and Metallurgy Research (CMR), Weapons Engineering and Tritium Facility (WETF), Radioactive Liquid Waste Treatment Facility (RLWTF), Waste Characterization, Reduction, Repackaging Facility (WCRRF), and Transuranic Waste Inspectable Storage Project (TWISP). The analysis methodology is based on the American Lifelines Alliance seismic fragility formulations for water systems. System fragilities are convolved with the 1995 LANL seismic hazards to develop failure frequencies. Acceptance is determined by comparing the failure frequencies to the DOE-1020 Performance Goals. This study concludes that: (1) If a significant number of existing isolation valves in the water distribution system are closed to dedicate the entire water system to fighting fires in specific nuclear facilities; (2) Then, the water distribution systems for WETF, RLWTF, WCRRF, and TWISP meet the PC-2 performance goal and the water distribution system for CMR is capable of surviving a 0.06g earthquake. A parametric study of the WETF water distribution system demonstrates that: (1) If a significant number of valves in the water distribution system are NOT closed to dedicate the entire water system to fighting fires in WETF; (2) Then, the water distribution system for WETF has an annual probability of failure on the order of 4 x 10 -3 that does not meet the PC-2 performance goal. Similar conclusions are expected for CMR, RLWTF, WCRRF, and TWISP. It is important to note that some of the assumptions made in deriving the results should be verified by personnel in the safety-basis office and may need to be incorporated in technical surveillance requirements in the existing authorization basis documentation if credit for availability of fire protection water is taken at the PC-2 level earthquake levels

  13. Towards the Understanding of Induced Seismicity in Enhanced Geothermal Systems

    Energy Technology Data Exchange (ETDEWEB)

    Gritto, Roland [Array Information Technology, Greenbelt, MD (United States); Dreger, Douglas [Univ. of California, Berkeley, CA (United States); Heidbach, Oliver [Helmholtz Centre Potsdam (Germany, German Research Center for Geosciences; Hutchings, Lawrence [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2014-08-29

    This DOE funded project was a collaborative effort between Array Information Technology (AIT), the University of California at Berkeley (UCB), the Helmholtz Centre Potsdam - German Research Center for Geosciences (GFZ) and the Lawrence Berkeley National Laboratory (LBNL). It was also part of the European research project “GEISER”, an international collaboration with 11 European partners from six countries including universities, research centers and industry, with the goal to address and mitigate the problems associated with induced seismicity in Enhanced Geothermal Systems (EGS). The goal of the current project was to develop a combination of techniques, which evaluate the relationship between enhanced geothermal operations and the induced stress changes and associated earthquakes throughout the reservoir and the surrounding country rock. The project addressed the following questions: how enhanced geothermal activity changes the local and regional stress field; whether these activities can induce medium sized seismicity M > 3; (if so) how these events are correlated to geothermal activity in space and time; what is the largest possible event and strongest ground motion, and hence the potential hazard associated with these activities. The development of appropriate technology to thoroughly investigate and address these questions required a number of datasets to provide the different physical measurements distributed in space and time. Because such a dataset did not yet exist for an EGS system in the United State, we used current and past data from The Geysers geothermal field in northern California, which has been in operation since the 1960s. The research addressed the need to understand the causal mechanisms of induced seismicity, and demonstrated the advantage of imaging the physical properties and temporal changes of the reservoir. The work helped to model the relationship between injection and production and medium sized magnitude events that have

  14. 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

  15. R and D on seismic emergency information system

    International Nuclear Information System (INIS)

    2001-06-01

    After the Great Hansin-Awaji Earthquake Disaster occurred in 1995, the Science and Technology Agency commenced 'Frontier Research Program on Earthquake' in FY1996. As a part of this research program, four-year program on 'Research on Real-time Earthquake Information Transmission' has been carried out at JAERI since FY1997. Through the experience of the above earthquake disaster, the importance of accurate and prompt seismic information transmission immediately after the occurrence of the earthquake has been recognized from the viewpoint of disaster mitigation. Under this circumstance, the main activity in Real-time Earthquake Information Transmission Research at JAERI has been placed on the development of a seismic emergency information system. In order to respond to the above R and D, Seismic Emergency Information System Research Team was organized in JAERI in FY1998. In the meantime, a part of this R and D program is performed under the coordinated research between JAERI and NIED (National Research Institute for Earth Science and Disaster Prevention). This report summarizes the results of four years program from FY1997 to FY2000 on the above R and D. The R and D has been conducted involving the latest progress in earthquake engineering with regard to estimation techniques on the hypocenter, fault and earthquake motion parameters and in Information Technologies. The R and D was divided into two parts, i.e., development of the basic system and application system. In the basic system, earthquake information with 500 m square mesh in a local area can be estimated and transmitted in a few minutes. In the application system, the concept of the disaster management system which consists of user site and disaster information center and is capable of mutual information transmission has been established. A prototype of the application system, which include the basic system in the disaster information center, has been developed. Test operation of the basic system in being

  16. Development of Seismic Isolation Systems Using Periodic Materials

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Yiqun [Univ. of Houston, Houston, TX (United States); Mo, Yi-Lung [Univ. of Houston, Houston, TX (United States); Menq, Farn-Yuh [Univ. of Texas, Austin, TX (United States); Stokoe, II, Kenneth H. [Univ. of Texas, Austin, TX (United States); Perkins, Judy [Prairie View A & M University, Prairie View, TX (United States); Tang, Yu [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-12-10

    Advanced fast nuclear power plants and small modular fast reactors are composed of thin-walled structures such as pipes; as a result, they do not have sufficient inherent strength to resist seismic loads. Seismic isolation, therefore, is an effective solution for mitigating earthquake hazards for these types of structures. Base isolation, on which numerous studies have been conducted, is a well-defined structure protection system against earthquakes. In conventional isolators, such as high-damping rubber bearings, lead-rubber bearings, and friction pendulum bearings, large relative displacements occur between upper structures and foundations. Only isolation in a horizontal direction is provided; these features are not desirable for the piping systems. The concept of periodic materials, based on the theory of solid-state physics, can be applied to earthquake engineering. The periodic material is a material that possesses distinct characteristics that prevent waves with certain frequencies from being transmitted through it; therefore, this material can be used in structural foundations to block unwanted seismic waves with certain frequencies. The frequency band of periodic material that can filter out waves is called the band gap, and the structural foundation made of periodic material is referred to as the periodic foundation. The design of a nuclear power plant, therefore, can be unified around the desirable feature of a periodic foundation, while the continuous maintenance of the structure is not needed. In this research project, three different types of periodic foundations were studied: one-dimensional, two-dimensional, and three-dimensional. The basic theories of periodic foundations are introduced first to find the band gaps; then the finite element methods are used, to perform parametric analysis, and obtain attenuation zones; finally, experimental programs are conducted, and the test data are analyzed to verify the theory. This procedure shows that the

  17. Seismicity within the Irpinia Fault System As Monitored By Isnet (Irpinia Seismic Network) and Its Possible Relation with Fluid Storage

    Science.gov (United States)

    Festa, G.; Zollo, A.; Amoroso, O.; Ascione, A.; Colombelli, S.; Elia, L.; Emolo, A.; Martino, C.; Mazzoli, S.; Orefice, A.; Russo, G.

    2014-12-01

    ISNet (http://isnet.fisica.unina.it) is deployed in Southern Apennines along the active fault system responsible for the 1980, M 6.9 Irpinia earthquake. ISNet consists of 32 seismic stations equipped with both strong motion and velocimetric instruments (either broadband or short-period), with the aim of capture a broad set of seismic signals, from ambient noise to strong motion. Real time and near real time procedures run at ISNet with the goal of monitoring the seismicity, check possible space-time anomalies, detect seismic sequences and launch an earthquake early warning in the case of potential significant ground shaking in the area. To understand the role of fluids on the seismicity of the area, we investigated velocity and attenuation models. The former is built from accurate cross-correlation picking and S wave detection based onto polarization analysis. Joint inversion of both P and S arrival times is then based on a linearized multi-scale tomographic approach. Attenuation is instead obtained from inversion of displacement spectra, deconvolving for the source effect. High VP/VS and QS/QP >1 were found within a ~15 km wide rock volume where intense microseismicity is located. This indicates that concentration of seismicity is possibly controlled by high pore fluid pressure. This earthquake reservoir may come from a positive feedback between the seismic pumping that controls the fluid transmission through the fractured damage zone and the low permeability of cross fault barrier, increasing the fluid pore pressure within the fault bounded block. In this picture, sequences mostly occur at the base of this fluid rich layer. They show an anomalous pattern in the earthquake occurrence per magnitude classes; main events evolve with a complex source kinematics, as obtained from backprojection of apparent source time functions, indicating possible directivity effects. In this area sequences might be the key for understanding the transition between the deep

  18. Automatic seismic support design of piping system by an object oriented expert system

    International Nuclear Information System (INIS)

    Nakatogawa, T.; Takayama, Y.; Hayashi, Y.; Fukuda, T.; Yamamoto, Y.; Haruna, T.

    1990-01-01

    The seismic support design of piping systems of nuclear power plants requires many experienced engineers and plenty of man-hours, because the seismic design conditions are very severe, the bulk volume of the piping systems is hyge and the design procedures are very complicated. Therefore we have developed a piping seismic design expert system, which utilizes the piping design data base of a 3 dimensional CAD system and automatically determines the piping support locations and support styles. The data base of this system contains the maximum allowable seismic support span lengths for straight piping and the span length reduction factors for bends, branches, concentrated masses in the piping, and so forth. The system automatically produces the support design according to the design knowledge extracted and collected from expert design engineers, and using design information such as piping specifications which give diameters and thickness and piping geometric configurations. The automatic seismic support design provided by this expert system achieves in the reduction of design man-hours, improvement of design quality, verification of design result, optimization of support locations and prevention of input duplication. In the development of this system, we had to derive the design logic from expert design engineers and this could not be simply expressed descriptively. Also we had to make programs for different kinds of design knowledge. For these reasons we adopted the object oriented programming paradigm (Smalltalk-80) which is suitable for combining programs and carrying out the design work

  19. Design and implement of system for browsing remote seismic waveform based on B/S schema

    International Nuclear Information System (INIS)

    Zheng Xuefeng; Shen Junyi; Wang Zhihai; Sun Peng; Jin Ping; Yan Feng

    2006-01-01

    Browsing remote seismic waveform based on B/S schema is of significance in modern seismic research and data service, and the technology should be improved urgently. This paper describes the basic plan, architecture and implement of system for browsing remote seismic waveform based on B/S schema. The problem to access, browse and edit the waveform data on serve from client only using browser has been solved. On this basis, the system has been established and been in use. (authors)

  20. Seismic behavior of steel storage pallet racking systems

    CERN Document Server

    Castiglioni, Carlo Andrea

    2016-01-01

    This book presents the main outcomes of the first European research project on the seismic behavior of adjustable steel storage pallet racking systems. In particular, it describes a comprehensive and unique set of full-scale tests designed to assess such behavior. The tests performed include cyclic tests of full-scale rack components, namely beam-to-upright connections and column base connections; static and dynamic tests to assess the friction factor between pallets and rack beams; full-scale pushover and pseudodynamic tests of storage racks in down-aisle and cross-aisle directions; and full-scale dynamic tests on two-bay, three-level rack models. The implications of the findings of this extensive testing regime on the seismic behavior of racking systems are discussed in detail, highlighting e.g. the confirmation that under severe dynamic conditions “sliding” is the main factor influencing rack response. This work was conceived during the development of the SEISRACKS project. Its outcomes will contribute...

  1. 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

  2. Seismic properties of fluid bearing formations in magmatic geothermal systems: can we directly detect geothermal activity with seismic methods?

    Science.gov (United States)

    Grab, Melchior; Scott, Samuel; Quintal, Beatriz; Caspari, Eva; Maurer, Hansruedi; Greenhalgh, Stewart

    2016-04-01

    Seismic methods are amongst the most common techniques to explore the earth's subsurface. Seismic properties such as velocities, impedance contrasts and attenuation enable the characterization of the rocks in a geothermal system. The most important goal of geothermal exploration, however, is to describe the enthalpy state of the pore fluids, which act as the main transport medium for the geothermal heat, and to detect permeable structures such as fracture networks, which control the movement of these pore fluids in the subsurface. Since the quantities measured with seismic methods are only indirectly related with the fluid state and the rock permeability, the interpretation of seismic datasets is difficult and usually delivers ambiguous results. To help overcome this problem, we use a numerical modeling tool that quantifies the seismic properties of fractured rock formations that are typically found in magmatic geothermal systems. We incorporate the physics of the pore fluids, ranging from the liquid to the boiling and ultimately vapor state. Furthermore, we consider the hydromechanics of permeable structures at different scales from small cooling joints to large caldera faults as are known to be present in volcanic systems. Our modeling techniques simulate oscillatory compressibility and shear tests and yield the P- and S-wave velocities and attenuation factors of fluid saturated fractured rock volumes. To apply this modeling technique to realistic scenarios, numerous input parameters need to be indentified. The properties of the rock matrix and individual fractures were derived from extensive literature research including a large number of laboratory-based studies. The geometries of fracture networks were provided by structural geologists from their published studies of outcrops. Finally, the physical properties of the pore fluid, ranging from those at ambient pressures and temperatures up to the supercritical conditions, were taken from the fluid physics

  3. Development of real time monitor system displaying seismic waveform data observed at seafloor seismic network, DONET, for disaster management information

    Science.gov (United States)

    Horikawa, H.; Takaesu, M.; Sueki, K.; Takahashi, N.; Sonoda, A.; Miura, S.; Tsuboi, S.

    2014-12-01

    Mega-thrust earthquakes are anticipated to occur in the Nankai Trough in southwest Japan. In the source areas, we have deployed seafloor seismic network, DONET (Dense Ocean-floor Network System for Earthquake and Tsunamis), in 2010 in order to monitor seismicity, crustal deformations, and tsunamis. DONET system consists of totally 20 stations, which is composed of six kinds of sensors, including strong-motion seismometers and quartz pressure gauges. Those stations are densely distributed with an average spatial interval of 15-20 km and cover near the trench axis to coastal areas. Observed data are transferred to a land station through a fiber-optical cable and then to JAMSTEC (Japan Agency for Marine-Earth Science and Technology) data management center through a private network in real time. After 2011 off the Pacific coast of Tohoku Earthquake, each local government close to Nankai Trough try to plan disaster prevention scheme. JAMSTEC will disseminate DONET data combined with research accomplishment so that they will be widely recognized as important earthquake information. In order to open DONET data observed for research to local government, we have developed a web application system, REIS (Real-time Earthquake Information System). REIS is providing seismic waveform data to some local governments close to Nankai Trough as a pilot study. As soon as operation of DONET is ready, REIS will start full-scale operation. REIS can display seismic waveform data of DONET in real-time, users can select strong motion and pressure data, and configure the options of trace view arrangement, time scale, and amplitude. In addition to real-time monitoring, REIS can display past seismic waveform data and show earthquake epicenters on the map. In this presentation, we briefly introduce DONET system and then show our web application system. We also discuss our future plans for further developments of REIS.

  4. A seismic data compression system using subband coding

    Science.gov (United States)

    Kiely, A. B.; Pollara, F.

    1995-01-01

    This article presents a study of seismic data compression techniques and a compression algorithm based on subband coding. The algorithm includes three stages: a decorrelation stage, a quantization stage that introduces a controlled amount of distortion to allow for high compression ratios, and a lossless entropy coding stage based on a simple but efficient arithmetic coding method. Subband coding methods are particularly suited to the decorrelation of nonstationary processes such as seismic events. Adaptivity to the nonstationary behavior of the waveform is achieved by dividing the data into separate blocks that are encoded separately with an adaptive arithmetic encoder. This is done with high efficiency due to the low overhead introduced by the arithmetic encoder in specifying its parameters. The technique could be used as a progressive transmission system, where successive refinements of the data can be requested by the user. This allows seismologists to first examine a coarse version of waveforms with minimal usage of the channel and then decide where refinements are required. Rate-distortion performance results are presented and comparisons are made with two block transform methods.

  5. R and D of seismic emergency information transmission system

    International Nuclear Information System (INIS)

    Ebisawa, Katsumi; Kuno, Tetsuya; Shibata, Katsuyuki; Abe, Ichiro; Tuzuki, Kazuhisa

    2002-01-01

    The R and D Seismic Emergency Information Transmission System has been conducted involving the latest progress in earthquake engineering with regard to estimation techniques on the hypocenter, fault and earthquake motion parameters and in information technologies. This system is the disaster management system which consists of user site and disaster information center and is capable of mutual information transmission through Inter-Net and walkie-talkie. The concept of the disaster management system which is adaptable with DiMSIS (Disaster Management Spatial Information System) developed by professor Kameda et al. of Kyoto University has been established. Based on this concept, a prototype system has been developed. This system has following functions, (1) compatible application both in usual condition and emergency time, (2) the decentralized independence, and (3) the integration of space and time information. The system can estimate the earthquake motion information with 500 m square mesh in a local area and transmit in a few minutes. In the development of the system, seismometer network, surface soil database and amplification functions were prepared for the examination of system function. Demonstration against the Tokai area was carried out and the function was verified. (author)

  6. Tracking changes in volcanic systems with seismic Interferometry

    Science.gov (United States)

    Haney, Matt; Alicia J. Hotovec-Ellis,; Bennington, Ninfa L.; Silvio De Angelis,; Clifford Thurber,

    2014-01-01

    The detection and evaluation of time-dependent changes at volcanoes form the foundation upon which successful volcano monitoring is built. Temporal changes at volcanoes occur over all time scales and may be obvious (e.g., earthquake swarms) or subtle (e.g., a slow, steady increase in the level of tremor). Some of the most challenging types of time-dependent change to detect are subtle variations in material properties beneath active volcanoes. Although difficult to measure, such changes carry important information about stresses and fluids present within hydrothermal and magmatic systems. These changes are imprinted on seismic waves that propagate through volcanoes. In recent years, there has been a quantum leap in the ability to detect subtle structural changes systematically at volcanoes with seismic waves. The new methodology is based on the idea that useful seismic signals can be generated “at will” from seismic noise. This means signals can be measured any time, in contrast to the often irregular and unpredictable times of earthquakes. With seismic noise in the frequency band 0.1–1 Hz arising from the interaction of the ocean with the solid Earth known as microseisms, researchers have demonstrated that cross-correlations of passive seismic recordings between pairs of seismometers yield coherent signals (Campillo and Paul 2003; Shapiro and Campillo 2004). Based on this principle, coherent signals have been reconstructed from noise recordings in such diverse fields as helioseismology (Rickett and Claerbout 2000), ultrasound (Weaver and Lobkis 2001), ocean acoustic waves (Roux and Kuperman 2004), regional (Shapiro et al. 2005; Sabra et al. 2005; Bensen et al. 2007) and exploration (Draganov et al. 2007) seismology, atmospheric infrasound (Haney 2009), and studies of the cryosphere (Marsan et al. 2012). Initial applications of ambient seismic noise were to regional surface wave tomography (Shapiro et al. 2005). Brenguier et al. (2007) were the first to

  7. 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)

  8. 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.)

  9. 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

  10. 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

  11. 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

  12. Design and implementation of a unified certification management system based on seismic business

    Science.gov (United States)

    Tang, Hongliang

    2018-04-01

    Many business software for seismic systems are based on web pages, users can simply open a browser and enter their IP address. However, how to achieve unified management and security management of many IP addresses, this paper introduces the design concept based on seismic business and builds a unified authentication management system using ASP technology.

  13. Evolution of seismic monitoring systems of nuclear power plants. Improvements and practical applications

    International Nuclear Information System (INIS)

    Sanchez Cabanero, J. G.; Jimenez Juan, A.

    2010-01-01

    The II. NN. Spanish have a seismic monitoring system (SVS) covering two objectives relevant to nuclear security: determining earthquake leave operation, and specific data that serve to limit or reduce the uncertainties associated with the seismic source, the site and design. Since its construction, the major SVS II. NN. have been equipped with the best time of seismic instrumentation to record earthquakes strong, but with limited resolution for recording in the free field and appropriately moderate earthquakes.

  14. 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.

  15. A guidebook for the operation and maintenance of HANARO seismic monitoring analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Jeong Soo; Yoon, Doo Byung; Kim, Hyung Kyoo

    2003-09-01

    Systems and structures related to HANARO safety are classified as seismic category I. Since 1995, the seismic monitoring system has been utilized for monitoring an earthquake at the HANARO site. The existing seismic monitoring system consists of field sensors and monitoring panel. The analog-type monitoring system with magnetic tape recorder is out-of-date model. In addition, the disadvantage of the existing system is that it does not include signal-analyzing equipment. Therefore, we have improved the analog seismic monitoring system into a new digital Seismic Monitoring Analysis System(SMAS) that can offer precise and detail information of the earthquake signals. This newly developed SMAS is operating at the HANARO instrument room to acquire and analyze the signal of an earthquake. This document is a guidebook for the operation and maintenance of the SMAS. The first chapter gives an outline of the SMAS. The second chapter describes functional capability and specification of the hardware. Chapters 3 and 4 describe starting procedure of the SMAS and how to operate the seismic monitoring program, respectively. Chapter 5 illustrates the seismic analysis algorithm used in the SMAS. The way of operating the seismic analysis program is described in chapter 6. Chapter 7 illustrates the calibration procedure for data acquisition module. Chapter 8 describes the symptoms of common malfunctions and its countermeasure suited to the occasions.

  16. The GNSS Component of the Seismic Monitoring System in Chile

    Science.gov (United States)

    Barrientos, S. E.

    2016-12-01

    Chile is amongst the most seismically active countries in the world. Since mid-XVI Century, a magnitude 8 or more earthquake has taken place every dozen of years, as an average. In the last 100 years, more than ten events with magnitudes around 8 or larger have taken place in this part of world. Three events with M>8 have taken place only in the last six years. The largest earthquake ever recorded took place in May, 1960, in southern Chile. Such extreme seismic activity is the result of the interaction of the Nazca, Antarctic, Scotia and South American plates in southwestern South America where Chile is located. These megathrust earthquakes exhibit long rupture regions reaching several hundreds of km with fault displacements of several tens of meters. At least eighteen of these earthquakes have generated local tsunamis with runups larger than 4 m -including events in 2010, 2014 and 2015- therefore it is mandatory to establish a system with capabilities to rapidly evaluate the tsunamigenic potential of these events. In 2013, the newly created National Seismological Center (CSN) of the University of Chile was tasked to upgrade the countrýs seismic network by increasing the numbers of real-time monitoring stations. The most important change to previous practices is the establishment of a GNSS network composed by 130 devices, in addition to the incorporation of 65 new collocated broadband and strong motion instruments. Additional 297 strong motion instruments for engineering purposes complement the system. Forty units -of the 130 devices- present an optional RTX capability, where satellite orbits and clock corrections are sent to the field device producing a 1-Hz position stream at 4-cm level. First records of ground displacement -using this technology-were recorded at the time of the largest aftershock (Mw=7.6) of the sequence that affected northern Chile in 2014. The CSN is currently developing automatic detectors and amplitude estimators of displacement from the

  17. Seismic ratchet-fatigue failure of piping systems

    International Nuclear Information System (INIS)

    Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

    1987-01-01

    Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-inch and a pressurized 6-inch diameter carbon steel nuclear pipe systems subjected to high-level shaking have been accomplished. The high-level shaking loads needed to cause failure were much higher than ASME Code rules would permit with present design limits. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occured in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate reasonably well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules to reduce unneeded conservatisms and to cover the ratchet-fatigue failure mode may be appropriate

  18. Seismic ratchet-fatigue failure of piping systems

    International Nuclear Information System (INIS)

    Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

    1986-01-01

    Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-in. and a pressurized 6-in. diameter carbon steel nuclear pipe systems subjected to high level shaking have been accomplished. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occurred in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate very well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules may be appropriate to cover the ratchet-fatigue failure mode

  19. 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

  20. Analysis of piping system response to seismic excitations

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1987-01-01

    This paper describes a numerical algorithm for analyzing piping system response to seismic excitations. The numerical model of the piping considers hoop, flexural, axial, and torsional modes of deformation. Hoop modes generated from internal hydrodynamic loading are superimposed on the bending and twisting modes by two extra degrees of freedom. A time-history analysis technique using the implicit temporal integration scheme is addressed. The time integrator uses a predictor-corrector successive iterative scheme which satisfies the equation of motion. Both geometrical and material nonlinearities are considered. Multiple support excitations, fluid effect, piping insulation, and material dampings can be included in the analysis. Two problems are presented to illustrate the method. The results are discussed in detail

  1. 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

  2. Seismic Safety Margins Research Program (Phase I). Project VII. Systems analysis specification of computational approach

    International Nuclear Information System (INIS)

    Wall, I.B.; Kaul, M.K.; Post, R.I.; Tagart, S.W. Jr.; Vinson, T.J.

    1979-02-01

    An initial specification is presented of a computation approach for a probabilistic risk assessment model for use in the Seismic Safety Margin Research Program. This model encompasses the whole seismic calculational chain from seismic input through soil-structure interaction, transfer functions to the probability of component failure, integration of these failures into a system model and thereby estimate the probability of a release of radioactive material to the environment. It is intended that the primary use of this model will be in sensitivity studies to assess the potential conservatism of different modeling elements in the chain and to provide guidance on priorities for research in seismic design of nuclear power plants

  3. 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)

  4. 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.

  5. The roles of the seismic safety and monitoring systems in the PEC fast reactor

    International Nuclear Information System (INIS)

    Masoni, P.; Di Tullio, E.M.; Massa, B.; Martelli, A.; Sano, T.

    1988-01-01

    Two different seismic systems are foreseen in the case of PEC: the seismic safety system, that provides the automatic scram, and the seismic monitoring system. During earthquake, three triaxial seismic switches are triggered if a threshold value of the ground acceleration is exceeded. In this case, the signals from the seismic switches are processed by the safety system (with a 2/3 logic) and the shutdown system is triggered. Peak acceleration is the parameter used by the safety system to quantify the seismic event. This way, however, no information is obtained with regard to earthquake frequency content. Thus, reactor safety is guaranteed by adopting a threshold considerably lower than the Z.P.A. of the Design Basis Earthquake. Furthermore, in the case of significant earthquakes, the seismic motion is measured by about 20 triaxial accelerometers, located both in the free field and on the plant's structures. Data are digitazed and recordered by the seismic monitoring system. This system also elaborates the recordered time-histories providing floor response spectra and compares such spectra to the design values. The above-mentioned elaborations and comparisons are performed in short time for two triaxial measuring positions, thus allowing the Operator to immediately get a more complete information on the seismic event. The complete set of data recorded by the seismic monitoring system also allows the actual dynamic response of the plant to be determined and compared to the design values. On the basis of this comparison the necessary safety analysis can be carried out to verify whether the design limits of the plant were respected: in the positive case the reactor can be restarted. (author)

  6. 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.

  7. 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

  8. A Seismic Transmission System for Continuous Monitoring of the Lithosphere : A Proposition

    NARCIS (Netherlands)

    Unger, R.

    2002-01-01

    The main objective of this thesis is to enhance earthquake prediction feasibility. We present the concept and the design layout of a novel seismic transmission system capable of continuously monitoring the Lithosphere for changes in Earth physics parameters governing seismic wave propagation.

  9. A Shear-Wave Seismic System to Look Ahead of a Tunnel Boring Machine

    NARCIS (Netherlands)

    Bharadwaj, Pawan; Drijkoningen, G.G.; Mulder, W.A.; Tscharner, Thomas; Jenneskens, Rob

    2016-01-01

    The Earth’s properties, composition and structure ahead of a tunnel boring machine (TBM) should be mapped for hazard assessment during excavation. We study the use of seismic-exploration techniques for this purpose. We focus on a seismic system for soft soils, where shear waves are better and easier

  10. Seismic qualification of the rotary relay for use in the solid state protection system

    International Nuclear Information System (INIS)

    Vogeding, E.L.; Jarecki, S.J.

    1976-01-01

    The seismic qualification of a rotary relay that can be used as a replacement for the type of relay located in the output section of the Solid State Protection System is described. The qualification test results indicate that the tested relays did not exhibit any contact bounce or abnormal operation; they performed satisfactorily before, during, and after the simulated seismic vibration tests

  11. Connection with seismic networks and construction of real time earthquake monitoring system

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Heon Cheol; Lee, H. I.; Shin, I. C.; Lim, I. S.; Park, J. H.; Lee, B. K.; Whee, K. H.; Cho, C. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-12-15

    It is natural to use the nuclear power plant seismic network which have been operated by KEPRI(Korea Electric Power Research Institute) and local seismic network by KIGAM(Korea Institute of Geology, Mining and Material). The real time earthquake monitoring system is composed with monitoring module and data base module. Data base module plays role of seismic data storage and classification and the other, monitoring module represents the status of acceleration in the nuclear power plant area. This research placed the target on the first, networking the KIN's seismic monitoring system with KIGAM and KEPRI seismic network and the second, construction the KIN's Independent earthquake monitoring system.

  12. 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

  13. 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.

  14. 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

  15. A Fiber-Optic Borehole Seismic Vector Sensor System for Geothermal Site Characterization and Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Paulsson, Bjorn N.P. [Paulsson, Inc., Van Nuys, CA (United States); Thornburg, Jon A. [Paulsson, Inc., Van Nuys, CA (United States); He, Ruiqing [Paulsson, Inc., Van Nuys, CA (United States)

    2015-04-21

    Seismic techniques are the dominant geophysical techniques for the characterization of subsurface structures and stratigraphy. The seismic techniques also dominate the monitoring and mapping of reservoir injection and production processes. Borehole seismology, of all the seismic techniques, despite its current shortcomings, has been shown to provide the highest resolution characterization and most precise monitoring results because it generates higher signal to noise ratio and higher frequency data than surface seismic techniques. The operational environments for borehole seismic instruments are however much more demanding than for surface seismic instruments making both the instruments and the installation much more expensive. The current state-of-the-art borehole seismic instruments have not been robust enough for long term monitoring compounding the problems with expensive instruments and installations. Furthermore, they have also not been able to record the large bandwidth data available in boreholes or having the sensitivity allowing them to record small high frequency micro seismic events with high vector fidelity. To reliably achieve high resolution characterization and long term monitoring of Enhanced Geothermal Systems (EGS) sites a new generation of borehole seismic instruments must therefore be developed and deployed. To address the critical site characterization and monitoring needs for EGS programs, US Department of Energy (DOE) funded Paulsson, Inc. in 2010 to develop a fiber optic based ultra-large bandwidth clamped borehole seismic vector array capable of deploying up to one thousand 3C sensor pods suitable for deployment into ultra-high temperature and high pressure boreholes. Tests of the fiber optic seismic vector sensors developed on the DOE funding have shown that the new borehole seismic sensor technology is capable of generating outstanding high vector fidelity data with extremely large bandwidth: 0.01 – 6,000 Hz. Field tests have shown

  16. 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.)

  17. 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

  18. Investigation of optimal seismic design methodology for piping systems supported by elasto-plastic dampers. Part. 2. Applicability for seismic waves with various frequency characteristics

    International Nuclear Information System (INIS)

    Ito, Tomohiro; Michiue, Masashi; Fujita, Katsuhisa

    2010-01-01

    In this study, the applicability of a previously developed optimal seismic design methodology, which can consider the structural integrity of not only piping systems but also elasto-plastic supporting devices, is studied for seismic waves with various frequency characteristics. This methodology employs a genetic algorithm and can search the optimal conditions such as the supporting location and the capacity and stiffness of the supporting devices. Here, a lead extrusion damper is treated as a typical elasto-plastic damper. Numerical simulations are performed using a simple piping system model. As a result, it is shown that the proposed optimal seismic design methodology is applicable to the seismic design of piping systems subjected to seismic waves with various frequency characteristics. The mechanism of optimization is also clarified. (author)

  19. System seismic analysis of an innovative primary system for a large pool type LMFBR plant

    International Nuclear Information System (INIS)

    Pan, Y.C.; Wu, T.S.; Cha, B.K.; Burelbach, J.; Seidensticker, R.

    1984-01-01

    The system seismic analysis of an innovative primary system for a large pool type liquid metal fast breeder reactor (LMFBR) plant is presented. In this primary system, the reactor core is supported in a way which differs significantly from that used in previous designs. The analytical model developed for this study is a three-dimensional finite element model including one-half of the primary system cut along the plane of symmetry. The model includes the deck and deck mounted components,the reactor vessel, the core support structure, the core barrel, the radial neutron shield, the redan, and the conical support skirt. The sodium contained in the primary system is treated as a lumped mass appropriately distributed among various components. The significant seismic behavior as well as the advantages of this primary system design are discussed in detail

  20. Protocol for Addressing Induced Seismicity Associated with Enhanced Geothermal Systems

    Energy Technology Data Exchange (ETDEWEB)

    Majer, Ernie [Office of Energy Efficiency and Renewable Energy (EERE), Washington, DC (United States); Nelson, James [Office of Energy Efficiency and Renewable Energy (EERE), Washington, DC (United States); Robertson-Tait, Ann [Office of Energy Efficiency and Renewable Energy (EERE), Washington, DC (United States); Savy, Jean [Office of Energy Efficiency and Renewable Energy (EERE), Washington, DC (United States); Wong, Ivan [Office of Energy Efficiency and Renewable Energy (EERE), Washington, DC (United States)

    2012-01-01

    This Protocol is a living guidance document for geothermal developers, public officials, regulators and the general public that provides a set of general guidelines detailing useful steps to evaluate and manage the effects of induced seismicity related to EGS projects.

  1. Seismic prospecting using a continuous shooting and continuous recording system

    International Nuclear Information System (INIS)

    Wason, C.B.

    1984-01-01

    A method of seismic prospecting is disclosed in which the seismic source is excited in such a manner as to maximize the use of the energy generated by the seismic source. In certain cases it may be desirable to convert the received seismic signals to their frequency domain counterparts before performing subsequent processing. Such conversion may be performed using the discrete Fourier transform with the result that transformed values are obtained only at certain discrete frequencies. It may further be desirable that processing be performed only at subsets of the total set of discrete frequencies with the values at the remaining frequencies being discarded. In the practice of the present invention, source energy is generated only at those discrete frequencies at which subsequent processing is to be performed. As a result there is substantially no source energy in the transform values at the frequencies which are discarded

  2. Views on seismic design standardization of structures, systems and components of nuclear facilities

    International Nuclear Information System (INIS)

    Reddy, G.R.

    2011-01-01

    Structures, Systems and Components (SSCs) of nuclear facilities have to be designed for normal operating loads such as dead weight, pressure, temperature etc., and accidental loads such as earthquakes, floods, extreme, wind air craft impact, explosions etc. Manmade accidents such as aircraft impact, explosions etc., sometimes may be considered as design basis event and sometimes taken care by providing administrative controls. This will not be possible in the case of natural events such as earthquakes, flooding, extreme winds etc. Among natural events earthquakes are considered as most devastating and need to be considered as design basis event which has certain annual frequency specified in design codes. For example nuclear power plants are designed for a seismic event has 10000 year return period. It is generally felt that design of SSCs for earthquake loads is very time consuming and expensive. Conventional seismic design approaches demands for large number of supports for systems and components. This results in large space occupation and in turn creates difficulties for maintenance and in service inspection of systems and components. In addition, complete exercise of design need to be repeated for plants being located at different sites due to different seismic demands. However, advanced seismic response control methods will help to standardize the seismic design meeting the safety and economy. These methods adopt passive, semi active and active devices, and base isolators to control the seismic response. In nuclear industry, it is advisable to go for passive devices to control the seismic responses. Ideally speaking, these methods will make the designs made for normal loads can also satisfy the seismic demand without calling for change in material, geometry, layout etc. in the SSCs. This paper explain the basic ideas of seismic response control methods, demonstrate the effectiveness of control methods through case studies and eventually give the procedure to

  3. Preliminary seismic design of dynamically coupled structural systems

    International Nuclear Information System (INIS)

    Pal, N.; Dalcher, A.W.; Gluck, R.

    1977-01-01

    In this paper, the analysis criteria for coupling and decoupling, which are most commonly used in nuclear design practice, are briefly reviewed and a procedure outlined and demonstrated with examples. Next, a criterion judged to be practical for preliminary seismic design purposes is defined. Subsequently, a technique compatible with this criterion is suggested. A few examples are presented to test the proposed procedure for preliminary seismic design purposes. Limitations of the procedure are also discussed and finally, the more important conclusions are summarized

  4. Development of 3-axis precise positioning seismic physical modeling system in the simulation of marine seismic exploration

    Science.gov (United States)

    Kim, D.; Shin, S.; Ha, J.; Lee, D.; Lim, Y.; Chung, W.

    2017-12-01

    Seismic physical modeling is a laboratory-scale experiment that deals with the actual and physical phenomena that may occur in the field. In seismic physical modeling, field conditions are downscaled and used. For this reason, even a small error may lead to a big error in an actual field. Accordingly, the positions of the source and the receiver must be precisely controlled in scale modeling. In this study, we have developed a seismic physical modeling system capable of precisely controlling the 3-axis position. For automatic and precise position control of an ultrasonic transducer(source and receiver) in the directions of the three axes(x, y, and z), a motor was mounted on each of the three axes. The motor can automatically and precisely control the positions with positional precision of 2''; for the x and y axes and 0.05 mm for the z axis. As it can automatically and precisely control the positions in the directions of the three axes, it has an advantage in that simulations can be carried out using the latest exploration techniques, such as OBS and Broadband Seismic. For the signal generation section, a waveform generator that can produce a maximum of two sources was used, and for the data acquisition section, which receives and stores reflected signals, an A/D converter that can receive a maximum of four signals was used. As multiple sources and receivers could be used at the same time, the system was set up in such a way that diverse exploration methods, such as single channel, multichannel, and 3-D exploration, could be realized. A computer control program based on LabVIEW was created, so that it could control the position of the transducer, determine the data acquisition parameters, and check the exploration data and progress in real time. A marine environment was simulated using a water tank 1 m wide, 1 m long, and 0.9 m high. To evaluate the performance and applicability of the seismic physical modeling system developed in this study, single channel and

  5. Seismic margin reviews of nuclear power plants: Identification of important functions and systems

    International Nuclear Information System (INIS)

    Prassinos, P.G.; Moore, D.L.; Amico, P.J.

    1987-01-01

    The results from the review of the seven utility-sponsored seismic PRAs plus the Zion SSMRP have been used to develop some insights regarding the importance of various systems and functions to seismic margins. By taking this information and combining it with the fragility insights we can develop some functional/systemic screening guideline for margin studies. This screening approach will greatly reduce the scope of the analysis. It is possible only to come to conclusions regarding the importance of plant systems and safety functions for PWRs, for which six plants were studied. For PWRs, it is possible to categorize plant safety functions as belonging to one of two groups, one of which is important to the assessment of seismic margins and one of which is not. The important functional group involves only two functions that must be considered for estimating seismic margin. These two functions are shutting down the nuclear reaction and providing cooling to the reactor core in the time period immediately following the seismic event (that is, the injection phase or pre-residual heat removal time period). It is possible to reasonably estimate the seismic margin of the plant by performing a study only involving the analysis of the plant systems and structure which are required in order to perform the two functions. Such analysis must include an assessment of a complete set of seismic initiating events. (orig./HP)

  6. Seismic Dynamic Damage Characteristics of Vertical and Batter Pile-supported Wharf Structure Systems

    Directory of Open Access Journals (Sweden)

    Li Jiren

    2015-10-01

    Full Text Available Considering a typical steel pipe pile-supported wharf as the research object, finite element analytical models of batter and vertical pile structures were established under the same construction site, service, and geological conditions to investigate the seismic dynamic damage characteristics of vertical and batter pile-supported wharf structures. By the numerical simulation and the nonlinear time history response analysis of structure system and the moment–axial force relation curve, we analyzed the dynamic damage characteristics of the two different structures of batter and vertical piles under different seismic ground motions to provide reasonable basis and reference for designing and selecting a pile-supported wharf structure. Results showed that the axial force of batter piles was dominant in the batter pile structure and that batter piles could effectively bear and share seismic load. Under the seismic ground motion with peak ground acceleration (PGA of 350 Gal and in consideration of the factors of the design requirement of horizontal displacement, the seismic performance of the batter pile structure was better than that of the vertical pile structure. Under the seismic ground motion with a PGA of 1000 Gal, plastic failure occurred in two different structures. The contrastive analysis of the development of plastic damage and the absorption and dissipation for seismic energy indicated that the seismic performance of the vertical pile structure was better than that of the batter pile structure.

  7. 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

  8. 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

  9. 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

  10. 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.

  11. GTOROTO: a simulation system for HTGR core seismic behavior

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Nakamura, Yasuhiro; Onuma, Yoshio

    1980-07-01

    One of the most important design of HTGR core is its aseismic structure. Therefore, it is necessary to predict the forces and motion of the core blocks. To meet the requirement, many efforts to develop analytical methods and computer programs are made. A graphic simulation system GTOROTO with a CRT graphic display and lightpen was developed to analyze the HTGR core behavior in seismic excitation. Feature of the GTOROTO are as follows: (1) Behavior of the block-type HTGR core during earthquake can be shown on the CRT-display. (2) Parameters of the computing scheme can be changed with the lightpen. (3) Routines of the computing scheme can be changed with the lightpen and an alteration switch. (4) Simulation pictures are shown automatically. Hardcopies are available by plotter in stopping the progress of simulation pictures. Graphic representation can be re-start with the predetermined program. (5) Graphic representation informations can be stored in assembly language on a disk for rapid representation. (6) A computer-generated cinema can be made by COM (Computer Output Microfilming) or filming directly the CRT pictures. These features in the GTOROTO are provided in on-line conversational mode. (author)

  12. Seismic risk control of nuclear power plants using seismic protection systems in stable continental regions: The UK case

    Energy Technology Data Exchange (ETDEWEB)

    Medel-Vera, Carlos, E-mail: cbmedel@uc.cl; Ji, Tianjian, E-mail: tianjian.ji@manchester.ac.uk

    2016-10-15

    Highlights: • Strategies to reduce seismic risk for nuclear power stations in the UK are analysed. • Efficiency of devices to reduce risk: viscous-based higher than hysteretic-based. • Scenario-based incremental dynamic analysis is introduced for use in nuclear stations. • Surfaces of seismic unacceptable performance for nuclear stations are proposed. - Abstract: This article analyses three different strategies on the use of seismic protection systems (SPS) for nuclear power plants (NPPs) in the UK. Such strategies are based on the experience reported elsewhere of seismically protected nuclear reactor buildings in other stable continental regions. Analyses are conducted using an example of application based on a 1000 MW Pressurised Water Reactor building located in a representative UK nuclear site. The efficiency of the SPS is probabilistically assessed to achieve possible risk reduction for both rock and soil sites in comparison with conventionally constructed NPPs. Further analyses are conducted to study how the reduction of risk changes when all controlling scenarios of the site are included. This is done by introducing a scenario-based incremental dynamic analysis aimed at the generation of surfaces for unacceptable performance of NPPs as a function of earthquake magnitude (M{sub w}) and distance-to-site (R{sub epi}). General guidelines are proposed to potentially use SPS in future NPPs in the UK. Such recommendations can be used by the British nuclear industry in the future development of 12 new reactors to be built in the next two decades to generate 16 GWe of new nuclear capacity.

  13. Quantitative identification and analysis of sub-seismic extensional structure system: technique schemes and processes

    International Nuclear Information System (INIS)

    Chenghua, Ou; Chen, Wei; Ma, Zhonggao

    2015-01-01

    Quantitative characterization of complex sub-seismic extensional structure system that essentially controls petroleum exploitation is difficult to implement in seismic profile interpretation. This research, based on a case study in block M of Myanmar, established a set of quantitative treatment schemes and technique processes for the identification of sub-seismic low-displacement (SSLD) extensional faults or fractures upon structural deformation restoration and geometric inversion. Firstly, the master-subsidiary inheritance relations and configuration of the seismic-scale extensional fault systems are determined by analyzing the structural pattern. Besides, three-dimensional (3D) pattern and characteristics of the seismic-scale extensional structure have been illustrated by a 3D structure model built upon seismic sections. Moreover, according to the dilatancy obtained from structural restoration on the basis of inclined shear method, as well as the fracture-flow index, potential SSLD extensional faults or fractures have been quantitatively identified. Application of the technique processes to the sub-seismic low-displacement extensional structures in block M in Myanmar is instructive to quantitatively interpret those SSLD extensional structure systems in practice. (paper)

  14. The passive seismic aftershock Monitoring system: testing program and preliminary results

    International Nuclear Information System (INIS)

    Mokhtari, M.

    2005-01-01

    The paper is dedicated to testing program (phase of the passive seismic aftershock monitoring system with RefTek equipment (Refraction Technology, Inc., USA) for On-Site Inspection purposes that was carried out near Vienna International Centre in 2000. Equipment and applied software are described. Testing results were analyzed; in particular, least needs in maintenance personnel during operation. Development perspectives of passive seismic aftershock monitoring system for On-Site Inspection have been discussed. (author)

  15. Intelligent seismic risk mitigation system on structure building

    Science.gov (United States)

    Suryanita, R.; Maizir, H.; Yuniorto, E.; Jingga, H.

    2018-01-01

    Indonesia located on the Pacific Ring of Fire, is one of the highest-risk seismic zone in the world. The strong ground motion might cause catastrophic collapse of the building which leads to casualties and property damages. Therefore, it is imperative to properly design the structural response of building against seismic hazard. Seismic-resistant building design process requires structural analysis to be performed to obtain the necessary building responses. However, the structural analysis could be very difficult and time consuming. This study aims to predict the structural response includes displacement, velocity, and acceleration of multi-storey building with the fixed floor plan using Artificial Neural Network (ANN) method based on the 2010 Indonesian seismic hazard map. By varying the building height, soil condition, and seismic location in 47 cities in Indonesia, 6345 data sets were obtained and fed into the ANN model for the learning process. The trained ANN can predict the displacement, velocity, and acceleration responses with up to 96% of predicted rate. The trained ANN architecture and weight factors were later used to build a simple tool in Visual Basic program which possesses the features for prediction of structural response as mentioned previously.

  16. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    International Nuclear Information System (INIS)

    Park, Jong-beom; Park, No-Cheol; Lee, Sang-Jeong; Park, Young-Pil; Choi, Youngin

    2017-01-01

    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  17. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-beom [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Park, No-Cheol, E-mail: pnch@yonsei.ac.kr [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Lee, Sang-Jeong; Park, Young-Pil [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Choi, Youngin [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 34142 (Korea, Republic of)

    2017-03-15

    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  18. Seismic analysis of the reactor coolant system of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Borsoi, L.; Sollogoub, P.

    1986-01-01

    For safety considerations, seismic analyses are performed of the Reactor Coolant System (R.C.S.) of PWR Plants. After a brief description of the R.C.S. and R.C.S. operation, the paper presents the two types of analysis used to determine the effect of earthquake on the R.C.S.: modal spectral analysis and nonlinear time history analysis. The paper finally shows how seismic loadings are combined with other types of loadings and illustrates how the consideration of seismic loads affects R.C.S. design [fr

  19. Seismic design of equipment and piping systems for nuclear power plants in Japan

    International Nuclear Information System (INIS)

    Minematsu, Akiyoshi

    1997-01-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on 'Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981' (referred to as 'Examination Guide' hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in 'Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association'. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  20. Seismic design of equipment and piping systems for nuclear power plants in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Minematsu, Akiyoshi [Tokyo Electric Power Co., Inc. (Japan)

    1997-03-01

    The philosophy of seismic design for nuclear power plant facilities in Japan is based on `Examination Guide for Seismic Design of Nuclear Power Reactor Facilities: Nuclear Power Safety Committee, July 20, 1981` (referred to as `Examination Guide` hereinafter) and the present design criteria have been established based on the survey of governmental improvement and standardization program. The detailed design implementation procedure is further described in `Technical Guidelines for Aseismic Design of Nuclear Power Plants, JEAG4601-1987: Japan Electric Association`. This report describes the principles and design procedure of the seismic design of equipment/piping systems for nuclear power plant in Japan. (J.P.N.)

  1. 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

  2. Sensitivity analyses of seismic behavior of spent fuel dry cask storage systems

    International Nuclear Information System (INIS)

    Luk, V.K.; Spencer, B.W.; Shaukat, S.K.; Lam, I.P.; Dameron, R.A.

    2003-01-01

    Sandia National Laboratories is conducting a research project to develop a comprehensive methodology for evaluating the seismic behavior of spent fuel dry cask storage systems (DCSS) for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission (NRC). A typical Independent Spent Fuel Storage Installation (ISFSI) consists of arrays of free-standing storage casks resting on concrete pads. In the safety review process of these cask systems, their seismically induced horizontal displacements and angular rotations must be quantified to determine whether casks will overturn or neighboring casks will collide during a seismic event. The ABAQUS/Explicit code is used to analyze three-dimensional coupled finite element models consisting of three submodels, which are a cylindrical cask or a rectangular module, a flexible concrete pad, and an underlying soil foundation. The coupled model includes two sets of contact surfaces between the submodels with prescribed coefficients of friction. The seismic event is described by one vertical and two horizontal components of statistically independent seismic acceleration time histories. A deconvolution procedure is used to adjust the amplitudes and frequency contents of these three-component reference surface motions before applying them simultaneously at the soil foundation base. The research project focused on examining the dynamic and nonlinear seismic behavior of the coupled model of free-standing DCSS including soil-structure interaction effects. This paper presents a subset of analysis results for a series of parametric analyses. Input variables in the parametric analyses include: designs of the cask/module, time histories of the seismic accelerations, coefficients of friction at the cask/pad interface, and material properties of the soil foundation. In subsequent research, the analysis results will be compiled and presented in nomograms to highlight the sensitivity of seismic response of DCSS to

  3. Statistical Analysis and ETAS Modeling of Seismicity Induced by Production of Geothermal Energy from Hydrothermal Systems

    Science.gov (United States)

    Dinske, C.; Langenbruch, C.; Shapiro, S. A.

    2017-12-01

    We investigate seismicity related to hydrothermal systems in Germany and Italy, focussing on temporal changes of seismicity rates. Our analysis was motivated by numerical simulations The modeling of stress changes caused by the injection and production of fluid revealed that seismicity rates decrease on a long-term perspective which is not observed in the considered case studies. We analyze the waiting time distributions of the seismic events in both time domain (inter event times) and fluid volume domain (inter event volume). We find clear indications that the observed seismicity comprises two components: (1) seismicity that is directly triggered by production and re-injection of fluid, i.e. induced events, and (2) seismicity that is triggered by earthquake interactions, i.e. aftershock triggering. In order to better constrain our numerical simulations using the observed induced seismicity we apply catalog declustering to seperate the two components. We use the magnitude-dependent space-time windowing approach introduced by Gardner and Knopoff (1974) and test several published algorithms to calculate the space-time windows. After declustering, we conclude that the different hydrothermal reservoirs show a comparable seismic response to the circulation of fluid and additional triggering by earthquake interactions. The declustered catalogs contain approximately 50 per cent of the number of events in the original catalogs. We then perform ETAS (Epidemic Type Aftershock; Ogata, 1986, 1988) modeling for two reasons. First, we want to know whether the different reservoirs are also comparable regarding earthquake interaction patterns. Second, if we identify systematic patterns, ETAS modeling can contribute to forecast seismicity during production of geothermal energy. We find that stationary ETAS models cannot accurately capture real seismicity rate changes. One reason for this finding is given by the rate of observed induced events which is not constant over time. Hence

  4. Integrating Social impacts on Health and Health-Care Systems in Systemic Seismic Vulnerability Analysis

    Science.gov (United States)

    Kunz-Plapp, T.; Khazai, B.; Daniell, J. E.

    2012-04-01

    This paper presents a new method for modeling health impacts caused by earthquake damage which allows for integrating key social impacts on individual health and health-care systems and for implementing these impacts in quantitative systemic seismic vulnerability analysis. In current earthquake casualty estimation models, demand on health-care systems is estimated by quantifying the number of fatalities and severity of injuries based on empirical data correlating building damage with casualties. The expected number of injured people (sorted by priorities of emergency treatment) is combined together with post-earthquake reduction of functionality of health-care facilities such as hospitals to estimate the impact on healthcare systems. The aim here is to extend these models by developing a combined engineering and social science approach. Although social vulnerability is recognized as a key component for the consequences of disasters, social vulnerability as such, is seldom linked to common formal and quantitative seismic loss estimates of injured people which provide direct impact on emergency health care services. Yet, there is a consensus that factors which affect vulnerability and post-earthquake health of at-risk populations include demographic characteristics such as age, education, occupation and employment and that these factors can aggravate health impacts further. Similarly, there are different social influences on the performance of health care systems after an earthquake both on an individual as well as on an institutional level. To link social impacts of health and health-care services to a systemic seismic vulnerability analysis, a conceptual model of social impacts of earthquakes on health and the health care systems has been developed. We identified and tested appropriate social indicators for individual health impacts and for health care impacts based on literature research, using available European statistical data. The results will be used to

  5. Seismic evaluation of a diesel generator system at the Savannah River Site using earthquake experience data

    International Nuclear Information System (INIS)

    Griffin, M.J.; Tong, Wen H.; Rawls, G.B.

    1990-01-01

    New equipment and systems have been seismically qualified traditionally by either two methods, testing or analysis. Testing programs are generally expensive and their input loadings are conservative. It is generally recognized that standard seismic analysis techniques produce conservative results. Seismic loads and response levels for equipment that are typically calculated exceed the values actually experienced in earthquakes. An alternate method for demonstrating the seismic adequacy of equipment has been developed which is based on conclusions derived from studying the performance of equipment that has been subjected to actual earthquake excitations. The conclusion reached from earthquake experience data is that damage or malfunction to most types of equipment subjected to earthquakes is less than that predicted by traditional testing and analysis techniques. The use of conclusions derived from experience data provides a realistic approach in assessing the seismic ruggedness of equipment. By recognizing the inherently higher capacity that exists in specific classes of equipment, commercial ''off-the-shelf'' equipment can be procured and qualified without the need to perform expensive modifications to meet requirements imposed by traditional conservative qualification analyses. This paper will present the seismic experience data methodology applied to demonstrate the seismic adequacy of several commercially supplied 800KW diesel powered engine driven generator sets with peripheral support components installed at the Savannah River Site (SRS)

  6. First experience concerning the seismic behavior of an electric power system in eastern North America

    International Nuclear Information System (INIS)

    Pierre, J.R.

    1991-01-01

    The November 25, 1988, Saguenay earthquake of magnitude M b L g = 6.5 occurred in the province of Quebec, Canada. It represents the first strong event in eastern North America for which the seismic behavior of a power system is documented. The paper describes the seismic performance of the main components of the power system with emphasis on damages to the substation's equipment and on the triggering of control and protection devices by the seismic waves. Performance of the network is analyzed taking in account the seismological and strong ground motion features. Attention is drawn to general observations related to soil conditions and topographical relief. These data, when extrapolated to the eastern North American context, indicate that caution must be exercised concerning the seismic resistance of lifelines in eastern Canada and United States

  7. User's manual of SECOM2: a computer code for seismic system reliability analysis

    International Nuclear Information System (INIS)

    Uchiyama, Tomoaki; Oikawa, Tetsukuni; Kondo, Masaaki; Tamura, Kazuo

    2002-03-01

    This report is the user's manual of seismic system reliability analysis code SECOM2 (Seismic Core Melt Frequency Evaluation Code Ver.2) developed at the Japan Atomic Energy Research Institute for systems reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as: Calculation of component failure probabilities based on the response factor method, Extraction of minimal cut sets (MCSs), Calculation of conditional system failure probabilities for given seismic motion levels at the site of an NPP, Calculation of accident sequence frequencies and the core damage frequency (CDF) with use of the seismic hazard curve, Importance analysis using various indicators, Uncertainty analysis, Calculation of the CDF taking into account the effect of the correlations of responses and capacities of components, and Efficient sensitivity analysis by changing parameters on responses and capacities of components. These analyses require the fault tree (FT) representing the occurrence condition of the system failures and core damage, information about response and capacity of components and seismic hazard curve for the NPP site as inputs. This report presents the models and methods applied in the SECOM2 code and how to use those functions. (author)

  8. 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

  9. Determination of seismic performance factors for CLT shear wall systems

    Science.gov (United States)

    M. Omar Amini; John W. van de Lindt; Douglas Rammer; Shiling Pei; Philip Line; Marjan Popovski

    2016-01-01

    This paper presents selected results of connector testing and wall testing which were part of a Forest Products Lab-funded project undertaken at Colorado State University in an effort to determine seismic performance factors for cross laminated timber (CLT) shear walls in the United States. Archetype development, which is required as part of the process, is also...

  10. Seismic retrofit system for single leaf masonry buildings in Groningen

    NARCIS (Netherlands)

    Türkmen, Ö.S.; Vermeltfoort, A.T.; Martens, D.R.W.

    2016-01-01

    Due to recent seismic activity in the Netherlands, the demand of adequate strengthening and retrofitting techniques increased, especially for single leaf masonry. Two Dutch companies founded in the re-gion have initiated an experimental program to study the applicability of existing stand-alone

  11. Implementation of a Seismic Early Warning System in Portugal Mainland

    Science.gov (United States)

    Madureira, Guilherme; Carrilho, Fernando

    2017-04-01

    Portugal mainland is located near the border between the Eurasian and Nubian plates, whose interaction is the main responsible for a significant seismic activity in the area, with historical occurrence of several catastrophic events (e.g. Lisbon 1755 earthquake [Mag 8.7]), most of which haviguilhng epicenter rise in submerged area, located in the Cadiz Gulf and Southwest of San Vincent Cape. Early Warning Systems (EEWS) is presently a very effective concept to be applied in the mitigation of the effects caused by large earthquakes. For the mentioned area a feasibility study of a EEWS was made in the ALERT-ES project. It was found that the system could be effective to protect cities and infrastructures located at larger distances (ex: Lisbon) from the areas, located south and southwest of PT mainland, where the larger earthquakes are expected to be originated. Considering the use of a new strong-motion network recently implemented in the south of PT mainland, we concluded that the lead-times could be improved. We opted by the implementation of the well known computational platform PRESTO. In the adaptation of the mentioned platform to the local reality one of the challenges was the computation of fast moment magnitude estimates, because regional attenuation must be properly considered, and a specific study was made on this issue. The several simulations that were performed showed a reasonably good performance of the system, both on magnitude evaluation and epicentre location. However we also noted that the problems in the acquisition instruments are a very important source of disturbance in the performance of the EEWS, pointing to a need of a very accurate quality control of the strong-motion network. Considering end-users, we are also developing specific software for intensity estimation at the target places and to trigger visual and audio alerts in accordance to the expected level of shaking. This work is supported by the EU project TSUMAPS-NEAM, Agreement Number

  12. Simplified seismic analysis applied to structures systems and components with limited radioactive inventories

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1989-01-01

    This paper presents a review of the current status of simplified methods of seismic design and analysis applicable to nuclear facility structures, systems and components important to public health and safety. In particular, the International Atomic Energy Agency, IAEA TEC DOC 348 procedure for structures and the Bounding Spectra Concept for equipment as being developed by Seismic Qualification Utility Group and the Electric Power Research Institute will be discussed in some detail

  13. Expert system GIP-WWER for verification of seismic adequacy of WWER equipment

    International Nuclear Information System (INIS)

    Masopust, R.

    1999-01-01

    The aim of this report is to describe the modified Generic Implementation Procedure (GIP) titled GIP-WWER which can be used to verify seismic adequacy of the safe shutdown mechanical and electrical equipment and distribution systems of operating or constructed WWER NPPs, namely WWER-440/213 type. The WWER-GIP procedure was prepared using available information contained in GIP and the experience taken from various seismic inspections and evaluations of WWER type NPPs performed in the last five years

  14. Development of seismic damage assessment system for nuclear power plant structures in Korea

    International Nuclear Information System (INIS)

    Hyun, Chang-Hun; Lee, Sung-Kyu; Choi, Kang-Ryoung; Koh, Hyun-Moo; Cho, HoHyun

    2003-01-01

    A seismic damage assessment system that analyses in real-time the actual seismic resistance capacity and the damage level of power plant structures has been developed. The system consists of three parts: a 3-D inelastic seismic analysis, a damage assessment using a damage index based on the previous 3-D analysis, and a 3-D graphic representation. PSC containment structures are modelled by finite shell elements using layered method and analysis is performed by means of time history inelastic seismic analysis method, which takes into account material nonlinearities. HHT-α, one kind of direct integration method, is adopted for the seismic analysis. Two damage indices at finite element and structural levels are applied for the seismic damage assessment. 3-D graphical representation of dynamic responses and damage index expedites procedure for evaluating the damage level. The developed system is now being installed at the Earthquake Monitoring Center of KINS (Korea Institute of Nuclear Safety) to support site inspections after an earthquake occurrence, and decisions about effective emergency measures, repair and operations of the plant. (author)

  15. Evaluation of seismic margins for an in-plant piping system

    International Nuclear Information System (INIS)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

    1991-01-01

    Earthquake experience as well as experiments indicate that, in general, piping systems are quite rugged in resisting seismic loadings. Therefore there is a basis to hold that the seismic margin against pipe failure is very high for systems designed according to current practice. However, there is very little data, either from tests or from earthquake experience, on the actual margin or excess capacity (against failure from seismic loading) of in-plant piping systems. Design of nuclear power plant piping systems in the US is governed by the criteria given in the ASME Boiler and Pressure Vessel (B ampersand PV) Code, which assure that pipe stresses are within specified allowable limits. Generally linear elastic analytical methods are used to determine the stresses in the pipe and forces in pipe supports. The objective of this study is to verify that piping designed according to current practice does indeed have a large margin against failure and to quantify the excess capacity for piping and dynamic pipe supports on the basis of data obtained in a series of high-level seismic experiments (designated SHAM) on an in-plant piping system at the HDR (Heissdampfreaktor) Test Facility in Germany. Note that in the present context, seismic margin refers to the deterministic excess capacities of piping or supports compared to their design capacities. The excess seismic capacities or margins of a prototypical in-plant piping system and its components are evaluated by comparing measured inputs and responses from high-level simulated seismic experiments with design loads and allowables. Large excess capacities are clearly demonstrated against pipe and overall system failure with the lower bound being about four. For snubbers the lower bound margin is estimated at two and for rigid strut supports at five. 4 refs., 2 figs., 2 tabs

  16. 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.

  17. Evaluation of Seismic Behavior of Steel Braced Frames with Controlled Rocking System and Energy Dissipating Fuses

    Directory of Open Access Journals (Sweden)

    Hassan Amirzehni

    2016-12-01

    Full Text Available The self-centering rocking steel braced frames are new type of seismic lateral-force resisting systems that are developed with aim to limiting structural damages, minimizing residual drifts on systems and creating easy and inexpensive reconstruction capability, after sever earthquakes. In Steel braced frames with controlled rocking system, column bases on seismic resisting frame are not attached to the foundation and the frame allowed to rock freely. The task of restoring the rotated frame to its initial location is on post-tensioned cables, which attaches top of the frame to foundation. The design of post tensioned stands and braced frame members is such that during earthquakes they remain in elastic region. Seismic energy, dissipates by plastic deformations in replaceable elements on each rock of frame. In current research work, the seismic behavior of this type of lateral resisting systems is evaluated. The research conducted on a one bay steel braced frame with controlled rocking system that is analyzed using nonlinear dynamic time history analysis (NLTHA procedure. The frame is subjected to JMA-Kobe and Northridge ground motions records that are scaled to unit, 1.2 and 1.5 times of maximum considered earthquake (MCE ground motion level intensity. Extracted results show that seismic behavior of this type of lateral force resisting systems are so desirable even under MCE ground motion levels. The only anxiety is about occurring fatigue in post-tensioned strands that endangers overall stability of system.

  18. Development of a 3-dimensional seismic isolation floor for computer systems

    International Nuclear Information System (INIS)

    Kurihara, M.; Shigeta, M.; Nino, T.; Matsuki, T.

    1991-01-01

    In this paper, we investigated the applicability of a seismic isolation floor as a method for protecting computer systems from strong earthquakes, such as computer systems in nuclear power plants. Assuming that the computer system is guaranteed for 250 cm/s 2 of input acceleration in the horizontal and vertical directions as the seismic performance, the basic design specification of the seismic isolation floor is considered as follows. Against S 1 level earthquakes, the maximum acceleration response of the seismic isolation floor in the horizontal and vertical directions is kept less than 250 cm/s 2 to maintain continuous computer operation. Against S 2 level earthquakes, the isolation floor allows large horizontal movement and large displacement of the isolation devices to reduce the acceleration response, although it is not guaranteed to be less than 250 cm/s 2 . By reducing the acceleration response, however, serious damage to the computer systems is reduced, so that they can be restarted after an earthquake. Usually, seismic isolation floor systems permit 2-dimensional (horizontal) isolation. However, in the case of just-under-seated earthquakes, which have large vertical components, the vertical acceleration response of this system is amplified by the lateral vibration of the frame of the isolation floor. Therefore, in this study a 3-dimensional seismic isolation floor, including vertical isolation, was developed. This paper describes 1) the experimental results of the response characteristics of the 3-dimensional seismic isolation floor built as a trial using a 3-dimensional shaking table, and 2) comparison of a 2-dimensional analytical model, for motion in one horizontal direction and the vertical direction, to experimental results. (J.P.N.)

  19. 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

  20. Testing, licensing, and code requirements for seismic isolation systems (for nuclear power plants)

    International Nuclear Information System (INIS)

    Seidensticker, R.W.

    1987-01-01

    The use of seismic isolation as an earthquake hazard mitigation strategy for nuclear reactor power plants is rapidly receiving interest throughout the world. Seismic isolation has already been used on at least two French PWR plants, was to have been used for plants to be built in Iran, and is under serious consideration for advanced LMR plants (in the US, UK, France, and Japan). In addition, there is a growing use of seismic isolation throughout the world for other critical facilities such as hospitals, emergency facilities, buildings with very high-cost equipment (e.g., computers) and as a strategy to reduce loss of life and expensive equipment in earthquakes. Such a design approach is in complete contrast to the conventional seismic design strategy in which the structure and components are provided with sufficient strength and ductility to resist the earthquake forces and to prevent structural collapses or failure. The use of seismic isolation for nuclear plants can, therefore, be expected to be a significant licensing issue. For isolation, the licensing process must shift away in large measure from the superstructure and concentrate on the behavior of the seismic isolation system. This paper is not intended to promote the advantages of seismic isolation system, but to explore in some detail those technical issues which must be satisfactorily addressed to achieve full licensability of the use of seismic isolation as a viable, attractive and economical alternative to current traditional design approaches. Special problems and topics associated with testing and codes and standards development are addressed. A positive program for approach or strategy to secure licensing is presented

  1. Testing, licensing, and code requirements for seismic isolation systems (for nuclear power plants)

    Energy Technology Data Exchange (ETDEWEB)

    Seidensticker, R.W.

    1987-01-01

    The use of seismic isolation as an earthquake hazard mitigation strategy for nuclear reactor power plants is rapidly receiving interest throughout the world. Seismic isolation has already been used on at least two French PWR plants, was to have been used for plants to be built in Iran, and is under serious consideration for advanced LMR plants (in the US, UK, France, and Japan). In addition, there is a growing use of seismic isolation throughout the world for other critical facilities such as hospitals, emergency facilities, buildings with very high-cost equipment (e.g., computers) and as a strategy to reduce loss of life and expensive equipment in earthquakes. Such a design approach is in complete contrast to the conventional seismic design strategy in which the structure and components are provided with sufficient strength and ductility to resist the earthquake forces and to prevent structural collapses or failure. The use of seismic isolation for nuclear plants can, therefore, be expected to be a significant licensing issue. For isolation, the licensing process must shift away in large measure from the superstructure and concentrate on the behavior of the seismic isolation system. This paper is not intended to promote the advantages of seismic isolation system, but to explore in some detail those technical issues which must be satisfactorily addressed to achieve full licensability of the use of seismic isolation as a viable, attractive and economical alternative to current traditional design approaches. Special problems and topics associated with testing and codes and standards development are addressed. A positive program for approach or strategy to secure licensing is presented.

  2. Self-Centering Seismic Lateral Force Resisting Systems: High Performance Structures for the City of Tomorrow

    Directory of Open Access Journals (Sweden)

    Nathan Brent Chancellor

    2014-09-01

    Full Text Available Structures designed in accordance with even the most modern buildings codes are expected to sustain damage during a severe earthquake; however; these structures are expected to protect the lives of the occupants. Damage to the structure can require expensive repairs; significant business downtime; and in some cases building demolition. If damage occurs to many structures within a city or region; the regional and national economy may be severely disrupted. To address these shortcomings with current seismic lateral force resisting systems and to work towards more resilient; sustainable cities; a new class of seismic lateral force resisting systems that sustains little or no damage under severe earthquakes has been developed. These new seismic lateral force resisting systems reduce or prevent structural damage to nonreplaceable structural elements by softening the structural response elastically through gap opening mechanisms. To dissipate seismic energy; friction elements or replaceable yielding energy dissipation elements are also included. Post-tensioning is often used as a part of these systems to return the structure to a plumb; upright position (self-center after the earthquake has passed. This paper summarizes the state-of-the art for self-centering seismic lateral force resisting systems and outlines current research challenges for these systems.

  3. Comparative performance of passive devices for piping system under seismic excitation

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Praveen, E-mail: pra_veen74@rediffmail.com [Bhabha Atomic Research Centre, Trombay, Mumbai, 400085 (India); Jangid, R.S. [Department of Civil Engineering, Indian Institute of Technology Bombay, Powai, Mumbai, 400076 (India); Reddy, G.R. [Bhabha Atomic Research Centre, Trombay, Mumbai, 400085 (India)

    2016-03-15

    Highlights: • Correlated the analytical results obtained from the proposed analytical procedures with experimental results in the case of XPD. • Substantial reduction of the seismic response of piping system with passive devices is observed. • Significant increase in the modal damping of the piping system is noted. • There exist an optimum parameters of the passive devices. • Good amount of energy dissipation is observed by using passive devices. - Abstract: Among several passive control devices, X-plate damper, viscous damper, visco-elastic damper, tuned mass damper and multiple tuned mass dampers are popular and used to mitigate the seismic response in the 3-D piping system. In the present paper detailed studies are made to see the effectiveness of the dampers when used in 3-D piping system subjected to artificial earthquake with increasing amplitudes. The analytical results obtained using Wen's model are compared with the corresponding experimental results available which indicated a good match with the proposed analytical procedure for the X-plate dampers. It is observed that there is significant reduction in the seismic response of interest like relative displacement, acceleration and the support reaction of the piping system with passive devices. In general, the passive devices under particular optimum parameters such as stiffness and damping are very effective and practically implementable for the seismic response mitigation, vibration control and seismic requalification of piping system.

  4. Active seismic response control systems for nuclear power plant equipment facilities

    International Nuclear Information System (INIS)

    Kobori, Takuji; Kanayama, Hiroo; Kamagata, Shuichi

    1989-01-01

    To sustain severe earthquake ground motion, a new type of anti-seismic structure is proposed, called a Dynamic Intelligent Building (DIB) system, which is positioned as an active seismic response controlled the structure. The structural concept starts from a new recognition of earthquake ground motion, and the structural natural frequency is actively adjusted to avoid resonant vibration, and similarly the external counter-force cancels the resonant force which comes from the dynamic structural motion energy. These concepts are verified using an analytical simulator program. The advanced application of the DIB system, is the Active Supporting system and the Active Stabilizer system for nuclear power plant equipment facilities. (orig.)

  5. Development of Friction Pendulum System to Reduce the Seismic Force

    International Nuclear Information System (INIS)

    Jang, J. B.; Kim, J. K.; Hwang, K. M.; Kwon, H. O.; Lee, C. W

    2007-01-01

    Most of the damages in electrical facilities are occurred at transformer due to the great earthquake. The damage types of transformer are the failure of upper bushing of transformer, overturning of transformer due to geometry with high height, and the failure of anchorage of transformer. The objective of this study is to develop the seismic isolator to prevent the damage of transformer due to earthquake considering the importance of transformer

  6. Seismic qualification of piping systems based on strain criteria

    International Nuclear Information System (INIS)

    Peters, K.; Rangette, A.

    1988-01-01

    Typical LMFBR piping is characterized by elevated temperature and low pressure levels. Taking into account operational conditions only these characteristics demand for and allow flexible piping design. The overestimation of the damage potential of seismic loading by e.g. improper failure criteria usually contradicts operational needs producing the known result of excessive ''snubberism'' and reduction of operational margins. As a matter of fact, due to its transiency seismic loading is essentially secondary provoking the natural design requirement ductility instead of stiffness and rigidity - i.e. exclusion of failure by strain control instead of stress control - and thus avoiding the LMFBR typical competition between operational needs and seismic qualification. The design requirement ductility needs judgement mechanisms, i.e. suitable load descriptions, allowed strain levels and strain evaluation tools. A simplified method for strain range estimation and the underlying basic ideas are roughly outlined. The status of verification and experience gained so far is described. The results achieved suggest that the qualification of piping based on ductility requirement controlled by strain criteria is not out of reach. (author)

  7. Analysis of Bi-directional Effects on the Response of a Seismic Base Isolation System

    International Nuclear Information System (INIS)

    Park, Hyung-Kui; Kim, Jung-Han; Kim, Min Kyu; Choi, In-Kil

    2014-01-01

    The floor response spectrum depends on the height of the floor of the structure. Also FRS depends on the characteristics of the seismic base isolation system such as the natural frequency, damping ratio. In the previous study, the floor response spectrum of the base isolated structure was calculated for each axis without considering bi-directional effect. However, the shear behavior of the seismic base isolation system of two horizontal directions are correlated each other by the bi-directional effects. If the shear behavior of the seismic isolation system changes, it can influence the floor response spectrum and displacement response of isolators. In this study, the analysis of a bi-directional effect on the floor response spectrum was performed. In this study, the response of the seismic base isolation system based on the bi-directional effects was analyzed. By analyzing the time history result, while there is no alteration in the maximum shear force of seismic base isolation system, it is confirmed that the shear force is generally more decreased in a one-directional that in a two-directional in most parts. Due to the overall decreased shear force, the floor response spectrum is more reduced in a two-directional than in a one-directional

  8. Study on seismic design margin based upon inelastic shaking test of the piping and support system

    International Nuclear Information System (INIS)

    Ishiguro, Takami; Eto, Kazutoshi; Ikeda, Kazutoyo; Yoshii, Toshiaki; Kondo, Masami; Tai, Koichi

    2009-01-01

    In Japan, according to the revised Regulatory Guide for Aseismic Design of Nuclear Power Reactor Facilities, September 2006, criteria of design basis earthquakes of Nuclear Power Reactor Facilities become more severe. Then, evaluating seismic design margin took on a great importance and it has been profoundly discussed. Since seismic safety is one of the major key issues of nuclear power plant safety, it has been demonstrated that nuclear piping system possesses large safety margins by various durability test reports for piping in ultimate conditions. Though the knowledge of safety margin has been accumulated from these reports, there still remain some technical uncertainties about the phenomenon when both piping and support structures show inelastic behavior in extremely high seismic excitation level. In order to obtain the influences of inelastic behavior of the support structures to the whole piping system response when both piping and support structures show inelastic behavior, we examined seismic proving tests and we conducted simulation analyses for the piping system which focused on the inelastic behavior of the support to the whole piping system response. This paper introduces major results of the seismic shaking tests of the piping and support system and the simulation analyses of these tests. (author)

  9. Utilities/industries joint study on seismic isolation systems for LWR: Part I. Experimental and analytical studies on seismic isolation systems

    International Nuclear Information System (INIS)

    Kato, Muneaki; Sato, Shoji; Shimomura, Issei

    1989-01-01

    This paper describes a joint study program on seismic isolation systems for light-water reactors (LWRs) performed by ten electric power companies, three manufacturers, and five construction companies. The fundamental response characteristics of base-isolated structures and base-isolation devices are described. Applications of a base-isolation system to LWR buildings are given. Finally, three-dimensional shaking table experiments are described

  10. Isolation systems influence in the seismic loading propagation analysis applied to an innovative near term reactor

    International Nuclear Information System (INIS)

    Lo Frano, R.; Forasassi, G.

    2010-01-01

    Integrity of a Nuclear Power Plant (NPP) must be ensured during the plant life in any design condition and, particularly, in the event of a severe earthquake. To investigate the seismic resistance capability of as-built structures systems and components, in the event of a Safe Shutdown Earthquake (SSE), and analyse its related effects on a near term deployment reactor and its internals, a deterministic methodological approach, based on the evaluation of the propagation of seismic waves along the structure, was applied considering, also, the use of innovative anti-seismic techniques. In this paper the attention is focused on the use and influence of seismic isolation technologies (e.g. isolators based on passive energy dissipation) that seem able to ensure the full integrity and operability of NPP structures, to enhance the seismic safety (improving the design of new NPPs and if possible, to retrofit existing facilities) and to attain a standardization plant design. To the purpose of this study a numerical assessment of dynamic response/behaviour of the structures was accomplished by means of the finite element approach and setting up, as accurately as possible, a representative three-dimensional model of mentioned NPP structures. The obtained results in terms of response spectra (carried out from both cases of isolated and not isolated seismic analyses) are herein presented and compared in order to highlight the isolation technique effectiveness.

  11. New perspectives on the damage estimation for buried pipeline systems due to seismic wave propagation

    Energy Technology Data Exchange (ETDEWEB)

    Pineda Porras, Omar Andrey [Los Alamos National Laboratory

    2009-01-01

    Over the past three decades, seismic fragility fonnulations for buried pipeline systems have been developed following two tendencies: the use of earthquake damage scenarios from several pipeline systems to create general pipeline fragility functions; and, the use of damage scenarios from one pipeline system to create specific-system fragility functions. In this paper, the advantages and disadvantages of both tendencies are analyzed and discussed; in addition, a summary of what can be considered the new challenges for developing better pipeline seismic fragility formulations is discussed. The most important conclusion of this paper states that more efforts are needed to improve the estimation of transient ground strain -the main cause of pipeline damage due to seismic wave propagation; with relevant advances in that research field, new and better fragility formulations could be developed.

  12. Sequence stratigraphy, seismic stratigraphy, and seismic structures of the lower intermediate confining unit and most of the Floridan aquifer system, Broward County, Florida

    Science.gov (United States)

    Cunningham, Kevin J.; Kluesner, Jared W.; Westcott, Richard L.; Robinson, Edward; Walker, Cameron; Khan, Shakira A.

    2017-12-08

    Deep well injection and disposal of treated wastewater into the highly transmissive saline Boulder Zone in the lower part of the Floridan aquifer system began in 1971. The zone of injection is a highly transmissive hydrogeologic unit, the Boulder Zone, in the lower part of the Floridan aquifer system. Since the 1990s, however, treated wastewater injection into the Boulder Zone in southeastern Florida has been detected at three treated wastewater injection utilities in the brackish upper part of the Floridan aquifer system designated for potential use as drinking water. At a time when usage of the Boulder Zone for treated wastewater disposal is increasing and the utilization of the upper part of the Floridan aquifer system for drinking water is intensifying, there is an urgency to understand the nature of cross-formational fluid flow and identify possible fluid pathways from the lower to upper zones of the Floridan aquifer system. To better understand the hydrogeologic controls on groundwater movement through the Floridan aquifer system in southeastern Florida, the U.S. Geological Survey and the Broward County Environmental Planning and Community Resilience Division conducted a 3.5-year cooperative study from July 2012 to December 2015. The study characterizes the sequence stratigraphy, seismic stratigraphy, and seismic structures of the lower part of the intermediate confining unit aquifer and most of the Floridan aquifer system.Data obtained to meet the study objective include 80 miles of high-resolution, two-dimensional (2D), seismic-reflection profiles acquired from canals in eastern Broward County. These profiles have been used to characterize the sequence stratigraphy, seismic stratigraphy, and seismic structures in a 425-square-mile study area. Horizon mapping of the seismic-reflection profiles and additional data collection from well logs and cores or cuttings from 44 wells were focused on construction of three-dimensional (3D) visualizations of eight

  13. 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)

  14. Open Source Seismic Software in NOAA's Next Generation Tsunami Warning System

    Science.gov (United States)

    Hellman, S. B.; Baker, B. I.; Hagerty, M. T.; Leifer, J. M.; Lisowski, S.; Thies, D. A.; Donnelly, B. K.; Griffith, F. P.

    2014-12-01

    The Tsunami Information technology Modernization (TIM) is a project spearheaded by National Oceanic and Atmospheric Administration to update the United States' Tsunami Warning System software currently employed at the Pacific Tsunami Warning Center (Eva Beach, Hawaii) and the National Tsunami Warning Center (Palmer, Alaska). This entirely open source software project will integrate various seismic processing utilities with the National Weather Service Weather Forecast Office's core software, AWIPS2. For the real-time and near real-time seismic processing aspect of this project, NOAA has elected to integrate the open source portions of GFZ's SeisComP 3 (SC3) processing system into AWIPS2. To provide for better tsunami threat assessments we are developing open source tools for magnitude estimations (e.g., moment magnitude, energy magnitude, surface wave magnitude), detection of slow earthquakes with the Theta discriminant, moment tensor inversions (e.g. W-phase and teleseismic body waves), finite fault inversions, and array processing. With our reliance on common data formats such as QuakeML and seismic community standard messaging systems, all new facilities introduced into AWIPS2 and SC3 will be available as stand-alone tools or could be easily integrated into other real time seismic monitoring systems such as Earthworm, Antelope, etc. Additionally, we have developed a template based design paradigm so that the developer or scientist can efficiently create upgrades, replacements, and/or new metrics to the seismic data processing with only a cursory knowledge of the underlying SC3.

  15. Wireless acquisition of multi-channel seismic data using the Seismobile system

    Science.gov (United States)

    Isakow, Zbigniew

    2017-11-01

    This paper describes the wireless acquisition of multi-channel seismic data using a specialized mobile system, Seismobile, designed for subsoil diagnostics for transportation routes. The paper presents examples of multi-channel seismic records obtained during system tests in a configuration with 96 channels (4 landstreamers of 24-channel) and various seismic sources. Seismic waves were generated at the same point using different sources: a 5-kg hammer, a Gisco's source with a 90-kg pile-driver, and two other the pile-drivers of 45 and 70 kg. Particular attention is paid to the synchronization of source timing, the measurement of geometry by autonomous GPS systems, and the repeatability of triggering measurements constrained by an accelerometer identifying the seismic waveform. The tests were designed to the registration, reliability, and range of the wireless transmission of survey signals. The effectiveness of the automatic numbering of measuring modules was tested as the system components were arranged and fixed to the streamers. After measurements were completed, the accuracy and speed of data downloading from the internal memory (SDHC 32GB WiFi) was determined. Additionally, the functionality of automatic battery recharging, the maximum survey duration, and the reliability of battery discharge signalling were assessed.

  16. 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.

  17. An Experimental Seismic Data and Parameter Exchange System for Tsunami Warning Systems

    Science.gov (United States)

    Hoffmann, T. L.; Hanka, W.; Saul, J.; Weber, B.; Becker, J.; Heinloo, A.; Hoffmann, M.

    2009-12-01

    For several years GFZ Potsdam is operating a global earthquake monitoring system. Since the beginning of 2008, this system is also used as an experimental seismic background data center for two different regional Tsunami Warning Systems (TWS), the IOTWS (Indian Ocean) and the interim NEAMTWS (NE Atlantic and Mediterranean). The SeisComP3 (SC3) software, developed within the GITEWS (German Indian Ocean Tsunami Early Warning System) project, capable to acquire, archive and process real-time data feeds, was extended for export and import of individual processing results within the two clusters of connected SC3 systems. Therefore not only real-time waveform data are routed to the attached warning centers through GFZ but also processing results. While the current experimental NEAMTWS cluster consists of SC3 systems in six designated national warning centers in Europe, the IOTWS cluster presently includes seven centers, with another three likely to join in 2009/10. For NEAMTWS purposes, the GFZ virtual real-time seismic network (GEOFON Extended Virtual Network -GEVN) in Europe was substantially extended by adding many stations from Western European countries optimizing the station distribution. In parallel to the data collection over the Internet, a GFZ VSAT hub for secured data collection of the EuroMED GEOFON and NEAMTWS backbone network stations became operational and first data links were established through this backbone. For the Southeast Asia region, a VSAT hub has been established in Jakarta already in 2006, with some other partner networks connecting to this backbone via the Internet. Since its establishment, the experimental system has had the opportunity to prove its performance in a number of relevant earthquakes. Reliable solutions derived from a minimum of 25 stations were very promising in terms of speed. For important events, automatic alerts were released and disseminated by emails and SMS. Manually verified solutions are added as soon as they become

  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. Seismic qualification of equipment

    International Nuclear Information System (INIS)

    Heidebrecht, A.C.; Tso, W.K.

    1983-03-01

    This report describes the results of an investigation into the seismic qualification of equipment located in CANDU nuclear power plants. It is particularly concerned with the evaluation of current seismic qualification requirements, the development of a suitable methodology for the seismic qualification of safety systems, and the evaluation of seismic qualification analysis and testing procedures

  20. 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)

  1. Artificial neural systems for interpretation and inversion of seismic data

    Science.gov (United States)

    Calderon-Macias, Carlos

    The goal of this work is to investigate the feasibility of using neural network (NN) models for solving geophysical exploration problems. First, a feedforward neural network (FNN) is used to solve inverse problems. The operational characteristics of a FNN are primarily controlled by a set of weights and a nonlinear function that performs a mapping between two sets of data. In a process known as training, the FNN weights are iteratively adjusted to perform the mapping. After training, the computed weights encode important features of the data that enable one pattern to be distinguished from another. Synthetic data computed from an ensemble of earth models and the corresponding models provide the training data. Two training methods are studied: the backpropagation method which is a gradient scheme, and a global optimization method called very fast simulated annealing (VFSA). A trained network is then used to predict models from new data (e.g., data from a new location) in a one-step procedure. The application of this method to the problems of obtaining formation resistivities and layer thicknesses from resistivity sounding data and 1D velocity models from seismic data shows that trained FNNs produce reasonably accurate earth models when observed data are input to the FNNs. In a second application, a FNN is used for automating the NMO correction process of seismic reflection data. The task of the FNN is to map CMP data at control locations along a seismic line into subsurface velocities. The network is trained while the velocity analyses are performed at the control locations. Once trained, the computed weights are used as an operator that acts on the remaining CMP data as a velocity interpolator, resulting in a fast method for NMO correction. The second part of this dissertation describes the application of a Hopfield neural network (HNN) to the problems of deconvolution and multiple attenuation. In these applications, the unknown parameters (reflection coefficients

  2. 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

  3. Development and Examination of Real-time Automatic Scram System Using Deep Vertical Array Seismic Observation System

    International Nuclear Information System (INIS)

    Sugaya, Katsunori

    2014-01-01

    In Japan, observed seismic motions in reactor buildings are currently used for seismic scram, but installing a seismometer at a great depth at the site may possibly shorten scram initiation time. JNES proposed a scram system with a seismometer set at a depth of 3,000 m on the premises of the Niigata Institute of Technology based on preliminary results for a scenario earthquake and is now planning quantitative evaluation. (authors)

  4. A high-speed transmission method for large-scale marine seismic prospecting systems

    International Nuclear Information System (INIS)

    KeZhu, Song; Ping, Cao; JunFeng, Yang; FuMing, Ruan

    2012-01-01

    A marine seismic prospecting system is a kind of data acquisition and transmission system with large-scale coverage and synchronous multi-node acquisition. In this kind of system, data transmission is a fundamental and difficult technique. In this paper, a high-speed data-transmission method is proposed, its implications and limitations are discussed, and conclusions are drawn. The method we propose has obvious advantages over traditional techniques with respect to long-distance operation, high speed, and real-time transmission. A marine seismic system with four streamers, each 6000 m long and capable of supporting up to 1920 channels, was designed and built based on this method. The effective transmission baud rate of this system was found to reach up to 240 Mbps, while the minimum sampling interval time was as short as 0.25 ms. This system was found to achieve a good synchronization: 83 ns. Laboratory and in situ experiments showed that this marine-prospecting system could work correctly and robustly, which verifies the feasibility and validity of the method proposed in this paper. In addition to the marine seismic applications, this method can also be used in land seismic applications and certain other transmission applications such as environmental or engineering monitoring systems. (paper)

  5. An innovative seismic bracing system based on a superelastic shape memory alloy ring

    International Nuclear Information System (INIS)

    Gao, Nan; Jeon, Jong-Su; DesRoches, Reginald; Hodgson, Darel E

    2016-01-01

    Shape memory alloys (SMAs) have great potential in seismic applications because of their remarkable superelasticity. Seismic bracing systems based on SMAs can mitigate the damage caused by earthquakes. The current study investigates a bracing system based on an SMA ring which is capable of both re-centering and energy dissipation. This lateral force resisting system is a cross-braced system consisting of an SMA ring and four tension-only cable assemblies, which can be applied to both new construction and seismic retrofit. The performance of this bracing system is examined through a quasi-static cyclic loading test and finite element (FE) analysis. This paper describes the experimental design in detail, discusses the experimental results, compares the performance with other bracing systems based on SMAs, and presents an Abaqus FE model calibrated on the basis of experimental results to simulate the superelastic behavior of the SMA ring. The experimental results indicate that the seismic performance of this system is promising in terms of damping and re-centering. The FE model can be used in the simulation of building structures using the proposed bracing system. (paper)

  6. A high-speed transmission method for large-scale marine seismic prospecting systems

    Science.gov (United States)

    KeZhu, Song; Ping, Cao; JunFeng, Yang; FuMing, Ruan

    2012-12-01

    A marine seismic prospecting system is a kind of data acquisition and transmission system with large-scale coverage and synchronous multi-node acquisition. In this kind of system, data transmission is a fundamental and difficult technique. In this paper, a high-speed data-transmission method is proposed, its implications and limitations are discussed, and conclusions are drawn. The method we propose has obvious advantages over traditional techniques with respect to long-distance operation, high speed, and real-time transmission. A marine seismic system with four streamers, each 6000 m long and capable of supporting up to 1920 channels, was designed and built based on this method. The effective transmission baud rate of this system was found to reach up to 240 Mbps, while the minimum sampling interval time was as short as 0.25 ms. This system was found to achieve a good synchronization: 83 ns. Laboratory and in situ experiments showed that this marine-prospecting system could work correctly and robustly, which verifies the feasibility and validity of the method proposed in this paper. In addition to the marine seismic applications, this method can also be used in land seismic applications and certain other transmission applications such as environmental or engineering monitoring systems.

  7. 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.

  8. Research on database realization technology of seismic information system in CTBT verification

    International Nuclear Information System (INIS)

    Zheng Xuefeng; Shen Junyi; Zhang Huimin; Jing Ping; Sun Peng; Zheng Jiangling

    2005-01-01

    Developing CTBT verification technology has become the most important method that makes sure CTBT to be fulfilled conscientiously. The seismic analysis based on seismic information system (SIS) is playing an important rule in this field. Based on GIS, the SIS will be very sufficient and powerful in spatial analysis, topologic analysis and visualization. However, the critical issue to implement the whole system function depends on the performance of SIS DB. Based on the ArcSDE Geodatabase data model, not only have the spatial data and attribute data seamless integrated management been realized with RDBMS ORACLE really, but also the most functions of ORACLE have been reserved. (authors)

  9. Scram and nonlinear reactor system seismic analysis for the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Morrone, A.

    1975-01-01

    A description is given of the analysis and results for the Fast Flux Test Facility (FFTF) reactor system which was analyzed for both scram times and seismic responses such as bending moments and impact forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node. The results give time history plots of various seismic responses and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. (U.S.)

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. Documentation of the workshop on R and D and application of seismic emergency information system

    International Nuclear Information System (INIS)

    2000-07-01

    This report describes the summary of the workshop on R and D and application of Seismic Emergency Information System (SEIS) organized by Japan Atomic Energy Research Institute (JAERI) and National Research Institute for Earth Science and Disaster Prevention (NIED) held on December 20, 1999. Documents presented in the workshop are attached. The workshop consists of the following five technical sessions. - Session I: Objectives of Workshop, - Session II: Progress of R and D of SEIS, - Session III: Current status of application of Seismic Information Systems, - Session IV: Free-discussion for issues and future prospects of Information Systems, - Session V: Briefing, Subsequently to the introduction of objectives of the workshop, the four topics on the progress of R and D of the seismic information system were presented by JAERI and NIED. The discussions are summarized in three viewpoints, i.e.; utilization of the potential of JAERI activities to the R and D, clarification on the objectives and philosophy of the system, effective utilization of the result of R and D. In addition, the current status on the application of seismic information systems was presented by staffs of local government and etc. Issues and future prospects of the information systems were discussed. The workshop was summarized in the final session. (author)

  15. Cities at risk: status of Italian planning system in reducing seismic and hydrogeological risks

    Directory of Open Access Journals (Sweden)

    Grazia Di Giovanni

    2016-03-01

    Full Text Available Italy and its urban systems are under high seismic and hydrogeological risks. The awareness about the role of human activities in the genesis of disasters is achieved in the scientific debate, as well as the role of urban and regional planning in reducing risks. The paper reviews the state of Italian major cities referred to hydrogeological and seismic risk by: 1 extrapolating data and maps about seismic hazard and landslide risk concerning cities with more than 50.000 inhabitants and metropolitan contexts, and 2 outlining how risk reduction is framed in Italian planning system (at national and regional levels. The analyses of available data and the review of the normative framework highlight the existing gaps in addressing risk reduction: nevertheless a wide knowledge about natural risks afflicting Italian territory and an articulated regulatory framework, the available data about risks are not exhaustive, and risk reduction policies and multidisciplinary pro-active approaches are only partially fostered and applied.

  16. 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

  17. Seismic re-evaluation of piping systems of heavy water plant, Kota

    International Nuclear Information System (INIS)

    Mishra, Rajesh; Soni, R.S.; Kushwaha, H.S.; Venkat Raj, V.

    2002-05-01

    Heavy Water Plant, Kota is the first indigenous heavy water plant built in India. The plant started operation in the year 1985 and it is approaching the completion of its originally stipulated design life. In view of the excellent record of plant operation for the past so many years, it has been planned to carry out various exercises for the life extension of the plant. In the first stage, evaluation of operation stresses was carried out for the process critical piping layouts and equipment, which are connected with 25 process critical nozzle locations, identified based on past history of the plant performance. Fatigue life evaluation has been carried out to fmd out the Cumulative Usage Factor, which helps in arriving at a decision regarding the life extension of the plant. The results of these exercises have been already reported separately vide BARC/200I /E/O04. In the second stage, seismic reevaluation of the plant has been carried out to assess its ability to maintain its integ:rity in case of a seismic event. The aim of this exercise is to assess the effects of the maximum probable earthquake at the plant site on the various systems and components of the plant. This exercise is further aimed at ensuring the adequacy of seismic supports to maintain the integrity of the system in case of a seismic event and to suggest some retrofitting measures, if required. Seismic re-evaluation of the piping of Heavy Water Plant, Kota has been performed taking into account the interaction effects from the connected equipment. Each layout has been qualified using the latest provisions of ASME Code Section III, Subsection ND wherein the earthquake loading has been considered as a reversing dynamic load. The maximum combined stresses for all the layouts due to pressure, weight and seismic loadings have been found to be well within the code allowable limit. Therefore, it has been concluded that during a maximum probable seismic event, the possibility of pipe rupture can be safely

  18. Comparison of elastic and inelastic seismic response of high temperature piping systems

    International Nuclear Information System (INIS)

    Thomas, F.M.; McCabe, S.L.; Liu, Y.

    1994-01-01

    A study of high temperature power piping systems is presented. The response of the piping systems is determined when subjected to seismic disturbances. Two piping systems are presented, a main steam line, and a cold reheat line. Each of the piping systems are modeled using the ANSYS computer program and two analyses are performed on each piping system. First, each piping system is subjected to a seismic disturbance and the pipe material is assumed to remain linear and elastic. Next the analysis is repeated for each piping system when the pipe material is modeled as having elastic-plastic behavior. The results of the linear elastic analysis and elastic-plastic analysis are compared for each of the two pipe models. The pipe stresses, strains, and displacements, are compared. These comparisons are made so that the effect of the material yielding can be determined and to access what error is made when a linear analysis is performed on a system that yields

  19. Seismic and structural characterization of the fluid bypass system using 3D and partial stack seismic from passive margin: inside the plumbing system.

    Science.gov (United States)

    Iacopini, David; Maestrelli, Daniele; Jihad, Ali; Bond, Clare; Bonini, Marco

    2017-04-01

    In recent years enormous attention has been paid to the understanding of the process and mechanism controlling the gas seepage and more generally the fluid expulsion affecting the earth system from onshore to offshore environment. This is because of their demonstrated impact to our environment, climate change and during subsea drilling operation. Several example from active and paleo system has been so far characterized and proposed using subsurface exploration, geophysical and geochemical monitoring technology approaches with the aims to explore what trigger and drive the overpressure necessary maintain the fluid/gas/material expulsion and what are the structure that act as a gateway for gaseous fluid and unconsolidated rock. In this contribution we explore a series of fluid escape structure (ranging from seepage pipes to large blowout pipes structure of km length) using 3D and partial stack seismic data from two distinctive passive margin from the north sea (Loyal field, West Shetland) and the Equatorial Brazil (Ceara' Basin). We will focuses on the characterization of the plumbing system internal architecture and, for selected example, exploring the AVO response (using partial stack) of the internal fluid/unconsolidated rock. The detailed seismic mapping and seismic attributes analysis of the conduit system helped us to recover some detail from the signal response of the chimney internal structures. We observed: (1) small to medium seeps and pipes following structural or sedimentary discontinuities (2) large pipes (probably incipient mud volcanoes) and blowup structures propagating upward irrespective of pre-existing fault by hydraulic fracturing and assisted by the buoyancy of a fluidised and mobilised mud-hydrocarbon mixture. The reflector termination observed inside the main conduits, the distribution of stacked bright reflectors and the AVO analysis suggests an evolution of mechanisms (involving mixture of gas, fluid and probably mud) during pipe birth and

  20. Development of Vertical Cable Seismic System for Hydrothermal Deposit Survey (2) - Feasibility Study

    Science.gov (United States)

    Asakawa, E.; Murakami, F.; Sekino, Y.; Okamoto, T.; Mikada, H.; Takekawa, J.; Shimura, T.

    2010-12-01

    In 2009, Ministry of Education, Culture, Sports, Science and Technology(MEXT) started the survey system development for Hydrothermal deposit. We proposed the Vertical Cable Seismic (VCS), the reflection seismic survey with vertical cable above seabottom. VCS has the following advantages for hydrothermal deposit survey. . (1) VCS is an effective high-resolution 3D seismic survey within limited area. (2) It achieves high-resolution image because the sensors are closely located to the target. (3) It avoids the coupling problems between sensor and seabottom that cause serious damage of seismic data quality. (4) Various types of marine source are applicable with VCS such as sea-surface source (air gun, water gun etc.) , deep-towed or ocean bottom sources. (5) Autonomous recording system. Our first experiment of 2D/3D VCS surveys has been carried out in Lake Biwa, JAPAN. in November 2009. The 2D VCS data processing follows the walk-away VSP, including wave field separation and depth migration. The result gives clearer image than the conventional surface seismic. Prestack depth migration is applied to 3D data to obtain good quality 3D depth volume. Uncertainty of the source/receiver poisons in water causes the serious problem of the imaging. We used several transducer/transponder to estimate these positions. The VCS seismic records themselves can also provide sensor position using the first break of each trace and we calibrate the positions. We are currently developing the autonomous recording VCS system and planning the trial experiment in actual ocean to establish the way of deployment/recovery and the examine the position through the current flow in November, 2010. The second VCS survey will planned over the actual hydrothermal deposit with deep-towed source in February, 2011.

  1. 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

  2. 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

  3. Data Analysis of Seismic Sequence in Central Italy in 2016 using CTBTO- International Monitoring System

    Science.gov (United States)

    Mumladze, Tea; Wang, Haijun; Graham, Gerhard

    2017-04-01

    The seismic network that forms the International Monitoring System (IMS) of the Comprehensive Nuclear-test-ban Treaty Organization (CTBTO) will ultimately consist of 170 seismic stations (50 primary and 120 auxiliary) in 76 countries around the world. The Network is still under the development, but currently more than 80% of the network is in operation. The objective of seismic monitoring is to detect and locate underground nuclear explosions. However, the data from the IMS also can be widely used for scientific and civil purposes. In this study we present the results of data analysis of the seismic sequence in 2016 in Central Italy. Several hundred earthquakes were recorded for this sequence by the seismic stations of the IMS. All events were accurately located the analysts of the International Data Centre (IDC) of the CTBTO. In this study we will present the epicentral and magnitude distribution, station recordings and teleseismic phases as obtained from the Reviewed Event Bulletin (REB). We will also present a comparison of the database of the IDC with the databases of the European-Mediterranean Seismological Centre (EMSC) and U.S. Geological Survey (USGS). Present work shows that IMS data can be used for earthquake sequence analyses and can play an important role in seismological research.

  4. A test of a global seismic system for monitoring earthquakes and underground nuclear explosions

    International Nuclear Information System (INIS)

    Bowman, J.R.; Muirhead, K.; Spiliopoulos, S.; Jepsen, D.; Leonard, M.

    1993-01-01

    Australia is a member of the Group of Scientific Experts (GSE) to consider international cooperative measures to detect and identify events, an ad hoc group of the United Nations Conference on Disarmament. The GSE conducted a large-scale technical test (GSETT-2) from 22 April to 9 June 1991 that focused on the exchange and analysis of seismic parameter and waveform data. Thirty-four countries participated in GSETT-2, and data were contributed from 60 stations on all continents. GSETT-2 demonstrated the feasibility of collecting and transmitting large volumes (around 1 giga-byte) of digital data around the world, and of producing a preliminary bulletin of global seismicity within 48 hours and a final bulletin within 7 days. However, the experiment also revealed the difficulty of keeping up with the flow of data and analysis with existing resources. The Final Event Bulletins listed 3715 events for the 42 recording days of the test, about twice the number reported routinely by another international agency 5 months later. The quality of the Final Event Bulletin was limited by the uneven spatial distribution of seismic stations that contributed to GSETT-2 and by the ambiguity of associating phases detected by widely separated stations to form seismic events. A monitoring system similar to that used in GSETT-2 could provide timely and accurate reporting of global seismicity. It would need an improved distribution of stations, application of more conservative event formation rules and further development of analysis software. 8 refs., 9 figs

  5. 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

  6. Study on anti-seismic test of control rod driving system suspended by magnetic force

    International Nuclear Information System (INIS)

    Zhang Zhihua; Qian Dazhi; Xu Xianqi; Huang Hongwen; Zhang Zhengming; Wu Xinxin; Hu Xiao

    2012-01-01

    To verify the stability, reliability and security function in extreme conditions, the anti-seismic test of control rod drive line was conducted. Drop-time of control rod drive line in different earthquake intensities was got. The response and strain values of control rod drive line acceleration on SL-1, SL-2 level were measured. Safety functions of control rod drive line were validated in different work conditions. Anti-seismic test data shows that the driving system can keep the structure's integrality and realize operation function under OBE and SSE. (authors)

  7. Seismic design criteria for the system 80+ advanced light water reactor

    International Nuclear Information System (INIS)

    Manrique, M.A.; Dermitzakis, S.N.; Gerdes, L.D.; Kennedy, R.P.; Idriss, I.M.; Cassidy, J.R.

    1991-01-01

    This paper presents the development of seismic design criteria in support of design certification by the Nuclear Regulatory Commission (NRC) of the ABB-Combustion Engineering's System 80+ Standard Design. The design certification effort is sponsored by the US Department of Energy (DOE). The development of the design criteria included: (a) development of the seismic control motion, (b) development of generic soil profiles for anticipated sites, (c) generation of in-structure response spectra and design loads for structures and equipment through soil-structure interaction (SSI) analyses, and (d) acceptance criteria for future construction sites

  8. Seismic Safety Margins Research Program. Phase 1. Project V. Structural sub-system response: subsystem response review

    International Nuclear Information System (INIS)

    Fogelquist, J.; Kaul, M.K.; Koppe, R.; Tagart, S.W. Jr.; Thailer, H.; Uffer, R.

    1980-03-01

    This project is directed toward a portion of the Seismic Safety Margins Research Program which includes one link in the seismic methodology chain. The link addressed here is the structural subsystem dynamic response which consists of those components and systems whose behavior is often determined decoupled from the major structural response. Typically the mathematical model utilized for the major structural response will include only the mass effects of the subsystem and the main model is used to produce the support motion inputs for subsystem seismic qualification. The main questions addressed in this report have to do with the seismic response uncertainty of safety-related components or equipment whose seismic qualification is performed by (a) analysis, (b) tests, or (c) combinations of analysis and tests, and where the seismic input is assumed to have no uncertainty

  9. Seismic Response of Power Transmission Tower-Line System Subjected to Spatially Varying Ground Motions

    Directory of Open Access Journals (Sweden)

    Li Tian

    2010-01-01

    Full Text Available The behavior of power transmission tower-line system subjected to spatially varying base excitations is studied in this paper. The transmission towers are modeled by beam elements while the transmission lines are modeled by cable elements that account for the nonlinear geometry of the cables. The real multistation data from SMART-1 are used to analyze the system response subjected to spatially varying ground motions. The seismic input waves for vertical and horizontal ground motions are also generated based on the Code for Design of Seismic of Electrical Installations. Both the incoherency of seismic waves and wave travel effects are accounted for. The nonlinear time history analytical method is used in the analysis. The effects of boundary conditions, ground motion spatial variations, the incident angle of the seismic wave, coherency loss, and wave travel on the system are investigated. The results show that the uniform ground motion at all supports of system does not provide the most critical case for the response calculations.

  10. A regulatory perspective on appropriate seismic loading stress criteria for advanced light water reactor piping systems

    International Nuclear Information System (INIS)

    Terao, D.

    1995-01-01

    In the foregoing sections, the author has discussed the NRC staff's perspective on the evolving seismic design criteria for piping systems. He also addressed the need for developing seismic loading stress criteria and provided several recommendations and considerations for ensuring piping functional capability, pressure integrity, and structural integrity. Overall, the general consensus in the NRC staff is that in the past several years, many initiatives have been developed and implemented by the industry and the NRC staff to reduce the excessive conservatisms that might have existed in nuclear piping system design criteria. The regulations, regulatory guides, and Standard Review Plan have been (or are currently in the process of being) revised to reflect these initiatives in an effort to produce requirements and guidelines that will continue to result in a safe and practical design of piping systems. However, further proposals to reduce margins are continually being submitted to the ASME Boiler and Pressure Vessel Code and the NRC for review and approval. Improvements to the piping seismic design criteria are always encouraged, but there is a point at which the benefits might be outweighed by drawbacks. Because of this rapidly evolving situation the need exists for the industry and the NRC staff to develop a course of action to ensure that piping seismic design criteria for future ALWR plants will result in piping system designs that provide adequate safety margins and practical designs at a reasonable cost

  11. Geological and Seismic Data Mining For The Development of An Interpretation System Within The Alptransit Project

    Science.gov (United States)

    Klose, C. D.; Giese, R.; Löw, S.; Borm, G.

    Especially for deep underground excavations, the prediction of the locations of small- scale hazardous geotechnical structures is nearly impossible when exploration is re- stricted to surface based methods. Hence, for the AlpTransit base tunnels, exploration ahead has become an essential component of the excavation plan. The project de- scribed in this talk aims at improving the technology for the geological interpretation of reflection seismic data. The discovered geological-seismic relations will be used to develop an interpretation system based on artificial intelligence to predict hazardous geotechnical structures of the advancing tunnel face. This talk gives, at first, an overview about the data mining of geological and seismic properties of metamorphic rocks within the Penninic gneiss zone in Southern Switzer- land. The data results from measurements of a specific geophysical prediction system developed by the GFZ Potsdam, Germany, along the 2600 m long and 1400 m deep Faido access tunnel. The goal is to find those seismic features (i.e. compression and shear wave velocities, velocity ratios and velocity gradients) which show a significant relation to geological properties (i.e. fracturing and fabric features). The seismic properties were acquired from different tomograms, whereas the geolog- ical features derive from tunnel face maps. The features are statistically compared with the seismic rock properties taking into account the different methods used for the tunnel excavation (TBM and Drill/Blast). Fracturing and the mica content stay in a positive relation to the velocity values. Both, P- and S-wave velocities near the tunnel surface describe the petrology better, whereas in the interior of the rock mass they correlate to natural micro- and macro-scopic fractures surrounding tectonites, i.e. cataclasites. The latter lie outside of the excavation damage zone and the tunnel loos- ening zone. The shear wave velocities are better indicators for rock

  12. An Experimental Seismic Data and Parameter Exchange System for Interim NEAMTWS

    Science.gov (United States)

    Hanka, W.; Hoffmann, T.; Weber, B.; Heinloo, A.; Hoffmann, M.; Müller-Wrana, T.; Saul, J.

    2009-04-01

    In 2008 GFZ Potsdam has started to operate its global earthquake monitoring system as an experimental seismic background data centre for the interim NEAMTWS (NE Atlantic and Mediterranean Tsunami Warning System). The SeisComP3 (SC3) software, developed within the GITEWS (German Indian Ocean Tsunami Early Warning System) project was extended to test the export and import of individual processing results within a cluster of SC3 systems. The initiated NEAMTWS SC3 cluster consists presently of the 24/7 seismic services at IMP, IGN, LDG/EMSC and KOERI, whereas INGV and NOA are still pending. The GFZ virtual real-time seismic network (GEOFON Extended Virtual Network - GEVN) was substantially extended by many stations from Western European countries optimizing the station distribution for NEAMTWS purposes. To amend the public seismic network (VEBSN - Virtual European Broadband Seismic Network) some attached centres provided additional private stations for NEAMTWS usage. In parallel to the data collection by Internet the GFZ VSAT hub for the secured data collection of the EuroMED GEOFON and NEAMTWS backbone network stations became operational and the first data links were established. In 2008 the experimental system could already prove its performance since a number of relevant earthquakes have happened in NEAMTWS area. The results are very promising in terms of speed as the automatic alerts (reliable solutions based on a minimum of 25 stations and disseminated by emails and SMS) were issued between 2 1/2 and 4 minutes for Greece and 5 minutes for Iceland. They are also promising in terms of accuracy since epicenter coordinates, depth and magnitude estimates were sufficiently accurate from the very beginning, usually don't differ substantially from the final solutions and provide a good starting point for the operations of the interim NEAMTWS. However, although an automatic seismic system is a good first step, 24/7 manned RTWCs are mandatory for regular manual verification

  13. Shallow seismicity in volcanic system: what role does the edifice play?

    Science.gov (United States)

    Bean, Chris; Lokmer, Ivan

    2017-04-01

    Seismicity in the upper two kilometres in volcanic systems is complex and very diverse in nature. The origins lie in the multi-physics nature of source processes and in the often extreme heterogeneity in near surface structure, which introduces strong seismic wave propagation path effects that often 'hide' the source itself. Other complicating factors are that we are often in the seismic near-field so waveforms can be intrinsically more complex than in far-field earthquake seismology. The traditional focus for an explanation of the diverse nature of shallow seismic signals is to call on the direct action of fluids in the system. Fits to model data are then used to elucidate properties of the plumbing system. Here we show that solutions based on these conceptual models are not unique and that models based on a diverse range of quasi-brittle failure of low stiffness near surface structures are equally valid from a data fit perspective. These earthquake-like sources also explain aspects of edifice deformation that are as yet poorly quantified.

  14. SEISMIC RESISTING PERFORMANCE OF A NEW DOUBLE TUBE HYBRID SYSTEM FOR MULTI-STORY BUILDINGS

    OpenAIRE

    Nasruddin

    2012-01-01

    Investigation on Double Tube Hybrid System (DTHS) through experimental work and analytical study are conducted as a part of the proposal on the seismic design method for Double Tubes Hybrid System (DTHS) for buildings. This structural system comprises Energy Dissipation Structural Walls (EDSW) as the interior tube and Spandrel Wall Frame (SWF) as the exterior tube. EDSW is composed of two reinforced concrete walls linked by steel coupling girders. The RC walls are not anchored to the foundati...

  15. The Experimental Research on Seismic Capacity of the Envelope Systems with Steel Frame

    Science.gov (United States)

    Li, Jiuyang; Wang, Bingbing; Li, Hengxu

    2017-09-01

    In this paper, according to the present application situation of the external envelope systems steel frame in the severe cold region, the stuffed composite wall panels are improved, the flexible connection with the steel frame is designed, the reduced scale specimens are made, the seismic capacity test is made and some indexes of the envelope systems such as bearing capacity, energy consumption and ductility, etc. are compared, which provide reference for the development and application of the steel frame envelope systems.

  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. The 2012 Emilia seismic sequence (Northern Italy): Imaging the thrust fault system by accurate aftershock location

    Science.gov (United States)

    Govoni, Aladino; Marchetti, Alessandro; De Gori, Pasquale; Di Bona, Massimo; Lucente, Francesco Pio; Improta, Luigi; Chiarabba, Claudio; Nardi, Anna; Margheriti, Lucia; Agostinetti, Nicola Piana; Di Giovambattista, Rita; Latorre, Diana; Anselmi, Mario; Ciaccio, Maria Grazia; Moretti, Milena; Castellano, Corrado; Piccinini, Davide

    2014-05-01

    Starting from late May 2012, the Emilia region (Northern Italy) was severely shaken by an intense seismic sequence, originated from a ML 5.9 earthquake on May 20th, at a hypocentral depth of 6.3 km, with thrust-type focal mechanism. In the following days, the seismic rate remained high, counting 50 ML ≥ 2.0 earthquakes a day, on average. Seismicity spreads along a 30 km east-west elongated area, in the Po river alluvial plain, in the nearby of the cities Ferrara and Modena. Nine days after the first shock, another destructive thrust-type earthquake (ML 5.8) hit the area to the west, causing further damage and fatalities. Aftershocks following this second destructive event extended along the same east-westerly trend for further 20 km to the west, thus illuminating an area of about 50 km in length, on the whole. After the first shock struck, on May 20th, a dense network of temporary seismic stations, in addition to the permanent ones, was deployed in the meizoseismal area, leading to a sensible improvement of the earthquake monitoring capability there. A combined dataset, including three-component seismic waveforms recorded by both permanent and temporary stations, has been analyzed in order to obtain an appropriate 1-D velocity model for earthquake location in the study area. Here we describe the main seismological characteristics of this seismic sequence and, relying on refined earthquakes location, we make inferences on the geometry of the thrust system responsible for the two strongest shocks.

  18. Use of the Delphi approach in seismic qualification of existing electrical and mechanical equipment and distribution systems

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1983-01-01

    In this paper a method of seismic reevaluation is described which permits every seismic Category I component and distribution system to be evaluated and an explicit estimate of seismic capability, in terms of an acceleration level which defines limits of structural adequacy and leak-tight integrity, be defined for each item. The procedure employees the Delphi method where a team of seismic design experts independently inspect each component and distribution system in its as installed condition in the field and estimate the acceleration level that component or system could withstand and still meet applicable seismic design code or other criteria requirements. The accuracy and any potential bias in the estimates can be evaluated by a small control sample where selected components which were also surveyed by the Delphi team are independently evaluated in detail using current analytical techniques. This entire procedure can be accomplished for an estimated 5000 to 10,000 manhours per plant

  19. Seismic fragility formulations for segmented buried pipeline systems including the impact of differential ground subsidence

    Energy Technology Data Exchange (ETDEWEB)

    Pineda Porras, Omar Andrey [Los Alamos National Laboratory; Ordaz, Mario [UNAM, MEXICO CITY

    2009-01-01

    Though Differential Ground Subsidence (DGS) impacts the seismic response of segmented buried pipelines augmenting their vulnerability, fragility formulations to estimate repair rates under such condition are not available in the literature. Physical models to estimate pipeline seismic damage considering other cases of permanent ground subsidence (e.g. faulting, tectonic uplift, liquefaction, and landslides) have been extensively reported, not being the case of DGS. The refinement of the study of two important phenomena in Mexico City - the 1985 Michoacan earthquake scenario and the sinking of the city due to ground subsidence - has contributed to the analysis of the interrelation of pipeline damage, ground motion intensity, and DGS; from the analysis of the 48-inch pipeline network of the Mexico City's Water System, fragility formulations for segmented buried pipeline systems for two DGS levels are proposed. The novel parameter PGV{sup 2}/PGA, being PGV peak ground velocity and PGA peak ground acceleration, has been used as seismic parameter in these formulations, since it has shown better correlation to pipeline damage than PGV alone according to previous studies. By comparing the proposed fragilities, it is concluded that a change in the DGS level (from Low-Medium to High) could increase the pipeline repair rates (number of repairs per kilometer) by factors ranging from 1.3 to 2.0; being the higher the seismic intensity the lower the factor.

  20. Analysis of a piping system under seismic load using incremental hinge technique

    International Nuclear Information System (INIS)

    Ravi Kiran, A.; Agrawal, M.K.; Reddy, G.R.; Singh, R.K.; Vaze, K.K.; Ghosh, A.K.; Kushwaha, H.S.; Ramesh Babu, R.

    2008-01-01

    ASME Boiler and Pressure Vessel Code treats piping system as a series of components but not as an overall structural system. Limit analyses and collapse tests at component level are used to establish stress allowables on seismic stresses. The code does not consider the load redistributions and structural redundancy existing in piping systems that prevent system collapse even when one or more individual components loaded beyond their collapse levels. This necessitates a simple analytical method for evaluation of inelastic seismic response at system level. The present paper presents a simplified analytical procedure for predicting inelastic response of a typical piping system subjected to seismic load. The analytical method known as incremental hinge technique is based on plastic system behavior in which the yielded components are replaced with hinge models when a critical hinge moment is reached. It also takes into account the inelastic response spectrum reduction factors and displacement ductility. The analytical method is used to obtain the inelastic response, location of hinge formation and level of base excitation needed for hinge formation. The predicted hinge locations and hinge ordering is compared with the results of a shake table test conducted on the piping system. (author)

  1. Study on seismic reliability for foundation grounds and surrounding slopes of nuclear power plants. Proposal of evaluation methodology and integration of seismic reliability evaluation system

    International Nuclear Information System (INIS)

    Ohtori, Yasuki; Kanatani, Mamoru

    2006-01-01

    This paper proposes an evaluation methodology of annual probability of failure for soil structures subjected to earthquakes and integrates the analysis system for seismic reliability of soil structures. The method is based on margin analysis, that evaluates the ground motion level at which structure is damaged. First, ground motion index that is strongly correlated with damage or response of the specific structure, is selected. The ultimate strength in terms of selected ground motion index is then evaluated. Next, variation of soil properties is taken into account for the evaluation of seismic stability of structures. The variation of the safety factor (SF) is evaluated and then the variation is converted into the variation of the specific ground motion index. Finally, the fragility curve is developed and then the annual probability of failure is evaluated combined with seismic hazard curve. The system facilitates the assessment of seismic reliability. A generator of random numbers, dynamic analysis program and stability analysis program are incorporated into one package. Once we define a structural model, distribution of the soil properties, input ground motions and so forth, list of safety factors for each sliding line is obtained. Monte Carlo Simulation (MCS), Latin Hypercube Sampling (LHS), point estimation method (PEM) and first order second moment (FOSM) implemented in this system are also introduced. As numerical examples, a ground foundation and a surrounding slope are assessed using the proposed method and the integrated system. (author)

  2. 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

  3. 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

  4. A development of three-dimensional seismic isolation for advanced reactor systems in Japan: Pt.2

    International Nuclear Information System (INIS)

    Kenji Takahashi; Kazuhiko Inoue; Asao Kato; Masaki Morishita; Takafumi Fujita

    2005-01-01

    Two types of three-dimensional seismic isolation systems were developed for the fast breeder reactor (FBR). One is the three-dimensional entire building base isolation system It was developed by collecting concepts Japanese companies from which a combination system with air springs and hydraulic rocking suppression devices was selected. The other is the vertically isolated system for main components with horizontally entire building base isolation, which was developed by adopting coned disk spring devices. In the study, seismic condition was assumed based on a strict reference ground motion. Design data of the building and components are referred to FBR being developed as the 'Commercialized Fast Reactor Cycle System'. Analysis based on these assumed conditions showed suitable combinations of natural frequencies and damping ratios for isolation. Devices were developed to satisfy the combinations. In five years research and development, several verification tests were performed including shake table tests with scaled models. Finally it is found that the two types of seismic isolation systems are available for FBR. The result is reflected in the preliminary design guideline for the three-dimensional isolation system. (authors)

  5. 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

  6. Seismic Target Classification Using a Wavelet Packet Manifold in Unattended Ground Sensors Systems

    Directory of Open Access Journals (Sweden)

    Enliang Song

    2013-07-01

    Full Text Available One of the most challenging problems in target classification is the extraction of a robust feature, which can effectively represent a specific type of targets. The use of seismic signals in unattended ground sensor (UGS systems makes this problem more complicated, because the seismic target signal is non-stationary, geology-dependent and with high-dimensional feature space. This paper proposes a new feature extraction algorithm, called wavelet packet manifold (WPM, by addressing the neighborhood preserving embedding (NPE algorithm of manifold learning on the wavelet packet node energy (WPNE of seismic signals. By combining non-stationary information and low-dimensional manifold information, WPM provides a more robust representation for seismic target classification. By using a K nearest neighbors classifier on the WPM signature, the algorithm of wavelet packet manifold classification (WPMC is proposed. Experimental results show that the proposed WPMC can not only reduce feature dimensionality, but also improve the classification accuracy up to 95.03%. Moreover, compared with state-of-the-art methods, WPMC is more suitable for UGS in terms of recognition ratio and computational complexity.

  7. Interaction of processes may explain induced seismicity after shut-in in Enhanced Geothermal Systems

    Science.gov (United States)

    De Simone, Silvia; Carrera, Jesus; Vilarrasa, Victor

    2015-04-01

    Deep fluid injection is a necessary operation in several engineering sectors, like geothermal energy production, natural gas storage, CO2 storage, etc. The seismicity associated to these activities has, in some occasions, reached unexpected magnitude, raising public concern. Moreover, the occurrence of such seismicity after the injection shut-in pointed out the incompleteness of the knowledge and the inability of fully managing these processes. On the other hand, the growing attention toward clean energy makes it clear that we cannot abandon these procedures, which have a huge potential. Therefore, deeply understanding the mechanisms that induce seismicity is crucial. In this study we consider hydraulic stimulation of deep geothermal systems and analyze the mechanisms that may induce or trigger seismicity. Given that the basic mechanism is fluid pressure increase, secondary triggering processes have been studied. In detail, we attempt to identify the potential mechanisms that may trigger seismicity in the post-injection phase, when the overpressure decreases. These mechanisms have been investigated with a coupled and uncoupled approach, in order to understand the individual effects of each one and the effects of the interactions between them on the reservoir stability. Besides fluid overpressure, another relevant process is the temperature variation. Indeed, in the case of enhanced geothermal systems, the temperature contrast between the injected cold fluid and the deep hot reservoir is great and induces thermal stress, which sensibly affects the in-situ stress field. Therefore, we have studied overpressure and temperature effects by means of analytic solutions and by means of hydro-mechanical and thermo-hydro-mechanical numerical simulations. Results show that in fractured rocks the spatial variability of hydraulic and mechanic parameters provokes no isotropic variation of the tensional field, in response to pressure and temperature perturbations. Another

  8. J-SHIS - an integrated system for knowing seismic hazard information in Japan

    Science.gov (United States)

    Azuma, H.; Fujiwara, H.; Kawai, S.; Hao, K. X.; Morikawa, N.

    2015-12-01

    An integrated system of Japan seismic hazard information station (J-SHIS) was established in 2005 for issuing and exchanging information of the National Seismic Hazard Maps for Japan that are based on seismic hazard assessment (SHA). A simplified app, also named J-SHIS, for smartphones is popularly used in Japan based on the integrated system of http://www.j-shis.bosai.go.jp/map/?lang=en. "Smartphone tells hazard" is realized on a cellphone, a tablet and/or a PC. At a given spot, the comprehensive information of SHA map can be easily obtained as below: 1) A SHA probability at given intensity (JMA=5-, 5+, 6-, 6+) within 30 years. 2) A site amplification factor varies within 0.5 ~ 3.0 and expectation is 1 based on surface geology map information. 3) A depth of seismic basement down to ~3,000m based on deeper borehole and geological structure. 4) Scenario earthquake maps: By choosing an active fault, one got the average case for different parameters of the modeling. Then choose a case, you got the shaking map of intensity with color scale. "Seismic Hazard Karte tells more hazard" is another app based on website of http://www.j-shis.bosai.go.jp/labs/karte/. (1) For every mesh of 250m x 250m, professional service SHA information is provided over national-world. (2) With five ranks for eight items, comprehensive SHA information could be delivered. (3) Site amplification factor with an average index is given. (4) Deeper geologic structure modeling is provided with borehole profiling. (5) A SHA probability is assessed within 30 and/or 50 years for the given site. (6) Seismic Hazard curves are given for earthquake sources from inland active fault, subduction zone, undetermined and their summarization. (7) The JMA seismic intensities are assessed in long-term averaged periods of 500-years to ~100,000 years. The app of J-SHIS can be downloaded freely from http://www.j-shis.bosai.go.jp/app-jshis.

  9. Seismic Response of a Platform-Frame System with Steel Columns

    Directory of Open Access Journals (Sweden)

    Davide Trutalli

    2017-04-01

    Full Text Available Timber platform-frame shear walls are characterized by high ductility and diffuse energy dissipation but limited in-plane shear resistance. A novel lightweight constructive system composed of steel columns braced with oriented strand board (OSB panels was conceived and tested. Preliminary laboratory tests were performed to study the OSB-to-column connections with self-drilling screws. Then, the seismic response of a shear wall was determined performing a quasi-static cyclic-loading test of a full-scale specimen. Results presented in this work in terms of force-displacement capacity show that this system confers to shear walls high in-plane strength and stiffness with good ductility and dissipative capacity. Therefore, the incorporation of steel columns within OSB bracing panels results in a strong and stiff platform-frame system with high potential for low- and medium-rise buildings in seismic-prone areas.

  10. Seismic response analysis of structural system subjected to multiple support excitation

    International Nuclear Information System (INIS)

    Wu, R.W.; Hussain, F.A.; Liu, L.K.

    1978-01-01

    In the seismic analysis of a multiply supported structural system subjected to nonuniform excitations at each support point, the single response spectrum, the time history, and the multiple response spectrum are the three commonly employed methods. In the present paper the three methods are developed, evaluated, and the limitations and advantages of each method assessed. A numerical example has been carried out for a typical piping system. Considerably smaller responses have been predicted by the time history method than that by the single response spectrum method. This is mainly due to the fact that the phase and amplitude relations between the support excitations are faithfully retained in the time history method. The multiple response spectrum prediction has been observed to compare favourably with the time history method prediction. Based on the present evaluation, the multiple response spectrum method is the most efficient method for seismic response analysis of structural systems subjected to multiple support excitation. (Auth.)

  11. New serial time codes for seismic short period and long period data acquisition systems

    International Nuclear Information System (INIS)

    Kolvankar, V.G.; Rao, D.S.

    1988-01-01

    This paper discusses a new time code for time indexing multichannel short period (1 to 25 hz) seismic event data recorded on a single track of magnetic tape in digital format and discusses its usefulness in contrast to Vela time code used in continuous analog multichannel data recording system on multitrack instrumentation tape deck. This paper also discusses another time code, used for time indexing of seismic long period (DC to 2.5 seconds) multichannel data recorded on a single track of magnetic tape in digital format. The time code decoding and display system developed to provide quick access to any desired portion of the tape in both data recording and repro duce system is also discussed. (author). 7 figs

  12. Quantitative estimation of lithofacies from seismic data in a tertiary turbidite system in the North Sea

    Energy Technology Data Exchange (ETDEWEB)

    Joerstad, A.K.; Avseth, P.Aa; Mukerji, T.; Mavko, G.; Granli, J.R.

    1998-12-31

    Deep water clastic systems and associated turbidite reservoirs are often characterized by very complex sand distributions and reservoir description based on conventional seismic and well-log stratigraphic analysis may be very uncertain in these depositional environments. There is shown that reservoirs in turbidite systems have been produced very inefficiently in conventional development. More than 70% of the mobile oil is commonly left behind, because of the heterogeneous nature of these reservoirs. In this study there is examined a turbidite system in the North Sea with five available wells and a 3-D seismic near and far offset stack to establish most likely estimates of facies and pore fluid within the cube. 5 figs.

  13. Upgrading the seismic performance of the interior water pipe supporting system of a cooling tower

    International Nuclear Information System (INIS)

    Manos, G.C.; Soulis, V.J.

    2005-01-01

    This paper presents results from a numerical study that was performed in order to simulate the seismic behavior of the interior support system of the piping and cooling features of a cooling tower in one of the old power stations located in an area at the North-Western part of Greece. This cooling tower has a diameter of 60 m and a height of 100 m. The interior piping support system consists mainly of a series of nine-meter high pre-cast vertical columns made by pre-stressed concrete; these columns, together with reinforced concrete pre-cast horizontal beams that are joined monolithically with the columns at their top, form the old interior supporting system. This system represented a very flexible structure, a fact that was verified from a preliminary numerical analysis of its seismic behavior. The maximum response to the design earthquake levels resulted in large horizontal displacements at the top of the columns as well as overstress to some of the columns. The most important part of the current numerical investigation was to examine various strengthening schemes of the old interior support system and to select one that will demonstrate acceptable seismic behavior. (authors)

  14. A filter circuit board for the Earthworm Seismic Data Acquisition System

    Science.gov (United States)

    Jensen, Edward Gray

    2000-01-01

    The Earthworm system is a seismic network data acquisition and processing system used by the Northern California Seismic Network as well as many other seismic networks. The input to the system is comprised of many realtime electronic waveforms fed to a multi-channel digitizer on a PC platform. The digitizer consists of one or more National Instruments Corp. AMUX–64T multiplexer boards attached to an A/D converter board located in the computer. Originally, passive filters were installed on the multiplexers to eliminate electronic noise picked up in cabling. It was later discovered that a small amount of crosstalk occurred between successive channels in the digitizing sequence. Though small, this crosstalk will cause what appear to be small earthquake arrivals at the wrong time on some channels. This can result in erroneous calculation of earthquake arrival times, particularly by automated algorithms. To deal with this problem, an Earthworm filter board was developed to provide the needed filtering while eliminating crosstalk. This report describes the tests performed to find a suitable solution, and the design of the circuit board. Also included are all the details needed to build and install this board in an Earthworm system or any other system using the AMUX–64T board. Available below is the report in PDF format as well as an archive file containing the circuit board manufacturing information.

  15. 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)

  16. Induced seismicity hazard and risk by enhanced geothermal systems: an expert elicitation approach

    Science.gov (United States)

    Trutnevyte, Evelina; Azevedo, Inês L.

    2018-03-01

    Induced seismicity is a concern for multiple geoenergy applications, including low-carbon enhanced geothermal systems (EGS). We present the results of an international expert elicitation (n = 14) on EGS induced seismicity hazard and risk. Using a hypothetical scenario of an EGS plant and its geological context, we show that expert best-guess estimates of annualized exceedance probabilities of an M ≥ 3 event range from 0.2%-95% during reservoir stimulation and 0.2%-100% during operation. Best-guess annualized exceedance probabilities of M ≥ 5 event span from 0.002%-2% during stimulation and 0.003%-3% during operation. Assuming that tectonic M7 events could occur, some experts do not exclude induced (triggered) events of up to M7 too. If an induced M = 3 event happens at 5 km depth beneath a town with 10 000 inhabitants, most experts estimate a 50% probability that the loss is contained within 500 000 USD without any injuries or fatalities. In the case of an induced M = 5 event, there is 50% chance that the loss is below 50 million USD with the most-likely outcome of 50 injuries and one fatality or none. As we observe a vast diversity in quantitative expert judgements and underlying mental models, we conclude with implications for induced seismicity risk governance. That is, we suggest documenting individual expert judgements in induced seismicity elicitations before proceeding to consensual judgements, to convene larger expert panels in order not to cherry-pick the experts, and to aim for multi-organization multi-model assessments of EGS induced seismicity hazard and risk.

  17. Fractal and chaotic laws on seismic dissipated energy in an energy system of engineering structures

    Science.gov (United States)

    Cui, Yu-Hong; Nie, Yong-An; Yan, Zong-Da; Wu, Guo-You

    1998-09-01

    Fractal and chaotic laws of engineering structures are discussed in this paper, it means that the intrinsic essences and laws on dynamic systems which are made from seismic dissipated energy intensity E d and intensity of seismic dissipated energy moment I e are analyzed. Based on the intrinsic characters of chaotic and fractal dynamic system of E d and I e, three kinds of approximate dynamic models are rebuilt one by one: index autoregressive model, threshold autoregressive model and local-approximate autoregressive model. The innate laws, essences and systematic error of evolutional behavior I e are explained over all, the short-term behavior predictability and long-term behavior probability of which are analyzed in the end. That may be valuable for earthquake-resistant theory and analysis method in practical engineering structures.

  18. The Effectiveness of Seismic Isolation System for Nuclear Equipment

    International Nuclear Information System (INIS)

    Kim, Min-Kyu; Choun, Young-Sun; Seo, Jeong-Moon

    2005-04-01

    In this study, the Emergency Diesel Generator and Off-site Transformer were selected for isolation. For the selection of the most suitable base isolation system, the literature review and the numerical analysis were performed. For the decision of the parameter of isolation system, the sensitivity analysis was performed. Finally the conceptual design of each equipment was performed. In case of EDG, the Coil Spring and Viscous Damper system was selected for isolation system and 45% isolation effect was determined. For the OST, the FPS was selected and 69% isolation effect was determined

  19. The Effectiveness of Seismic Isolation System for Nuclear Equipment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min-Kyu; Choun, Young-Sun; Seo, Jeong-Moon

    2005-04-15

    In this study, the Emergency Diesel Generator and Off-site Transformer were selected for isolation. For the selection of the most suitable base isolation system, the literature review and the numerical analysis were performed. For the decision of the parameter of isolation system, the sensitivity analysis was performed. Finally the conceptual design of each equipment was performed. In case of EDG, the Coil Spring and Viscous Damper system was selected for isolation system and 45% isolation effect was determined. For the OST, the FPS was selected and 69% isolation effect was determined.

  20. Seismic testing

    International Nuclear Information System (INIS)

    Sollogoub, Pierre

    2001-01-01

    This lecture deals with: qualification methods for seismic testing; objectives of seismic testing; seismic testing standards including examples; main content of standard; testing means; and some important elements of seismic testing

  1. 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)

  2. FN-tandem accelerator in Bucharest after the seismic protection system installation

    International Nuclear Information System (INIS)

    Marinescu, L.; Petrascu, M.; Dima, R.; Serban, D.

    1996-01-01

    Subsequent to the installation of the seismic protection system in the HVEC FN-tandem accelerator in Bucharest, observations have been made of the efficiency of the protection devices. Performance and improvements of the new voltage divider are discussed. The SF 6 content of the insulating gas is shown to be conveniently measured using the elastic recoil technique and a robust ioniser heater is described for the HICONEX source. (orig.)

  3. 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

  4. Risk assessment and early warning systems for industrial facilities in seismic zones

    International Nuclear Information System (INIS)

    Salzano, Ernesto; Garcia Agreda, Anita; Di Carluccio, Antonio; Fabbrocino, Giovanni

    2009-01-01

    Industrial equipments and systems can suffer structural damage when hit by earthquakes, so that accidental scenarios as fire, explosion and dispersion of toxic substances can take place. As a result, overall damage to people, environment and properties increases. The present paper deals with seismic risk analysis of industrial facilities where atmospheric storage tanks (anchored or unanchored to ground), horizontal pressurised tanks, reactors and pumps are installed. Simplified procedures and methodologies based on historical database and literature data on natural-technological (Na-Tech) accidents for seismic risk assessment are discussed. Equipment-specific fragility curves have been thus derived depending on a single earthquake measure, peak ground acceleration (PGA). Fragility parameters have been then transformed to linear probit coefficients in order to obtain reliable threshold values for earthquake intensity measure, both for structural damage and loss of containment. These threshold values are of great interest when development of active and passive mitigation actions and systems, safety management, and the implementation of early warning system are concerned. The approach is general and can be implemented in any available code or procedure for risk assessment. Some results of seismic analysis of atmospheric storage tanks are also presented for validation.

  5. A seismic network to investigate the sedimentary hosted hydrothermal Lusi system

    Science.gov (United States)

    Javad Fallahi, Mohammad; Mazzini, Adriano; Lupi, Matteo; Obermann, Anne; Karyono, Karyono

    2016-04-01

    The 29th of May 2006 marked the beginning of the sedimentary hosted hydrothermal Lusi system. During the last 10 years we witnessed numerous alterations of the Lusi system behavior that coincide with the frequent seismic and volcanic activity occurring in the region. In order to monitor the effect that the seismicity and the activity of the volcanic arc have on Lusi, we deployed a ad hoc seismic network. This temporary network consist of 10 broadband and 21 short period stations and is currently operating around the Arjuno-Welirang volcanic complex, along the Watukosek fault system and around Lusi, in the East Java basin since January 2015. We exploit this dataset to investigate surface wave and shear wave velocity structure of the upper-crust beneath the Arjuno-Welirang-Lusi complex in the framework of the Lusi Lab project (ERC grant n° 308126). Rayleigh and Love waves travelling between each station-pair are extracted by cross-correlating long time series of ambient noise data recorded at the stations. Group and phase velocity dispersion curves are obtained by time-frequency analysis of cross-correlation functions, and are tomographically inverted to provide 2D velocity maps corresponding to different sampling depths. 3D shear wave velocity structure is then acquired by inverting the group velocity maps.

  6. FE Modelling of the Seismic Behavior of Wide Beam-Column Joints Strengthened with CFRP Systems

    Directory of Open Access Journals (Sweden)

    Giuseppe Santarsiero

    2018-02-01

    Full Text Available A large share of reinforced concrete (RC framed buildings is provided with wide beams being a type of beam allowing greater freedom in the architectural arrangement of interiors, beyond further advantage due to fewer formworks needed during the construction. Nevertheless, little attention has been devoted to the seismic vulnerability of this kind of framed RC buildings as well as to the study of strengthening systems purposely developed for wide beams and wide beam-column connections. Under these premises, this paper proposes simple strengthening solutions made by Fibre Reinforced Polymers (FRP systems able to effectively improve seismic capacity through feasible arrangement suitable in case a wide beam is present. On the basis of wide beam-column joints previously tested without strengthening system, detailed nonlinear finite element models were calibrated. Then, an FRP strengthening intervention based on a brand new arrangement was modeled in order to perform additional simulations under seismic actions. This way, the effectiveness of the strengthening intervention was assessed finding out that significant strength and ductility increments were achieved with a relatively simple and cheap strengthening arrangement. Additional research would be desirable in the form of experimental tests on the simulated wide beam-column joints.

  7. Response of piping system with semi-active variable stiffness damper under tri-directional seismic excitation

    International Nuclear Information System (INIS)

    Praveen Kumar; Jangid, R.S.; Reddy, G.R.

    2013-01-01

    Highlights: ► Piping system with semi-active variable stiffness damper is investigated under different seismic excitations. ► Switching control law and modified switching control law are adopted. ► There exist an optimum parameters of the SAVSD. ► Substantial reduction of the seismic response of piping system with SAVSD is observed. ► Good amount of energy dissipation is observed. -- Abstract: Seismic loads on piping system due to earthquakes can cause excessive vibrations, which can lead to serious instability resulting in damage or complete failure. In this paper, semi-active variable stiffness dampers (SAVSDs) have been studied to mitigate seismic response and vibration control of piping system used in the process industries, fossil and fissile fuel power plant. The SAVSD changes its stiffness depending upon the piping response and accordingly adds the control forces in the piping system. A study is conducted on the performance of SAVSD due to variation in device stiffness ratios in the switching control law and modified switching control law, which plays an important role in the present control algorithm of the damper. The effectiveness of the SAVSD in terms of reduction in the responses, namely, displacements, accelerations and base shear of the piping system is investigated by comparing uncontrolled responses under four different artificial earthquake motions with increasing amplitudes. The analytical results demonstrate that the SAVSDs under particular optimum parameters are very effective and practically implementable for the seismic response mitigation, vibration control and seismic requalification of piping systems

  8. Response of piping system with semi-active variable stiffness damper under tri-directional seismic excitation

    Energy Technology Data Exchange (ETDEWEB)

    Praveen Kumar, E-mail: praveen@barc.gov.in [Department of Civil Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India); Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Jangid, R.S. [Department of Civil Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India); Reddy, G.R. [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2013-05-15

    Highlights: ► Piping system with semi-active variable stiffness damper is investigated under different seismic excitations. ► Switching control law and modified switching control law are adopted. ► There exist an optimum parameters of the SAVSD. ► Substantial reduction of the seismic response of piping system with SAVSD is observed. ► Good amount of energy dissipation is observed. -- Abstract: Seismic loads on piping system due to earthquakes can cause excessive vibrations, which can lead to serious instability resulting in damage or complete failure. In this paper, semi-active variable stiffness dampers (SAVSDs) have been studied to mitigate seismic response and vibration control of piping system used in the process industries, fossil and fissile fuel power plant. The SAVSD changes its stiffness depending upon the piping response and accordingly adds the control forces in the piping system. A study is conducted on the performance of SAVSD due to variation in device stiffness ratios in the switching control law and modified switching control law, which plays an important role in the present control algorithm of the damper. The effectiveness of the SAVSD in terms of reduction in the responses, namely, displacements, accelerations and base shear of the piping system is investigated by comparing uncontrolled responses under four different artificial earthquake motions with increasing amplitudes. The analytical results demonstrate that the SAVSDs under particular optimum parameters are very effective and practically implementable for the seismic response mitigation, vibration control and seismic requalification of piping systems.

  9. Development of system design and seismic performance evaluation for reactor pool working platform of a research reactor

    International Nuclear Information System (INIS)

    Kwag, Shinyoung; Lee, Jong-Min; Oh, Jinho; Ryu, Jeong-Soo

    2014-01-01

    Highlights: • Design of reactor pool working platform (RPWP) is newly proposed for an open-tank-in-pool type research reactor. • Main concept of RPWP is to minimize the pool top radiation level. • Framework for seismic performance evaluation of nuclear SSCs in a deterministic and a probabilistic manner is proposed. • Structural integrity, serviceability, and seismic margin of the RPWP are evaluated during and after seismic events. -- Abstract: The reactor pool working platform (RPWP) has been newly designed for an open-tank-in-pool type research reactor, and its seismic response, structural integrity, serviceability, and seismic margin have been evaluated during and after seismic events in this paper. The main important concept of the RPWP is to minimize the pool top radiation level by physically covering the reactor pool of the open-tank-in-pool type research reactor and suppressing the rise of flow induced by the primary cooling system. It is also to provide easy handling of the irradiated objects under the pool water by providing guide tubes and refueling cover to make the radioisotopes irradiated and protect the reactor structure assembly. For this concept, the new three dimensional design model of the RPWP is established for manufacturing, installation and operation, and the analytical model is developed to analyze the seismic performance. Since it is submerged under and influenced by water, the hydrodynamic effect is taken into account by using the hydrodynamic added mass method. To investigate the dynamic characteristics of the RPWP, a modal analysis of the developed analytical model is performed. To evaluate the structural integrity and serviceability of the RPWP, the response spectrum analysis and response time history analysis have been performed under the static load and the seismic load of a safe shutdown earthquake (SSE). Their stresses are analyzed for the structural integrity. The possibility of an impact between the RPWP and the most

  10. 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

  11. SEISVIZ3D: Stereoscopic system for the representation of seismic data - Interpretation and Immersion

    Science.gov (United States)

    von Hartmann, Hartwig; Rilling, Stefan; Bogen, Manfred; Thomas, Rüdiger

    2015-04-01

    The seismic method is a valuable tool for getting 3D-images from the subsurface. Seismic data acquisition today is not only a topic for oil and gas exploration but is used also for geothermal exploration, inspections of nuclear waste sites and for scientific investigations. The system presented in this contribution may also have an impact on the visualization of 3D-data of other geophysical methods. 3D-seismic data can be displayed in different ways to give a spatial impression of the subsurface.They are a combination of individual vertical cuts, possibly linked to a cubical portion of the data volume, and the stereoscopic view of the seismic data. By these methods, the spatial perception for the structures and thus of the processes in the subsurface should be increased. Stereoscopic techniques are e. g. implemented in the CAVE and the WALL, both of which require a lot of space and high technical effort. The aim of the interpretation system shown here is stereoscopic visualization of seismic data at the workplace, i.e. at the personal workstation and monitor. The system was developed with following criteria in mind: • Fast rendering of large amounts of data so that a continuous view of the data when changing the viewing angle and the data section is possible, • defining areas in stereoscopic view to translate the spatial impression directly into an interpretation, • the development of an appropriate user interface, including head-tracking, for handling the increased degrees of freedom, • the possibility of collaboration, i.e. teamwork and idea exchange with the simultaneous viewing of a scene at remote locations. The possibilities offered by the use of a stereoscopic system do not replace a conventional interpretation workflow. Rather they have to be implemented into it as an additional step. The amplitude distribution of the seismic data is a challenge for the stereoscopic display because the opacity level and the scaling and selection of the data have to

  12. New portable sensor system for rotation seismic motion measurements

    Czech Academy of Sciences Publication Activity Database

    Brokešová, J.; Málek, Jiří

    2010-01-01

    Roč. 81, č. 8 (2010), 084501 ISSN 0034-6748 R&D Projects: GA ČR GAP210/10/0925 Institutional research plan: CEZ:AV0Z30460519 Keywords : rotation al seismology * sensor system Subject RIV: DC - Siesmology, Volcanology, Earth Structure Impact factor: 1.598, year: 2010

  13. Power System Assessment for the Burnt Mountain Seismic Observatory

    Science.gov (United States)

    1994-03-01

    monocrystalline cells and 15- 20% for polycrystalline cells. Less expensive cells, such as copper indium diselenide and cadmium telluride, have...solar cells are recommended. The silicon solar cells themselves are very reliable, however the reliability of panels and systems for field use can...is composed of four solar-cell panels each measuring 48 inches X 21 inches. The four panels in the array module are arranged, electrically, as two

  14. Seismic stratigraphic architecture of the Disko Bay trough-mouth fan system, West Greenland

    Science.gov (United States)

    Hofmann, Julia C.; Knutz, Paul C.

    2015-04-01

    Spatial and temporal changes of the Greenland Ice Sheet on the continental shelf bordering Baffin Bay remain poorly constrained. Then as now, fast-flowing ice streams and outlet glaciers have played a key role for the mass balance and stability of polar ice sheets. Despite their significance for Greenland Ice Sheet dynamics and evolution, our understanding of their long-term behaviour is limited. The central West Greenland margin is characterized by a broad continental shelf where a series of troughs extend from fjords to the shelf margin, acting as focal points for trough-mouth fan (TMF) accummulations. The sea-ward bulging morphology and abrupt shelf-break of these major depositional systems is generated by prograding depocentres that formed during glacial maxima when ice streams reached the shelf edge, delivering large amounts of subglacial sediment onto the continental slope (Ó Cofaigh et al., 2013). The aim of this study is to unravel the seismic stratigraphic architecture and depositional processes of the Disko Bay TMF, aerially the largest single sedimentary system in West Greenland, using 2D and 3D seismic reflection data, seabed bathymetry and stratigraphic information from exploration well Hellefisk-1. The south-west Disko Bay is intersected by a deep, narrow trough, Egedesminde Dyb, which extends towards the southwest and links to the shallower and broader cross-shelf Disko Trough (maximum water depths of > 1000 m and a trough length of c. 370 km). Another trough-like depression (trough length of c. 120 km) in the northern part of the TMF, indicating a previous position of the ice stream, can be distinguished on the seabed topographic map and the seismic images. The Disko Bay TMF itself extends from the shelf edge down to the abyssal plain (abyssal floor depths of 2000 m) of the southern Baffin Bay. Based on seismic stratigraphic configurations relating to reflection terminations, erosive patterns and seismic facies (Mitchum et al., 1977), the TMF

  15. 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

  16. Seismic calculation of buildings using a modular programme system

    International Nuclear Information System (INIS)

    Meskouris, K.; Weber, B.

    1980-01-01

    The calculation of the stress resultant and of the deformation forces affecting a building during an earthquake is shown here by two examples ( a multi-span bridge and a high multistage frame), the calculation being made in accordance with DIN E 4149 of December 1976, using the modular programme MISS-SMIS. In contrast to the usual programme systems, it is the user who determines the required algorithm with the aid of a mnemonic language and a collection of independent programme parts (modules). Thus, simple adjustment to the individual tasks to be carried out together with a high degree of transparency and flexibility of the calculation process is achieved. (orig.) [de

  17. 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

  18. Development of an evaluation method for seismic isolation systems of nuclear power facilities. Development of crossover piping design method for seismic isolation systems

    International Nuclear Information System (INIS)

    Otoyo, Teruyoshi; Otani, Akihito; Otani, Akihito; Fukushima, Shunsuke; Jimbo, Masakazu; Yamamoto, Tomofumi; Sakakida, Takaaki; Onishi, Shigenobu

    2014-01-01

    In the conceptual design of seismic isolation systems of nuclear power facilities, there exist two types of installation. The first type is to isolate both the reactor and the turbine buildings, the other is to isolate only the reactor building. In the latter type, the crossover piping, which installed between the isolated and the non-isolated buildings, is excited and deformed by the different motions of those buildings. In this study, shaking tests of 1/10 scaled model of the main steam piping and FEM analyses under multiple support excitation conditions have been performed to investigate the vibration behavior of the crossover piping. It was confirmed that modal time-history analyses could be in good agreement with the shaking test results. Also, Numerous combination methods were investigated by comparing response spectrum analyses and modal time-history analyses. In conclusion, response spectrum analyses using SRSS combinations could correspond to time-history analyses. (author)

  19. United States Geological Survey (USGS) FM cassette seismic-refraction recording system

    International Nuclear Information System (INIS)

    Murphy, J.M.

    1988-01-01

    In this two chapter report, instrumentation used to collect seismic data is described. This data acquisition system has two parts: (1) portable anolog seismic recorders and related ''hand-held-testers'' (HHT) and (2) portable digitizing units. During the anolog recording process, ground motion is sensed by a 2-Hz vertical-component seismometer. The voltage output from the seismometer is split without amplification and sent to three parallel amplifier circuit boards. Each circuit board amplifiers the seismic signal in three stages and then frequency modulates the signal. Amplification at the last two stages can be set by the user. An internal precision clock signal is also frequency modulated. The three data carrier frequencies, the clock carrier frequency, and a tape-speed compensation carrier frequency are summed and recorded on a recorded on a cassette tape. During the digitizing process, the cassette tapes are played back and the signals are demultiplexed and demodulated. An anolog-to-digital converter converts the signals to digital data which are stored on 8-inch floppy disks. 7 refs., 19 figs., 6 tabs

  20. MAHA: A comprehensive system for the storage and visualization of subsoil data for seismic microzonation

    Science.gov (United States)

    Di Felice, P.; Spadoni, M.

    2013-04-01

    MAHA is a database-centred software system for the storage and visualization of subsoil data used for the production of seismic microzonation maps in Italy. The application was implemented using open source software in order to grant its maximum diffusion and customization. A conceptual model of the subsoil, jointly developed by the Italian National Research Council and the National Department of Civil Protection, inspired the structure of the underlying database, consisting of 15 tables, 3 of which of spatial nature to accommodate geo-referenced data associated to points, lines and polygons. A web-GIS interface acts as a bridge between the user and the database, drives the input of geo-referenced data and enables the users to formulate different types of spatial queries. A series of forms designed "ad hoc" and enriched with combo boxes provide guided procedures to maximize the fluency of data entry and to reduce the possibility of erroneous typing. One of these procedures helps to transform the descriptions of the geological units (granular materials), given in technical paper documents by using a conversational style, into standardized numeric codes. Summary reports, produced in the pdf format, can be generated through decoding and graphic display of the parameters previously entered in the database. MAHA was approved by the national commission for seismic microzonation established by the Italian Prime Minister and, in the next years, it is expected to significantly support the entire process of map production in the urban areas more exposed to seismic hazard.

  1. New seismic monitoring observation system and data accessibility at Syowa Station

    Directory of Open Access Journals (Sweden)

    Masaki Kanao

    1999-03-01

    Full Text Available The seismic observation system at Syowa Station, East Antarctica was fully replaced in the wintering season of the 38th Japanese Antarctic Research Expedition (JARE-38 in 1996-1998. The old seismographic vault constructed in 1970 was closed at the end of JARE-38 because of cumulative damage to the inner side of the vault by continuous flowing in of water from walls in summer and its freezing in winter. All the seismometers were moved to a new seismographic hut (69°00′24.0″S, 39°35′06.0″E and 20m above mean sea level in April 1997. Seismic signals of the short-period (HES and broadband (STS-1 seismometers in the new hut are transmitted to the Earth Science Laboratory (ESL via analog cable 600m in length. The new acquisition system was installed in the ESL with 6-channel 24-bit A/D converters for both sensor signals. All digitized data are automatically transmitted from the A/D converter to a workstation via TCP/IP protocol. After parallel observations with the old acquisition system by personal computers and the new system during the wintering season of JARE-38,the main system was changed to the new one, which has some advantages for both the reduction of daily maintenance efforts and the data transport/communication processes via Internet by use of LAN at the station. In this report, details of the new seismographic hut and the recording system are described. Additionally, the seismic data accessibility for public use, including Internet service, is described.

  2. A seismic analysis of the driving system for the pulsed reactor

    International Nuclear Information System (INIS)

    Hu Yongtao; Fu Shixiang; Zeng Jianhua; Hong Jingfeng

    1991-01-01

    The driving system of the pulsed reactor contains control rods, pulsing o rod and sample rack. They are slender, and their drive function is required more strictly. First, a complete model which contains all driving system and reactor bridge is used. Then the substructure models are adopted. The results of calculation are compared with the experimental results. It shows that the analysis results are reliable and the substructure method is simple, available and utility. The seismic safety is evaluated by the results from response spectra method

  3. Design and realization of real-time processing system for seismic exploration

    International Nuclear Information System (INIS)

    Zhang Sifeng; Cao Ping; Song Kezhu; Yao Lin

    2010-01-01

    For solving real-time seismic data processing problems, a high-speed, large-capacity and real-time data processing system is designed based on FPGA and ARM. With the advantages of multi-processor, DRPS has the characteristics of high-speed data receiving, large-capacity data storage, protocol analysis, data splicing, data converting from time sequence into channel sequence, no dead time data ping-pong storage, etc. And with the embedded Linux operating system, DRPS has the characteristics of flexibility and reliability. (authors)

  4. Shaking table test and simulation analysis on failure characteristics of seismic isolation system

    International Nuclear Information System (INIS)

    Fukushima, Yasuaki; Iizuka, Maao; Satoh, Nobuhisa; Yoshikawa, Kazuhide; Katoh, Asao; Tanimoto, Eisuke

    2000-01-01

    Seismic safety and dynamic characteristics of the rubber bearing breaks of three types of base isolation system, natural rubber bearing + steel damper, lead rubber bearing and high damping rubber bearing, for nuclear power plant facilities were conducted by confirmed shaking table tests. The simulation analyses were conducted for the shaking table tests until the rubber broke. These results demonstrate that the dynamic behavior of base isolation system could be simulated closely until the rubber broke using simple analytical model based on static test. (author)

  5. Seismic Risk of the Base Isolation System Protected by the Hard Stop

    International Nuclear Information System (INIS)

    Kim, Jung Han; Choi, In-Kil; Kim, Min Kyu

    2015-01-01

    The concept of base isolation is to permit the deformation of isolator for absorbing seismic input wave from the ground. In a nuclear power plant design, allowable shear deformation of isolators should be enough to absorb the displacement response by extended design basis (EDB) ground motions. However isolators cannot resist over its displacement capacity. So, the clearance of hard stop (CHS) needs to be set between the response of base isolation system excited by the EDB ground motion and the displacement capacity of isolators. The isolation system must survive with high confidence in any seismic accident because it is a non-redundant system. Therefore, the CHS should be determined carefully based on the failure risk of base isolation system considering the uncertainties of earthquake responses and isolator capacities. In this research, the fragility curve of isolation system and its failure risk were estimated. The procedure to calculate the acceleration based fragility curve of the isolation system was developed. The fragility curve and failure risk for example case was estimated and its result was compared with different isolator capacities

  6. Development of real-time on-line vibration testing system for seismic experiments

    International Nuclear Information System (INIS)

    Horiuchi, T.; Nakagawa, M.; Kametani, M.

    1993-01-01

    An on-line vibration testing method is being developed for seismic experiments. This method combines computer simulation and an actuator for vibration testing of structures. A real-time, on-line testing system was developed to improve the method. In the system, the timing of the vibration testing and the computer simulation are the same. This allows time-dependent reaction forces, such as damping force, to be immediately considered in the computer simulation. The real-time system has many requirements, such as complicated matrix calculations within a small time step, and communication with outer devices like sensors and actuators through A/D and D/A converters. These functions arc accomplished by using a newly-developed, real-time controller that employs a parallel processing technique. A small structural model is used to demonstrate the system. The reliability and applicability of the system for seismic experiments can be demonstrated by comparing the results of the system and a shaking table, which are in almost agreement. (author)

  7. Seismic Risk of the Base Isolation System Protected by the Hard Stop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Han; Choi, In-Kil; Kim, Min Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The concept of base isolation is to permit the deformation of isolator for absorbing seismic input wave from the ground. In a nuclear power plant design, allowable shear deformation of isolators should be enough to absorb the displacement response by extended design basis (EDB) ground motions. However isolators cannot resist over its displacement capacity. So, the clearance of hard stop (CHS) needs to be set between the response of base isolation system excited by the EDB ground motion and the displacement capacity of isolators. The isolation system must survive with high confidence in any seismic accident because it is a non-redundant system. Therefore, the CHS should be determined carefully based on the failure risk of base isolation system considering the uncertainties of earthquake responses and isolator capacities. In this research, the fragility curve of isolation system and its failure risk were estimated. The procedure to calculate the acceleration based fragility curve of the isolation system was developed. The fragility curve and failure risk for example case was estimated and its result was compared with different isolator capacities.

  8. Information system evolution at the French National Network of Seismic Survey (BCSF-RENASS)

    Science.gov (United States)

    Engels, F.; Grunberg, M.

    2013-12-01

    The aging information system of the French National Network of Seismic Survey (BCSF-RENASS), located in Strasbourg (EOST), needed to be updated to satisfy new practices from Computer science world. The latter means to evolve our system at different levels : development method, datamining solutions, system administration. The new system had to provide more agility for incoming projects. The main difficulty was to maintain old system and the new one in parallel the time to validate new solutions with a restricted team. Solutions adopted here are coming from standards used by the seismological community and inspired by the state of the art of devops community. The new system is easier to maintain and take advantage of large community to find support. This poster introduces the new system and choosen solutions like Puppet, Fabric, MongoDB and FDSN Webservices.

  9. Doosan Experience on I and C Upgrade for Operating NPPs: Control Rod Control System and Automatic Seismic Trip System

    International Nuclear Information System (INIS)

    Nam, C.H.; Kim, K.H.; Lee, D.H.

    2012-01-01

    This paper describes DHIC's experience on upgrading 3 coil type control rod control system(CRCS), 4 coil type control element drive mechanism control system(CEDMCS) and automatic seismic trip system(ASTS). Common main feature of the above systems are full duplex system to prevent unwanted trip and mis-operation. 5 CRCS and CEDMCS have been supplied to Kori 1,2, Ulchin 1,2 and Younggwang 3 since 2010 and 7 CEDMCS are contracted to supply Korea Hydro and Nuclear Power Co.(KHNP) site. Also 16 ASTS are supplied and 12 ASTS will be supplied to operating and new NPPs within 3 years. (author)

  10. 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

  11. 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.).

  12. 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.).

  13. Basic concepts about application of dual vibration absorbers to seismic design of nuclear piping systems

    International Nuclear Information System (INIS)

    Hara, F.; Seto, K.

    1987-01-01

    The design value of damping for nuclear piping systems is a vital parameter in ensuring safety in nuclear plants during large earthquakes. Many experiments and on-site tests have been undertaken in nuclear-industry developed countries to determine rational design values. However damping value in nuclear piping systems is so strongly influenced by many piping parameters that it shows a tremendous dispersion in its experimental values. A new trend has recently appeared in designing nuclear pipings, where they attempt to use a device to absorb vibration energy induced by seismic excitation. A typical device is an energy absorbing device, made of a special material having a high capacity of plasticity, which is installed between the piping and the support. This paper deals with the basic study of application of dual vibration absorbers to nuclear piping systems to accomplish high damping value and reduce consequently seismic response at resonance frequencies of a piping system, showing their effectiveness from not only numerical calculation but also experimental evaluation of the vibration responses in a 3D model piping system equipped with dual two vibration absorbers

  14. Can Vrancea earthquakes be accurately predicted from unusual bio-system behavior and seismic-electromagnetic records?

    International Nuclear Information System (INIS)

    Enescu, D.; Chitaru, C.; Enescu, B.D.

    1999-01-01

    The relevance of bio-seismic research for the short-term prediction of strong Vrancea earthquakes is underscored. An unusual animal behavior before and during Vrancea earthquakes is described and illustrated in the individual case of the major earthquake of March 4, 1977. Several hypotheses to account for the uncommon behavior of bio-systems in relation to earthquakes in general and strong Vrancea earthquakes in particular are discussed in the second section. It is reminded that promising preliminary results concerning the identification of seismic-electromagnetic precursor signals have been obtained in the Vrancea seismogenic area using special, highly sensitive equipment. The need to correlate bio-seismic and seismic-electromagnetic researches is evident. Further investigations are suggested and urgent steps are proposed in order to achieve a successful short-term prediction of strong Vrancea earthquakes. (authors)

  15. Finding Trapped Miners by Using a Prototype Seismic Recording System Made from Music-Recording Hardware

    Science.gov (United States)

    Pratt, Thomas L.

    2009-01-01

    The goal of this project was to use off-the-shelf music recording equipment to build and test a prototype seismic system to listen for people trapped in underground chambers (mines, caves, collapsed buildings). Previous workers found that an array of geophones is effective in locating trapped miners; displaying the data graphically, as well as playing it back into an audio device (headphones) at high speeds, was found to be effective for locating underground tapping. The desired system should record the data digitally to allow for further analysis, be capable of displaying the data graphically, allow for rudimentary analysis (bandpass filter, deconvolution), and allow the user to listen to the data at varying speeds. Although existing seismic reflection systems are adequate to record, display and analyze the data, they are relatively expensive and difficult to use and do not have an audio playback option. This makes it difficult for individual mines to have a system waiting on the shelf for an emergency. In contrast, music recording systems, like the one I used to construct the prototype system, can be purchased for about 20 percent of the cost of a seismic reflection system and are designed to be much easier to use. The prototype system makes use of an ~$3,000, 16-channel music recording system made by Presonus, Inc., of Baton Rouge, Louisiana. Other manufacturers make competitive systems that would serve equally well. Connecting the geophones to the recording system required the only custom part of this system - a connector that takes the output from the geophone cable and breaks it into 16 microphone inputs to be connected to the music recording system. The connector took about 1 day of technician time to build, using about $300 in off-the-shelf parts. Comparisons of the music recording system and a standard seismic reflection system (A 24-channel 'Geode' system manufactured by Geometrics, Inc., of San Jose, California) were carried out at two locations. Initial

  16. Seismic response analysis of a piping system subjected to multiple support excitations in a base isolated NPP building

    International Nuclear Information System (INIS)

    Surh, Han-Bum; Ryu, Tae-Young; Park, Jin-Sung; Ahn, Eun-Woo; Choi, Chul-Sun; Koo, Ja Choon; Choi, Jae-Boong; Kim, Moon Ki

    2015-01-01

    Highlights: • Piping system in the APR 1400 NPP with a base isolation design is studied. • Seismic response of piping system in base isolated building are investigated. • Stress classification method is examined for piping subjected to seismic loading. • Primary stress of piping is reduced due to base isolation design. • Substantial secondary stress is observed in the main steam piping. - Abstract: In this study, the stress response of the piping system in the advanced power reactor 1400 (APR 1400) with a base isolation design subjected to seismic loading is addressed. The piping system located between the auxiliary building with base isolation and the turbine building with a fixed base is considered since it can be subjected to substantial relative support movement during seismic events. First, the support responses with respect to the base characteristic are investigated to perform seismic analysis for multiple support excitations. Finite element analyses are performed to predict the piping stress response through various analysis methods such as the response spectrum, seismic support movement and time history method. To separately evaluate the inertial effect and support movement effect on the piping stress, the stress is decomposed into a primary and secondary stress using the proposed method. Finally, influences of the base isolation design on the piping system in the APR 1400 are addressed. The primary stress based on the inertial loading is effectively reduced in a base isolation design, whereas a considerable amount of secondary stress is generated in the piping system connecting a base isolated building with a fixed base building. It is also confirmed that both the response spectrum analysis and seismic support movement analysis provide more conservative estimations of the piping stress compared to the time history analysis

  17. Development of a Seismic Setpoint Calculation Methodology Using a Safety System Approach

    International Nuclear Information System (INIS)

    Lee, Chang Jae; Baik, Kwang Il; Lee, Sang Jeong

    2013-01-01

    The Automatic Seismic Trip System (ASTS) automatically actuates reactor trip when it detects seismic activities whose magnitudes are comparable to a Safe Shutdown Earthquake (SSE), which is the maximum hypothetical earthquake at the nuclear power plant site. To ensure that the reactor is tripped before the magnitude of earthquake exceeds the SSE, it is crucial to reasonably determine the seismic setpoint. The trip setpoint and allowable value for the ASTS for Advanced Power Reactor (APR) 1400 Nuclear Power Plants (NPPs) were determined by the methodology presented in this paper. The ASTS that trips the reactor when a large earthquake occurs is categorized as a non safety system because the system is not required by design basis event criteria. This means ASTS has neither specific analytical limit nor dedicated setpoint calculation methodology. Therefore, we developed the ASTS setpoint calculation methodology by conservatively considering that of PPS. By incorporating the developed methodology into the ASTS for APR1400, the more conservative trip setpoint and allowable value were determined. In addition, the ZPA from the Operating Basis Earthquake (OBE) FRS of the floor where the sensor module is located is 0.1g. Thus, the allowance of 0.17g between OBE of 0.1 g and ASTS trip setpoint of 0.27 g is sufficient to prevent the reactor trip before the magnitude of the earthquake exceeds the OBE. In result, the developed ASTS setpoint calculation methodology is evaluated as reasonable in both aspects of the safety and performance of the NPPs. This will be used to determine the ASTS trip setpoint and allowable for newly constructed plants

  18. Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor

    International Nuclear Information System (INIS)

    Morrone, A.; Brussalis, W.G.

    1975-01-01

    The paper presents the analysis and results for a LMFBR system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% coefficient of restitution. The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a ten seconds Safe Shutdown Earthquake acceleration-time history at 0.005 seconds intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then used by the second program for the scram time determination. The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions. (orig./HP) [de

  19. Seismic analysis of a LNG storage tank isolated by a multiple friction pendulum system

    Science.gov (United States)

    Zhang, Ruifu; Weng, Dagen; Ren, Xiaosong

    2011-06-01

    The seismic response of an isolated vertical, cylindrical, extra-large liquefied natural gas (LNG) tank by a multiple friction pendulum system (MFPS) is analyzed. Most of the extra-large LNG tanks have a fundamental frequency which involves a range of resonance of most earthquake ground motions. It is an effective way to decrease the response of an isolation system used for extra-large LNG storage tanks under a strong earthquake. However, it is difficult to implement in practice with common isolation bearings due to issues such as low temperature, soft site and other severe environment factors. The extra-large LNG tank isolated by a MFPS is presented in this study to address these problems. A MFPS is appropriate for large displacements induced by earthquakes with long predominant periods. A simplified finite element model by Malhotra and Dunkerley is used to determine the usefulness of the isolation system. Data reported and statistically sorted include pile shear, wave height, impulsive acceleration, convective acceleration and outer tank acceleration. The results show that the isolation system has excellent adaptability for different liquid levels and is very effective in controlling the seismic response of extra-large LNG tanks.

  20. Seismic analysis with FEM for fuel transfer system of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Jia Xiaofeng; Liu Pengliang; Bi Xiangjun; Ji Shunying

    2012-01-01

    In the PWR nuclear power plant, the function of the fuel transfer system (FTS) is to transfer the fuel assembly between the reactor building and the fuel building. The seismic analysis of the transfer system structure should be carried out to ensure the safety under OBE and SSE. Therefore, the ANASYS 12.0 software is adopted to construct the finite element analysis model for the fuel transfer system in a million kilowatt nuclear power plant. For the various configurations of FTS in the operating process, the stresses of the main structures, such as the transfer tube, fuel assembly container, fuel conveyor car, lifting frame in the reactor building, lifting frame in the fuel building, support and guide structure of conveyor car and the lifting frame in both buildings, are computed. The stresses are combined with the method of square root of square sum (SRSS) and assessed under various seismic conditions based on RCCM code, the results of the assessment satisfy the code. The results show that the stresses of the fuel transfer system structure meet the strength requirement, meanwhile, it can withstand the earthquake well. (authors)

  1. Evaluation of seismic damage to bridges and highway systems in Shelby County, Tennessee

    Science.gov (United States)

    Jernigan, John Bailey

    Past earthquakes have demonstrated that bridges are one of the most vulnerable components of highway transportation systems. In addition to bridges, roadways may also be subject to damage, particularly in an area prone to earthquake-induced liquefaction. As a consequence, the highway transportation systems after an earthquake might be impaired and the post-earthquake emergency response might be compromised. Furthermore, the impact on the regional economy might be very significant from the damage to highway systems. Since highway transportation systems are critical lifelines for people living in an urban area, it is important to evaluate the vulnerability of bridges and highway systems in earthquake-prone regions. Memphis and Shelby County, Tennessee are located close to the southwestern segment of the New Madrid seismic zone (NMSZ). This zone produced three of the largest earthquakes in North America in 1811--1812. Presently, the NMSZ is still active and is considered by engineers, seismologists, and public officials as the most hazardous seismic zone in the central and eastern United States. Bridges in the Memphis area were generally not designed for seismic resistance until 1990. Therefore, the majority of existing bridges might suffer damage from earthquakes occurring in the NMSZ. The overall objective of this study is to evaluate the expected damage to bridges and roadways on the major routes in Memphis and Shelby County resulting from New Madrid earthquakes with the aid of geographic information system (GIS) technology. The road network selected for this study includes all the Interstate highway system, all the primary and secondary routes maintained by the state, and most of the major arterial routes. There are 452 bridges on the selected roadway systems and data pertinent to these bridges and roadway systems were collected and implemented as a GIS database. The bridges in the Memphis area were classified into several types and damage states were determined

  2. Study on the seismic monitoring system development against the adjacent countries nuclear test

    International Nuclear Information System (INIS)

    Min, Kyung Sik; Ahn, Jong Sung; Lee, Jong Wook; Chang, In Soon; Seo, In Seok; Kwak, Eun Ho

    1995-12-01

    The project was carried out to construct foundation for the monitoring of the neighboring countries's nuclear test by seismic method. For this, we collected, organized and analyzed the information about the Comparative Test Ban Treaty (CTBT) and investigated theoretical backgrounds of the elastic wave generation by the Nuclear test and the identification of the nuclear tests from the natural earthquakes. And the computer system was setup to obtain realtime data from the broadband seismograph in Inchon of the Korean Meteorological Agency. 15 refs. (Author)

  3. Ice drilling for blasting boreholes in deep seismic surveys (JARE-43 by steam type drilling system

    Directory of Open Access Journals (Sweden)

    Atsushi Watanabe

    2003-03-01

    Full Text Available A seismic exploration was accomplished in the austral summer of 2001-2002 by the 43rd Japanese Antarctic Research Expedition (JARE-43 along a profile oblique to that held by JARE-41 on the Mizuho Plateau, East Antarctica. We used a steam type drilling system to obtain seven blasting boreholes. We spent 7 to 8 hours to make an enough depth of the hole for one shot point. The holes were 35 to 40 cm in diameter and 23.5 to 28.7 m in depth. The average drilling speed was 3.25 m/hr.

  4. Capabilities of seismic and georadar 2D/3D imaging of shallow subsurface of transport route using the Seismobile system

    Science.gov (United States)

    Pilecki, Zenon; Isakow, Zbigniew; Czarny, Rafał; Pilecka, Elżbieta; Harba, Paulina; Barnaś, Maciej

    2017-08-01

    In this work, the capabilities of the Seismobile system for shallow subsurface imaging of transport routes, such as roads, railways, and airport runways, in different geological conditions were presented. The Seismobile system combines the advantages of seismic profiling using landstreamer and georadar (GPR) profiling. It consists of up to four seismic measuring lines and carriage with a suspended GPR antenna. Shallow subsurface recognition may be achieved to a maximum width of 10.5 m for a distance of 3.5 m between the measurement lines. GPR measurement is performed in the axis of the construction. Seismobile allows the measurement time, labour and costs to be reduced due to easy technique of its installation, remote data transmission from geophones to accompanying measuring modules, automated location of the system based on GPS and a highly automated method of seismic wave excitation. In this paper, the results of field tests carried out in different geological conditions were presented. The methodologies of acquisition, processing and interpretation of seismic and GPR measurements were broadly described. Seismograms and its spectrum registered by Seismobile system were compared to the ones registered by Geode seismograph of Geometrix. Seismic data processing and interpretation software allows for the obtaining of 2D/3D models of P- and S-wave velocities. Combined seismic and GPR results achieved sufficient imaging of shallow subsurface to a depth of over a dozen metres. The obtained geophysical information correlated with geological information from the boreholes with good quality. The results of performed tests proved the efficiency of the Seismobile system in seismic and GPR imaging of a shallow subsurface of transport routes under compound conditions.

  5. Seismic evaluation of existing liquid low level waste system at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Hammond, C.R.; Holmes, R.M.; Kincaid, J.H.; Singhal, M.K.; Stockdale, B.I.; Walls, J.C.; Webb, D.S.

    1993-01-01

    The existing liquid low level waste (LLLW) system at the Oak Ridge National Laboratory is used to collect, neutralize, concentrate, and store the radioactive and toxic waste from various sources at the Laboratory. The waste solutions are discharged from source facilities to individual collection tanks, transferred by underground piping to an evaporator facility for concentration, and pumped through the underground piping to storage in underground tanks. The existing LLLW system was installed in the 1950s with several system additions up to the present. The worst-case accident postulated is an earthquake of sufficient magnitude to rupture the tanks and/or piping so as to damage the containment integrity to the surrounding soil and environment. The objective of an analysis of the system is to provide a level of confidence in the seismic resistance of the LLLW system to withstand the postulated earthquake

  6. Frontal compression along the Apennines thrust system: The Emilia 2012 example from seismicity to crustal structure

    Science.gov (United States)

    Chiarabba, Claudio; De Gori, Pasquale; Improta, Luigi; Lucente, Francesco Pio; Moretti, Milena; Govoni, Aladino; Di Bona, Massimo; Margheriti, Lucia; Marchetti, Alessandro; Nardi, Anna

    2014-12-01

    The evolution of the Apennines thrust-and-fold belt is related to heterogeneous process of subduction and continental delamination that generates extension within the mountain range and compression on the outer front of the Adria lithosphere. While normal faulting earthquakes diffusely occur along the mountain chain, the sparse and poor seismicity in the compressional front does not permit to resolve the ambiguity that still exists about which structure accommodates the few mm/yr of convergence observed by geodetic data. In this study, we illustrate the 2012 Emilia seismic sequence that is the most significant series of moderate-to-large earthquakes developed during the past decades on the compressional front of the Apennines. Accurately located aftershocks, along with P-wave and Vp/Vs tomographic models, clearly reveal the geometry of the thrust system, buried beneath the Quaternary sediments of the Po Valley. The seismic sequence ruptured two distinct adjacent thrust faults, whose different dip, steep or flat, accounts for the development of the arc-like shape of the compressional front. The first shock of May 20 (Mw 6.0) developed on the middle Ferrara thrust that has a southward dip of about 30°. The second shock of May 29 (Mw 5.8) ruptured the Mirandola thrust that we define as a steep dipping (50-60°) pre-existing (Permo-Triassic) basement normal fault inverted during compression. The overall geometry of the fault system is controlled by heterogeneity of the basement inherited from the older extension. We also observe that the rupture directivity during the two main-shocks and the aftershocks concentration correlate with low Poisson ratio volumes, probably indicating that portions of the fault have experienced intense micro-damage.

  7. Rubber bearing and bitumen infill support system for seismic protection of nuclear power plant structures

    International Nuclear Information System (INIS)

    Ahmed, K.M.

    1981-01-01

    The prestressed concrete pressure vessel (PCPV) for the AGR is of cylindrical type. The whole of the reactor primary circuit is contained within the PCPV vault and includes the reactor core and support structures, boilers and gas circulators. The PCPV is essentially free standing on its foundation raft. In order to transmit gravitational and seismic loads between the PCPV and the foundation raft, a support system is used which consists of concentric rings of neoprene pads and a thin annulus of bitumen infill. In order to assess the importance of both stiffness and damping of the PCPV support system on the overall AGR response, detailed parametric studies were carried out using time-history dynamic analysis in conjunction with the modal superposition technique. The effects of both stiffness and damping are compared in terms of the maximum dynamic response (maximum accelerations and maximum relative displacements) and also floor response spectra at various locations on the nuclear island. It is clearly apparent from these investigations that for an appropriate range of structures on the nuclear island (such as the PCPV and its internals), greater reduction in seismic loading can be achieved by proper selection of stiffnesses and damping of the PCPV support system without resorting to strengthening techniques. (orig./HP)

  8. Energy-Based Design Criterion of Dissipative Bracing Systems for the Seismic Retrofit of Frame Structures

    Directory of Open Access Journals (Sweden)

    Gloria Terenzi

    2018-02-01

    Full Text Available Direct sizing criteria represent useful tools in the design of dissipative bracing systems for the advanced seismic protection of existing frame structures, especially when incorporated dampers feature a markedly non-linear behaviour. An energy-based procedure is proposed herein to this aim, focusing attention on systems including fluid viscous devices. The procedure starts by assuming prefixed reduction factors of the most critical response parameters in current conditions, which are evaluated by means of a conventional elastic finite element analysis. Simple formulas relating the reduction factors to the equivalent viscous damping ratio of the dampers, ξeq, are proposed. These formulas allow calculating the ξeq values that guarantee the achievement of the target factors. Finally, the energy dissipation capacity of the devices is deduced from ξeq, finalizing their sizing process. A detailed description of the procedure is presented in the article, by distinguishing the cases where the prevailing structural deficiencies are represented by poor strength of the constituting members, from the cases having excessive horizontal displacements. A demonstrative application to the retrofit design of a reinforced concrete gym building is then offered to explicate the steps of the sizing criterion in practice, as well as to evaluate the enhancement of the seismic response capacities generated by the installation of the dissipative system.

  9. Seismic Retrofit of Reinforced Concrete Frame Buildings with Hysteretic Bracing Systems: Design Procedure and Behaviour Factor

    Directory of Open Access Journals (Sweden)

    Antonio Di Cesare

    2017-01-01

    Full Text Available This paper presents a design procedure to evaluate the mechanical characteristics of hysteretic Energy Dissipation Bracing (EDB systems for seismic retrofitting of existing reinforced concrete framed buildings. The proposed procedure, aiming at controlling the maximum interstorey drifts, imposes a maximum top displacement as function of the seismic demand and, if needed, regularizes the stiffness and strength of the building along its elevation. In order to explain the application of the proposed procedure and its capacity to involve most of the devices in the energy dissipation with similar level of ductility demand, a simple benchmark structure has been studied and nonlinear dynamic analyses have been performed. A further goal of this work is to propose a simplified approach for designing dissipating systems based on linear analysis with the application of a suitable behaviour factor, in order to achieve a widespread adoption of the passive control techniques. At this goal, the increasing of the structural performances due to the addition of an EDB system designed with the above-mentioned procedure has been estimated considering one thousand case studies designed with different combinations of the main design parameters. An analytical formulation of the behaviour factor for braced buildings has been proposed.

  10. Experimental investigation of the seismic control of a nonlinear soil-structure system using MR dampers

    International Nuclear Information System (INIS)

    Li, Hui; Wang, Jian

    2011-01-01

    This paper reports the results of an experimental study conducted to demonstrate the feasibility and capability of magnetorheological (MR) dampers commanded by a decentralized control algorithm for seismic control of nonlinear civil structures considering soil-structure interaction (SSI). A two-story reinforced concrete (RC) frame resting in a laminar soil container is employed as the test specimen, and two MR dampers equipped in the first story are used to mitigate the response of this frame subjected to various intensity seismic excitations. A hyperbolic tangent function is used to represent the hysteretic behavior of the MR damper and a decentralized control approach for commanding MR dampers is proposed and implemented in the shaking table tests. Only the response of the first story is feedback for control command calculation of the MR dampers. The results indicate that the MR damper can effectively reduce the response of the soil-structure system, even when the soil-structure system presents complex nonlinear hysteretic behavior. The robustness of the proposed decentralized control algorithm is validated through the shaking table tests on the soil-structure system with large uncertainty. The most interesting findings in this paper are that MR dampers not only mitigate the superstructure response, but also reduce the soil response, pile response and earth pressure on the pile foundation

  11. Review: Marine Seismic And Side-Scan Sonar Investigations For Seabed Identification With Sonar System

    Directory of Open Access Journals (Sweden)

    Muhammad Zainuddin Lubis

    2017-06-01

    Full Text Available Marine seismic reflection data have been collected for decades and since the mid-to late- 1980s much of this data is positioned relatively accurately. Marine geophysical acquisition of data is a very expensive process with the rates regularly ship through dozens of thousands of euros per day. Acquisition of seismic profiles has the position is determined by a DGPS system and navigation is performed by Hypack and Maxview software that also gives all the offsets for the equipment employed in the survey. Examples of some projects will be described in terms of the project goals and the geophysical equipment selected for each survey and specific geophysical systems according to with the scope of work. For amplitude side scan sonar image, and in the multi-frequency system, color, becoming a significant properties of the sea floor, the effect of which is a bully needs to be fixed. The main confounding effect is due to absorption of water; geometric spread; shape beam sonar function (combined transmit-receive sonar beam intensity as a function of tilt angle obtained in this sonar reference frame; sonar vehicle roll; form and function of the seabed backscatter (proportion incident on the seabed backscattered signal to sonar as a function of the angle of incidence relative to the sea floor; and the slope of the seabed. The different angles of view are generated by the translation of the sonar, because of the discrete steps involved by the sequential pings, the angular sampling of the bottom.

  12. Adaptive neuro-fuzzy inference systems for semi-automatic discrimination between seismic events: a study in Tehran region

    Science.gov (United States)

    Vasheghani Farahani, Jamileh; Zare, Mehdi; Lucas, Caro

    2012-04-01

    Thisarticle presents an adaptive neuro-fuzzy inference system (ANFIS) for classification of low magnitude seismic events reported in Iran by the network of Tehran Disaster Mitigation and Management Organization (TDMMO). ANFIS classifiers were used to detect seismic events using six inputs that defined the seismic events. Neuro-fuzzy coding was applied using the six extracted features as ANFIS inputs. Two types of events were defined: weak earthquakes and mining blasts. The data comprised 748 events (6289 signals) ranging from magnitude 1.1 to 4.6 recorded at 13 seismic stations between 2004 and 2009. We surveyed that there are almost 223 earthquakes with M ≤ 2.2 included in this database. Data sets from the south, east, and southeast of the city of Tehran were used to evaluate the best short period seismic discriminants, and features as inputs such as origin time of event, distance (source to station), latitude of epicenter, longitude of epicenter, magnitude, and spectral analysis (fc of the Pg wave) were used, increasing the rate of correct classification and decreasing the confusion rate between weak earthquakes and quarry blasts. The performance of the ANFIS model was evaluated for training and classification accuracy. The results confirmed that the proposed ANFIS model has good potential for determining seismic events.

  13. 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.

  14. Seismic Symphonies

    Science.gov (United States)

    Strinna, Elisa; Ferrari, Graziano

    2015-04-01

    The project started in 2008 as a sound installation, a collaboration between an artist, a barrel organ builder and a seismologist. The work differs from other attempts of sound transposition of seismic records. In this case seismic frequencies are not converted automatically into the "sound of the earthquake." However, it has been studied a musical translation system that, based on the organ tonal scale, generates a totally unexpected sequence of sounds which is intended to evoke the emotions aroused by the earthquake. The symphonies proposed in the project have somewhat peculiar origins: they in fact come to life from the translation of graphic tracks into a sound track. The graphic tracks in question are made up by copies of seismograms recorded during some earthquakes that have taken place around the world. Seismograms are translated into music by a sculpture-instrument, half a seismograph and half a barrel organ. The organ plays through holes practiced on paper. Adapting the documents to the instrument score, holes have been drilled on the waves' peaks. The organ covers about three tonal scales, starting from heavy and deep sounds it reaches up to high and jarring notes. The translation of the seismic records is based on a criterion that does match the highest sounds to larger amplitudes with lower ones to minors. Translating the seismogram in the organ score, the larger the amplitude of recorded waves, the more the seismogram covers the full tonal scale played by the barrel organ and the notes arouse an intense emotional response in the listener. Elisa Strinna's Seismic Symphonies installation becomes an unprecedented tool for emotional involvement, through which can be revived the memory of the greatest disasters of over a century of seismic history of the Earth. A bridge between art and science. Seismic Symphonies is also a symbolic inversion: the instrument of the organ is most commonly used in churches, and its sounds are derived from the heavens and

  15. Tool for generation of seismic floor response spectra for secondary system design

    International Nuclear Information System (INIS)

    Cardoso, Tarcisio F.; Almeida, Andreia A. Diniz de

    2009-01-01

    The spectral analysis is still a valuable method to the seismic structure design, especially when one focalizes the topics of secondary systems in large industrial installations, as nuclear power plants. Two aspects of this situation add their arguments to recommend the use of this kind of analysis: the random character of the excitation and the multiplicity and the variability of the secondary systems. The first aspect can be managed if one assumes the site seismicity represented by a power spectrum density function of the ground acceleration, and then, by the systematic resolution of a first passage problem, to develop a uniformly probable response spectrum. The second one suggests also a probabilistic approach to the response spectrum in order to be representative all over the extensive group of systems with different characteristics, which can be enrolled in a plant. The present paper proposes a computational tool to achieve in-structure floor response spectra for secondary system design, which includes a probabilistic approach and considers coupling effects between primary and inelastic secondary systems. The analysis is performed in the frequency domain, with SASSI2000 system. A set of auxiliary programs are developed to consider three-dimensional models and their responses to a generic base excitation, acting in 3 orthogonal directions. The ground excitation is transferred to a secondary system SDOF model conveniently attached to the primary system. Then, a uniformly probable coupled response spectrum is obtained using a first passage analysis. In this work, the ExeSASSI program is created to manage SASSI2000 several modules and a set of auxiliary programs created to perform the probabilistic analyses. (author)

  16. 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

  17. Simplified inelastic seismic response analysis of piping system using improved capacity spectrum method

    International Nuclear Information System (INIS)

    Iijima, Tadashi

    2005-01-01

    We applied improved capacity spectrum method (ICSM) to a piping system with an asymmetric load-deformation relationship in a piping elbow. The capacity spectrum method can predict an inelastic response by balancing the structural capacity obtained from the load-deformation relationship with the seismic demand defined by an acceleration-displacement response spectrum. The ICSM employs (1) effective damping ratio and period that are based on a statistical methodology, (2) practical procedures necessary to obtain a balance between the structural capacity and the seismic demand. The effective damping ratio and period are defined so as to maximize the probability that predicted response errors lie inside the -10 to 20% range. However, without taking asymmetry into consideration the displacement calculated by using the load-deformation relationship on the stiffer side was 39% larger than that of a time history analysis by a direct integral method. On the other hand, when asymmetry was taken into account, the calculated displacement was only 14% larger than that of a time history analysis. Thus, we verified that the ICSM could predict the inelastic response with errors lying within the -10 to 20% range, by taking into account the asymmetric load-deformation relationship of the piping system. (author)

  18. Prioritization of information using decision support systems for seismic risk in Bucharest city

    Science.gov (United States)

    Armas, Iuliana; Gheorghe, Diana

    2014-05-01

    Nowadays, because of the ever increasing volume of information, policymakers are faced with decision making problems. Achieving an objective and suitable decision making may become a challenge. In such situations decision support systems (DSS) have been developed. DSS can assist in the decision making process, offering support on how a decision should be made, rather than what decision should be made (Simon, 1979). This in turn potentially involves a huge number of stakeholders and criteria. Regarding seismic risk, Bucharest City is highly vulnerable (Mandrescu et al., 2007). The aim of this study is to implement a spatial decision support system in order to secure a suitable shelter in case of an earthquake occurrence in the historical centre of Bucharest City. In case of a seismic risk, a shelter is essential for sheltering people who lost their homes or whose homes are in danger of collapsing while people at risk receive first aid in the post-disaster phase. For the present study, the SMCE Module for ILWIS 3.4 was used. The methodology included structuring the problem by creating a decision tree, standardizing and weighting of the criteria. The results showed that the most suitable buildings are Tania Hotel, Hanul lui Manuc, The National Bank of Romania, The Romanian Commercial Bank and The National History Museum.

  19. Mechanical design of a single-axis monolithic accelerometer for advanced seismic attenuation systems

    Energy Technology Data Exchange (ETDEWEB)

    Bertolini, Alessandro [Dipartimento di Fisica dell' Universita di Pisa and INFM, Largo Pontecorvo 2, I-56127 Pisa (Italy) and LIGO Project, California Institute of Technology, 1200 E. California Blvd., Pasadena, CA 91125 (United States)]. E-mail: alessandro.bertolini@desy.de; DeSalvo, Riccardo [LIGO Project, California Institute of Technology, 1200 E. California Blvd., Pasadena, CA 91125 (United States); Fidecaro, Francesco [Dipartimento di Fisica dell' Universita di Pisa and INFM, Largo Pontecorvo 2, I-56127 Pisa (Italy); Francesconi, Mario [Dipartimento di Fisica dell' Universita di Pisa and INFM, Largo Pontecorvo 2, I-56127 Pisa (Italy); Marka, Szabolcs [Department of Physics, Columbia University, 538 W. 120th St., New York, NY 10027 (United States); Sannibale, Virginio [LIGO Project, California Institute of Technology, 1200 E. California Blvd., Pasadena, CA 91125 (United States); Simonetti, Duccio [Dipartimento di Fisica dell' Universita di Pisa and INFM, Largo Pontecorvo 2, I-56127 Pisa (Italy); Takamori, Akiteru [LIGO Project, California Institute of Technology, 1200 E. California Blvd., Pasadena, CA 91125 (United States); Earthquake Research Institute, University of Tokyo, 1-1-1 Yayoi, Bunkyo-Ku, Tokyo 113-0032 (Japan); Tariq, Hareem [LIGO Project, California Institute of Technology, 1200 E. California Blvd., Pasadena, CA 91125 (United States)

    2006-01-15

    The design and mechanics for a new very-low noise low frequency horizontal accelerometer is presented. The sensor has been designed to be integrated in an advanced seismic isolation system for interferometric gravitational wave detectors. The motion of a small monolithic folded-pendulum (FP) is monitored by a high resolution capacitance displacement sensor; a feedback force actuator keeps the mass at the equilibrium position. The feedback signal is proportional to the ground acceleration in the frequency range 0-150Hz. The very high mechanical quality factor, Q{approx}3000 at a resonant frequency of 0.5Hz, reduces the Brownian motion of the proof mass of the accelerometer below the resolution of the displacement sensor. This scheme enables the accelerometer to detect the inertial displacement of a platform with a root-mean-square noise less than 1nm, integrated over the frequency band from 0.01 to 150Hz. The FP geometry, combined with the monolithic design, allows the accelerometer to be extremely directional. A vertical-horizontal coupling ranging better than 10{sup -3} has been achieved. A detailed account of the design and construction of the accelerometer is reported here. The instrument is fully ultra-high vacuum compatible and has been tested and approved for integration in seismic attenuation system of japanese TAMA 300 gravitational wave detector. The monolithic design also makes the accelerometer suitable for cryogenic operation.

  20. Experimental studies of the seismic response of structures incorporating base isolation systems

    International Nuclear Information System (INIS)

    Kelly, J.M.; Aiken, I.D.

    1989-01-01

    Whereas the concept of base isolating structures from the damaging effects of earthquake motions is not new, implementation of the technique is a relatively new occurrence. This has mainly been due to the need for several important developments in materials science and experimental and analytical modeling before base isolation could evolve into a practical approach for seismic design. One of these developments has been the ability to test large-scale isolation systems using simulated seismic loads. These tests have not only proven the performance and reliability of the isolation systems and hardware, but have enabled correlation studies to be undertaken which have confirmed the accuracy of analytical methods and the acceptability of current design procedures. The Earthquake Engineering Research Center (EERC) at the University of California at Berkeley has been an active participant in this work, and this paper reviews some of the achievements of the Center in the last few years. Component tests on single isolators are described. Tests on plain and high damping natural rubber bearings, lead-rubber bearings, sliding bearings, and bearings incorporating uplift resistance mechanisms have been performed. High-shear strain tests on large (up to full scale) elastomeric bearings have been conducted to determine the stability characteristics and limit states of the isolators

  1. Nonlinear Seismic Behavior of Different Boundary Conditions of Transmission Line Systems under Earthquake Loading

    Directory of Open Access Journals (Sweden)

    Li Tian

    2016-01-01

    Full Text Available Nonlinear seismic behaviors of different boundary conditions of transmission line system under earthquake loading are investigated in this paper. The transmission lines are modeled by cable element accounting for the nonlinearity of the cable. For the suspension type, three towers and two span lines with spring model (Model 1 and three towers and four span lines’ model (Model 2 are established, respectively. For the tension type, three towers and two span lines’ model (Model 3 and three towers and four span lines’ model (Model 4 are created, respectively. The frequencies of the transmission towers and transmission lines of the suspension type and tension type are calculated, respectively. The responses of the suspension type and tension type are investigated using nonlinear time history analysis method, respectively. The results show that the responses of the transmission tower and transmission line of the two models of the suspension type are slightly different. However, the responses of transmission tower and transmission line of the two models of the tension type are significantly different. Therefore, in order to obtain accurate results, a reasonable model should be considered. The results could provide a reference for the seismic analysis of the transmission tower-line system.

  2. New Site Coefficients and Site Classification System Used in Recent Building Seismic Code Provisions

    Science.gov (United States)

    Dobry, R.; Borcherdt, R.D.; Crouse, C.B.; Idriss, I.M.; Joyner, W.B.; Martin, G.R.; Power, M.S.; Rinne, E.E.; Seed, R.B.

    2000-01-01

    Recent code provisions for buildings and other structures (1994 and 1997 NEHRP Provisions, 1997 UBC) have adopted new site amplification factors and a new procedure for site classification. Two amplitude-dependent site amplification factors are specified: Fa for short periods and Fv for longer periods. Previous codes included only a long period factor S and did not provide for a short period amplification factor. The new site classification system is based on definitions of five site classes in terms of a representative average shear wave velocity to a depth of 30 m (V?? s). This definition permits sites to be classified unambiguously. When the shear wave velocity is not available, other soil properties such as standard penetration resistance or undrained shear strength can be used. The new site classes denoted by letters A - E, replace site classes in previous codes denoted by S1 - S4. Site classes A and B correspond to hard rock and rock, Site Class C corresponds to soft rock and very stiff / very dense soil, and Site Classes D and E correspond to stiff soil and soft soil. A sixth site class, F, is defined for soils requiring site-specific evaluations. Both Fa and Fv are functions of the site class, and also of the level of seismic hazard on rock, defined by parameters such as Aa and Av (1994 NEHRP Provisions), Ss and S1 (1997 NEHRP Provisions) or Z (1997 UBC). The values of Fa and Fv decrease as the seismic hazard on rock increases due to soil nonlinearity. The greatest impact of the new factors Fa and Fv as compared with the old S factors occurs in areas of low-to-medium seismic hazard. This paper summarizes the new site provisions, explains the basis for them, and discusses ongoing studies of site amplification in recent earthquakes that may influence future code developments.

  3. Using the automized system ''section'' to forecast velocity sections using data on borehole velocity measurement and seismic field prospecting

    Energy Technology Data Exchange (ETDEWEB)

    Dorman, M.I.; Gein, F.F.; Zubairov, F.B.

    1981-01-01

    A system of automated processing of seismic data is examined which makes it possible to set up rate functions at arbitrary points of a seismic prospecting section or at points conciding with boreholes in which rate measurements have not been completed. The basis for the forecasting method is data on seismic well logging investigations, seismic prospecting and some indirect observations on sections. The bases of a procedure realizing a forecasting method are set forth, as are those requirements which satisfy the system as a whole. The results of using the ''section'' system in a terrestrial section of Western Siberia are set forth.

  4. Application of new developments in coupled seismic analysis of piping systems

    International Nuclear Information System (INIS)

    Gupta, A.; Gupta, A.K.

    1995-01-01

    The current practice of calculating the seismic response is to perform the analysis of the primary structure (buildings) and the secondary systems (piping) separately. Earthquake input to the primary system in terms of a design response spectrum. An acceleration time history compatible with the design response spectrum is developed (a non-unique process) and primary system is analyzed to obtain the acceleration histories at the desired floors. Floor time histories are used for generating the corresponding instructure response spectrum (IRIS). The instructure response spectra are used as input at the supports of secondary systems. Further, in case of multiple supports, an envelope spectrum (introducing conservatism) is obtained from the individual support IRS. The effect of relative support motion is incorporated by a worst-case separate static analysis (adding to the conservatism). In the above method, mass interaction between the secondary and primary system is ignored, which may have significant effect at resonant frequencies (further adding to the conservatism). The calculated response may be an order of magnitude higher than they should be. Two computer programs, CREST and CREST-IRIS, were developed at Center for NUclear Power Plant Structures, Equipment and Piping. Any one of the two computer programs together with a piping analysis program can be used to perform an accurate coupled seismic analysis of piping systems. The two computer programs have been validated against the time history analysis for simple problems. In the present study, we have applied CREST to analyze two real-life piping systems. The piping analysis program used in this research is the commercial software PIPESTRESS, developed by DST Computer Services of Geneva, Switzerland. (author). 4 refs., 3 figs., 2 tabs

  5. 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

  6. 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.

  7. The results of the Seismic Alert System of Mexico SASMEX, during the earthquakes of 7 and 19 of September 2017

    Science.gov (United States)

    Espinosa Aranda, J. M., Sr.; Cuellar Martinez, A.

    2017-12-01

    The Seismic Alert System of Mexico, SASMEX began in 1991, is integrated by the seismic alert system of Mexico City and the seismic alert system of Oaxaca. SASMEX has 97 seismic sensors which are distributed in the seismic regions of the Pacific coast and the South of the Trans-Mexican Volcanic Belt of states of Jalisco, Colima, Michoacán, Guerrero, Oaxaca and Puebla. The alert dissemination covers the cities of: Acapulco, Chilpancingo, Morelia, Puebla, Oaxaca, Toluca and Mexico City, reaching the earthquake warnings to more than 25 millions of people. SASMEX has detected correctly more than 5600 earthquakes and warned 156. Mexico City has different alert dissemination systems like several Radio and Tv commercial broadcasters, dedicated radio receivers, EAS-SAME-SARMEX radio receivers and more tha 6700 public loud speakers. The other cities have only some of those systems. The Mw 8.2 Chiapas earthquake on September 7, despite the epicentral distance far of the first seismic detections (more than 180 km) and the low amplitudes of the P waves, the earthquake warning time gave more than 90 seconds to Mexico City before the arrivals of S waves with minor damages to the city in contrast with high damages in towns in the coast. This earthquake offered an opportunity to show the developments and lacks to reduce the risk, such as the need to increase the seismic detection coverage and the earthquake warning dissemination in towns with high seismic vulnerability. The Mw 7.1 Morelos earthquake on September 19 caused thousands of damages and hundreds of deaths and injuries in Mexico City, this earthquake is the second with the most damages after the Mw 8.1 Michoacán earthquake of September 19 on 1985. The earthquake early warning gave 11 seconds after the arrivals of S waves, however the activation occurred few seconds after the P waves arrives to Mexico City, and due to the seismic focus was near to the city, the P waves were felt for the people. The Accelerographic Network

  8. REWSET: A prototype seismic