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Sample records for auxiliary feedwater system

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.

  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. 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

  10. 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

  11. 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)

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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

  5. Steam generator auxiliary systems

    International Nuclear Information System (INIS)

    Heinz, A.

    1982-01-01

    The author deals with damage and defect types obtaining in auxiliary systems of power plants. These concern water/steam auxiliary systems (feed-water tank, injection-control valves, slide valves) and air/fluegas auxiliary systems (blowers, air preheaters, etc.). Operating errors and associated damage are not dealt with; by contrast, weak spots are pointed out which result from planning and design. Damage types and events are collected in statistics in order to facilitate damage evaluation for arriving at improved design solutions. (HAG) [de

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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)

  3. 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

  4. Aging and service wear of auxiliary feedwater pumps for PWR nuclear power plants

    International Nuclear Information System (INIS)

    Greenstreet, W.L.

    1989-01-01

    This paper describes investigations on auxiliary feedwater pumps being done under the Nuclear Plant Aging Research (NPAR) Program. Objectives of these studies are: to identify and evaluate practical, cost-effective methods for detecting, monitoring, and assessing the severity of time-dependent degradation (aging and service wear); recommend inspection and maintenance practices; establish acceptance criteria; and help facilitate use of the results. Emphasis is given to identifying and assessing methods for detecting failure in the incipient stage and to developing degradation trends to allow timely maintenance, repair or replacement actions. 3 refs

  5. 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

  6. Feedwater control system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

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

  7. Auxiliary feedwater system risk-based inspection guide for the 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

  8. 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)

  9. 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)

  10. 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.)

  11. Feedwater control system in nuclear power plants

    International Nuclear Information System (INIS)

    Masuyama, Hideo.

    1981-01-01

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

  12. 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.)

  13. 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

  14. 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

  15. 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)

  16. 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

  17. 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

  18. 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)

  19. 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.

  20. 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.)

  1. 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

  2. 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

  3. 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.)

  4. 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

  5. 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

  6. 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.)

  7. 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)

  8. 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.)

  9. 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

  10. 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)

  11. 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.

  12. 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.)

  13. 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

  14. 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)

  15. 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)

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

    International Nuclear Information System (INIS)

    1985-07-01

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

  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. Regulatory analysis for the resolution of Generic Issue 125.II.7 ''Reevaluate Provision to Automatically Isolate Feedwater from Steam Generator During a Line Break''

    International Nuclear Information System (INIS)

    Basdekas, D.L.

    1988-09-01

    Generic Issue 125.II.7 addresses the concern related to the automatic isolation of auxiliary feedwater (AFW) to a steam generator with a broken steam or feedwater line. This regulatory analysis provides a quantitative assessment of the costs and benefits associated with the removal of the AFW automatic isolation and concludes that no new regulatory requirements are warranted. 21 refs., 7 tabs

  19. 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

  20. 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

  1. 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.

  2. Improvement of Equipment reliability for Auxiliary Feed Water System

    Energy Technology Data Exchange (ETDEWEB)

    Deok, Lee Sang; Kwan, Lee Yong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2013-10-15

    According to AP913 ER) of INPO, Number of the event related to equipment is higher than others like external or human performance. In the top 25 systems, Auxiliary feed water system is the seventh highest among systems. AWFS consists of many component and complex system and Main Function of AFWS is to supply feedwater to the steam generators for the removal of heat from the RCS(Reactor Coolant System) in event the main feedwater system is unavailable following a transient or accident. Reliability of component means how well operate on demands and monitoring is necessary to keep track of condition of component. If component performance is lower than the required value, corrective action for failure mode should be done. The objective of this study is focused to improve of AF pump by adding the tasks of SHR(System Health Report) into the task of system engineer walkdown of PMT(Preventive Maintenance Template). Increasing the reliability of AF pump will contribute to improvement of reliability of AFWS. Based on operating history, there was high vibration of AF pump during performance test. In that case, there were a lot of maintenance works for normal operation of AF pump. Vibration problem related pump can't be detected by tasks of SE walkdown because it's not running during normal operation except for surveillance test. CHR(Component Health Report) of AF pump in AFWS can be made from necessary part which means monitoring and functional failure because problem of Stand-by pump can be covered by conducting monitoring and analysis of functional failure. To improve reliability of AF pump, walkdown of PMT and SHR should be conducted both in accordance with surveillance test frequency. Health of AF pump based on operation history can be verified first and then can find out which parts of pump are weak. Finally, weak part can be managed intensively and failure can be reduced according to SE walkdown. But this work can be risky and burdensome because all parts of

  3. Proceedings: Power Plant Electric Auxiliary Systems Workshop

    International Nuclear Information System (INIS)

    1992-06-01

    The EPRI Power Plant Electric Auxiliary Systems Workshop, held April 24--25, 1991, in Princeton, New Jersey, brought together utilities, architect/engineers, and equipment suppliers to discuss common problems with power plant auxiliary systems. Workshop participants presented papers on monitoring, identifying, and solving problems with auxiliary systems. Panel discussions focused on improving systems and existing and future plants. The solutions presented to common auxiliary system problems focused on practical ideas that can enhance plant availability, reduce maintenance costs, and simplify the engineering process. The 13 papers in these proceedings include: Tutorials on auxiliary electrical systems and motors; descriptions of evaluations, software development, and new technologies used recently by electric utilities; an analysis of historical performance losses caused by power plant auxiliary systems; innovative design concepts for improving auxiliary system performance in future power plants

  4. 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

  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. 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

  7. 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)

  8. 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

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

    International Nuclear Information System (INIS)

    Faure, Celine; Eade, Kevin; Fontan, Guillaume

    2012-09-01

    offered increased operational flexibility. Its use is limited to the start-up of a unit up to the point where the chemical criteria are met for the changeover from auxiliary feedwater to final feedwater and, at most, up to 20% Pn. In 2006, the will to optimise the discharge during start-up in order to increase the availability of the waste tanks, and to reduce the quantity of metallic oxides in the SGs at the start-up following the appearance of the clogging of SGs broached-hole tube support plates, lead EDF to recreate identical SMEs for the rest of the fleet. Using the SME at the start-up is beneficial for safety, availability, lifespan and discharge reduction. This paper develops the context and technical choices which lead EDF to implement the SME; it goes on to present the design of the treatment system, as well as the implementation feedback of its use on a number of plants. (authors)

  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. 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)

  12. IE Information Notice No. 85-75: Improperly installed instrumentation, inadequate quality control and inadequate postmodification testing

    International Nuclear Information System (INIS)

    Jordan, E.L.

    1992-01-01

    On June 10, 1985, the licensee informed the NRC Resident Inspector that for approximately 5 days LaSalle Unit 2 had been without the capability of automatic actuation of emergency core cooling (ECCS) and that for approximately 3 days during this period the plant had been without secondary containment integrity. The major cause of this condition was improper installation (the variable and reference legs were reversed) of the two reactor vessel level actuation switches which control Division 1 automatic depressurization system (ADS), low pressure core spray (LPCS), and reactor core isolation cooling (RCIC). On July 20, 1985, the Trojan Nuclear Power Plant tripped from 100% power because of a turbine trip that was caused by the loss of the unit auxiliary transformer. All systems functioned normally except that low suction pressure caused one auxiliary feedwater pump to trip and then the other auxiliary feedwater pump to trip after restart of the first auxiliary feedwater pump. The cause of the trips of the auxiliary feedwater pumps can be traced back to improper postmodification adjustment and inadequate postmodification testing following retrofit of environmentally qualified controllers for the auxiliary feedwater system. The auxiliary feedwater pump trips on low suction pressure were caused by excessive combined flow from the two auxiliary feedwater pumps that draw from a single header from the condensate storage tank. The flow control valves were open farther than required after new environmentally qualified controllers had been installed during a recent refueling outage

  13. 30 CFR 57.8529 - Auxiliary fan systems

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Auxiliary fan systems 57.8529 Section 57.8529 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE... Underground Only § 57.8529 Auxiliary fan systems When auxiliary fan systems are used, such systems shall...

  14. Comparison of auxiliary feedwater and EDRS operation during natural circulation of MRX

    International Nuclear Information System (INIS)

    Kim, Jae Hak; Park, Goon Cherl

    1997-01-01

    The MRX is an integral type ship reactor with 100 MWt power, which is designed by Japan Atomic Energy Research Institute. It is characterized by integral type PWR, in-vessel type control rod drive mechanism, water-filled containment vessel and passive decay heat removal system. Marine reactor should have high passive safety. Therefore, in this study, we simulated the loss of flow accident to verify the passive decay heat removal by natural circulation using RETRAN-03 code. auxiliary feed water systems are used for decay heat removal mechanism and results are compared with the loss of flow accident analysis using emergency decay heat removal system by JAERI. Results are very similar to case of EDRS 1 loop operation in JAERI analysis and decay heat is successfully removed by natural circulation

  15. 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

  16. 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)

  17. 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)

  18. 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)

  19. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  1. 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.

  2. 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

  3. CAREM-25. Auxiliary systems

    International Nuclear Information System (INIS)

    Acosta, Eduardo; Amaya, Daniel; Carlevaris, Rodolfo; Patrignani, A.; Ramilo, L.; Santecchia, A.; Vindrola, C.

    2000-01-01

    CAREM is an innovative PWR reactor whose prototype will be of small power generation capacity (100MWt, about 25MWe).CAREM design is based on light water integrated reactor with slightly enriched uranium.In this work, a summary of the functions and most relevant design characteristics of main auxiliary systems associated to the chain of heat removal and physicochemical - radiological treatment of the cooling fluids of the CAREM-25 prototype is presented.Even though these auxiliary systems of the reactor are not safety system, they fulfill functions related with the nuclear safety at different operative modes of the reactor

  4. CAREM-25. Auxiliary systems

    International Nuclear Information System (INIS)

    Acosta, Eduardo; Amaya, Daniel; Carlevaris, Rodolfo; Patrignani, Alberto; Santecchia, Alberto; Vindrola, Carlos; Ramilo, Lucia B.

    2000-01-01

    CAREM is an innovative PWR reactor whose prototype will be of small power generation capacity (100 M Wt, about 25 M We). CAREM design is based on light water integrated reactor with slightly enriched uranium. In this work, a summary of the functions and most relevant design characteristics of main auxiliary systems associated to the chain of heat removal and physicochemical - radiological treatment of the cooling fluids of the CAREM-25 prototype is presented. Even though these auxiliary systems of the reactor are not safety system, they fulfill functions related with the nuclear safety at different operative modes of the reactor. (author)

  5. 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

  6. Lead corrosion and transport in simulated secondary feedwater

    Energy Technology Data Exchange (ETDEWEB)

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

    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)

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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)

  12. The installation of helium auxiliary systems in HTGR

    International Nuclear Information System (INIS)

    Qin Zhenya; Fu Xiaodong

    1993-01-01

    The inert gas Helium was chosen as reactor coolant in high temperature gas coolant reactor, therefore a set of Special and uncomplex helium auxiliary systems will be installed, the safe operation of HTR-10 can be safeguarded. It does not effect the inherent safety of HTR-10 MW if any one of all those systems were damaged during operation condition. This article introduces the design function and the system principle of all helium auxiliary systems to be installed in HTR-10. Those systems include: helium purification and its regeneration system, helium supply and storage system, pressure control and release system of primary system, dump system for helium auxiliary system and fuel handling, gaseous waste storage system, water extraction system for helium auxiliary systems and evacuation system for primary system

  13. Major factors in critical equipment reliability - Auxiliary systems; The development of an auxiliary system

    International Nuclear Information System (INIS)

    Forsthoffer, W.E.

    1992-01-01

    In this article, the author details the development of an actual auxiliary system in order to fully understand the function of each major component and how it contributes to the total operation and reliability of the system. Only after the function of an auxiliary system is thoroughly understood, can one proceed to discuss specifications, design audits, testing, operation and preventive maintenance. The application selected will be to develop a pressurized lubrication and steam turbine control oil system for the critical equipment unit. This example was selected since many readers will be familiar with this type and because it provides a good foundation towards understanding fluid sealing systems. In the exercise that follow, he will define the system requirements and determine the system parameters. This information will then be used for component sizing

  14. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

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

  15. 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...

  16. 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

  17. Operation auxiliary system (SAO)

    International Nuclear Information System (INIS)

    Lolich, J.; Santome, D.; Drexler, J.

    1990-01-01

    This work presents an auxiliary system for nuclear power plants operation (SAO). The development purpose consisted in a computing supervision system to be installed at different sites of a reactor, mainly in the control room. The inclusion of this system to a nuclear power plant minimizes the possibility of human error for the facility operation. (Author) [es

  18. 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

  19. CRBRP decay heat removal systems

    International Nuclear Information System (INIS)

    Hottel, R.E.; Louison, R.; Boardman, C.E.; Kiley, M.J.

    1977-01-01

    The Decay Heat Removal Systems for the Clinch River Breeder Reactor Plant (CRBRP) are designed to adequately remove sensible and decay heat from the reactor following normal shutdown, operational occurrences, and postulated accidents on both a short term and a long term basis. The Decay Heat Removal Systems are composed of the Main Heat Transport System, the Main Condenser and Feedwater System, the Steam Generator Auxiliary Heat Removal System (SGAHRS), and the Direct Heat Removal Service (DHRS). The overall design of the CRBRP Decay Heat Removal Systems and the operation under normal and off-normal conditions is examined. The redundancies of the system design, such as the four decay heat removal paths, the emergency diesel power supplies, and the auxiliary feedwater pumps, and the diversities of the design such as forced circulation/natural circulation and AC Power/DC Power are presented. In addition to overall design and system capabilities, the detailed designs for the Protected Air Cooled Condensers (PACC) and the Air Blast Heat Exchangers (ABHX) are presented

  20. 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

  1. 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.)

  2. Nuclear reactor auxiliary heat removal system

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  3. 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.

  4. Auxiliary System Load Schemes in Large Thermal and Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kuzle, I.; Bosnjak, D.; Pandzic, H.

    2010-01-01

    Uninterrupted auxiliary system power supply in large power plants is a key factor for normal operation, transient states, start-ups and shutdowns and particularly during fault conditions. Therefore, there are many challenges in designing the main electrical system as well as the auxiliary systems power supply. Depending upon the type of fuel used and the environmental control system required, a thermal power plant may consume as much as 10% of its total generation for auxiliary power, while a nuclear power plant may require only 4 - 6% auxiliaries. In general, the larger the power generating plant, the higher the voltage selected for the AC auxiliary electric system. Most stations in the 75 to 500 MW range utilize 4,2 kV as the base auxiliary system voltage. Large generating stations 500 - 1000 MW and more use voltage levels of 6,9 kV and more. Some single dedicated loads such as electric driven boiler feed pumps are supplied ba a 13,8 kV bus. While designing the auxiliary electric system, the following areas must be considered: motor starting requirements, voltage regulation requirements, short-circuit duty requirements, economic considerations, reliability and alternate sources. Auxiliary power supply can't be completely generalized and each situation should be studied on its own merits to determine the optimal solution. Naturally, nuclear power plants have more reliability requirements and safety design criteria. Main coolant-pump power supply and continuity of service to other vital loads deserve special attention. This paper presents an overview of some up-to-date power plant auxiliary load system concepts. The main types of auxiliary loads are described and the electric diagrams of the modern auxiliary system supply concepts are given. Various alternative sources of auxiliary electrical supply are considered, the advantages and disadvantages of these are compared and proposals are made for high voltage distribution systems around the thermal and nuclear plant

  5. Energy consumption of auxiliary systems of electric cars

    Directory of Open Access Journals (Sweden)

    Evtimov Ivan

    2017-01-01

    Full Text Available The paper analyzes the power demand of the auxiliary systems of electric cars. On the basis of existing electric cars an analysis of energy consumption of different auxiliary systems is done. As a result possibilities for rational use of these systems have been proposed, which can increase the mileage per one charge of the battery.

  6. 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)

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  8. 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...

  9. 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

  10. Safety device and machine system of nuclear power plant

    International Nuclear Information System (INIS)

    1978-10-01

    It introduces principle and kinds of heat power including heat balance and nuclear power. It explains a lot of technical terms about the nuclear power system, which are primary loop, reactor, steam generator, primary coolant pump and pressurizer in PWR, chemical and volume control system, component cooling system, safety injection system, and spent fuel cooling and storage system in auxiliary system, liquid solid and gaseous waste disposal system in radwaste disposal, gland sealing system, turbine instrumentation, turning gear, hydrogen cooling system, condenser, feedwater heater, degenerate heater, auxiliary heat exchanger, centrifugal pump, rotary reciprocating and tank and pressure vessel.

  11. 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

  12. 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.)

  13. 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)

  14. Probabilistic cloning with supplementary information contained in the quantum states of two auxiliary systems

    International Nuclear Information System (INIS)

    Li, Lvjun; Qiu, Daowen

    2007-01-01

    In probabilistic cloning with two auxiliary systems, we consider and compare three different protocols for the success probabilities of cloning. We show that, in certain circumstances, it may increase the success probability to add an auxiliary system to the probabilistic cloning machine having one auxiliary system, but we always can find another cloning machine with one auxiliary system having the same success probability as that with two auxiliary systems

  15. 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

  16. 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)

  17. Development of KALIMER auxiliary sodium and cover gas management system

    International Nuclear Information System (INIS)

    Kwon, Sang Woon; Hwang, Sung Tae

    1996-11-01

    The objectives of this report are to develop and to describe the auxiliary liquid metal and cover gas management systems of KALIMER. the system includes following system: (1) Auxiliary liquid metal system (2) Inert gas receiving and processing system (3) Impurity monitoring and analysis system. Auxiliary liquid metal and cover gas management system of KALIMER was developed. Functions of each systems and design basis were describes. The auxiliary liquid metal system receives, transfers, and purifies all sodium used in the plant. The system furnishes the required sodium quantity at the pressure, temperature, flow rate, and purity specified by the interfacing system. The intermediated sodium processing subsystem (ISPS) provides continuous purification of IHTS sodium, as well as performs the initial fill operation for both the IHTS and reactor vessel. The primary sodium processing subsystem provides purification (cold trapping) for sodium used in the reactor vessel. The inert gas receiving and processing (IGRP) system provides liquefied and ambient gas storage, delivers inert gases of specified composition and purity at regulated flow rates and pressures to points of usage throughout the KALIMER, and accepts the contaminated gases through its vacuum facilities for storage and transfer to the gas radwaste system. Three gases are used in the KALIMER: helium, argon, and nitrogen. 11 tabs., 12 figs. (Author)

  18. Development of KALIMER auxiliary sodium and cover gas management system

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Sang Woon; Hwang, Sung Tae [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-11-01

    The objectives of this report are to develop and to describe the auxiliary liquid metal and cover gas management systems of KALIMER. the system includes following system: (1) Auxiliary liquid metal system (2) Inert gas receiving and processing system (3) Impurity monitoring and analysis system. Auxiliary liquid metal and cover gas management system of KALIMER was developed. Functions of each systems and design basis were describes. The auxiliary liquid metal system receives, transfers, and purifies all sodium used in the plant. The system furnishes the required sodium quantity at the pressure, temperature, flow rate, and purity specified by the interfacing system. The intermediated sodium processing subsystem (ISPS) provides continuous purification of IHTS sodium, as well as performs the initial fill operation for both the IHTS and reactor vessel. The primary sodium processing subsystem provides purification (cold trapping) for sodium used in the reactor vessel. The inert gas receiving and processing (IGRP) system provides liquefied and ambient gas storage, delivers inert gases of specified composition and purity at regulated flow rates and pressures to points of usage throughout the KALIMER, and accepts the contaminated gases through its vacuum facilities for storage and transfer to the gas radwaste system. Three gases are used in the KALIMER: helium, argon, and nitrogen. 11 tabs., 12 figs. (Author).

  19. 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

  20. 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.

  1. Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

    International Nuclear Information System (INIS)

    Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

    1995-01-01

    Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

  2. 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.

  3. Development of Intelligent Auxiliary System for Customized Physical Fitness and Healthcare

    Directory of Open Access Journals (Sweden)

    Huang Chung-Chi

    2016-01-01

    Full Text Available With the advent of global high-tech industry and commerce era, the sedentary reduces opportunities of physical activity. And physical fitness and health of people is getting worse and worse. At present, the shortage of physical fitness instructors greatly affected the effectiveness of health promotion. Therefore, it is necessary to develop an auxiliary system which can reduce the workload of instructors and enhance physical fitness and health for people. But current general physical fitness and healthcare system is hard to meet individualized needs. The main purpose of this research is to develop an intelligent auxiliary system for customized physical fitness and healthcare. It records all processes of physical fitness and healthcare system by wireless sensors network. The results of intelligent auxiliary systems for customized physical fitness and healthcare will be generated by fuzzy logic Inference. It will improve individualized physical fitness and healthcare. Finally, we will demonstrate the advantages of the intelligent auxiliary system for customized physical fitness and healthcare.

  4. Mechanics of integrating root causes into PRAs

    International Nuclear Information System (INIS)

    Bruske, S.Z.; Cadwallader, L.C.; Stepina, P.L.; Vesely, W.E.

    1985-01-01

    This paper presents a derivation of root cause importance, root cause data for selected components of a pressurized water reactor auxiliary feedwater system, an Accident Sequence Evaluation Program (ASEP) auxiliary feedwater system model, and the results of root cause importance calculations. The methodology shown herein is straightforward and is easily applied to existing probabilistic risk assessments. Root cause importance can greatly benefit the areas of design, maintenance, and inspection. Root cause importance for various components and circumstances can be evaluated

  5. 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

  6. 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.)

  7. Auxiliary cooling device for power plant

    International Nuclear Information System (INIS)

    Yamanoi, Kozo.

    1996-01-01

    An auxiliary cooling sea water pipeline for pumping up cooling sea water, an auxiliary cooling sea water pipeline and a primary side of an auxiliary cooling heat exchanger are connected between a sea water taking vessel and a sea water discharge pit. An auxiliary cooling water pump is connected to an auxiliary water cooling pipeline on the second side of the auxiliary cooling heat exchanger. The auxiliary cooling water pipeline is connected with each of auxiliary equipments of a reactor system and each of auxiliary equipments of the turbine system connected to a turbine auxiliary cooling water pipeline in parallel. During ordinary operation of the reactor, heat exchange for each of the auxiliary equipments of the reactor and heat exchange for each of the equipments of the turbine system are conducted simultaneously. Since most portions of the cooling devices of each of the auxiliary equipments of the reactor system and each of the auxiliary equipments of the turbine system can be used in common, the operation efficiency of the cooling device is improved. In addition, the space for the pipelines and the cost for the equipments can be reduced. (I.N.)

  8. 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

  9. Experimental fast reactor JOYO MK-III functional test. Primary auxiliary cooling system test

    International Nuclear Information System (INIS)

    Karube, Koji; Akagi, Shinji; Terano, Toshihiro; Onuki, Osamu; Ito, Hideaki; Aoki, Hiroshi; Odo, Toshihiro

    2004-03-01

    This paper describes the results of primary auxiliary cooling system, which were done as a part of JOYO MK-III function test. The aim of the tests was to confirm the operational performance of primary auxiliary EMP and the protection system including siphon breaker of primary auxiliary cooling system. The items of the tests were: (Test No.): (Test item). 1) SKS-117: EMP start up test. 2) SKS-118-1: EMP start up test when pony motor running. 3) SKS-121: Function test of siphon breaker. The results of the tests satisfied the required performance, and demonstrated successful operation of primary auxiliary cooling system. (author)

  10. Cooling system for auxiliary systems of a nuclear power plant

    International Nuclear Information System (INIS)

    Maerker, W.; Mueller, K.; Roller, W.

    1981-01-01

    From the reactor auxiliary and ancillary systems of a nuclear facility heat has to be removed without the hazard arising that radioactive liquids or gases may escape from the safe area of the nuclear facility. A cooling system is described allowing at every moment to make available cooling fluid at a temperature sufficiently low for heat exchangers to be able to remove the heat from such auxiliary systems without needing fresh water supply or water reservoirs. For this purpose a dry cooling tower is connected in series with a heat exchanger that is cooled on the secondary side by means of a refrigerating machine. The cooling pipes are filled with a nonfreezable fluid. By means of a bypass a minimum temperature is guaranteed at cold weather. (orig.) [de

  11. 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

  12. 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

  13. Entanglement Capacity of Two-Qubit Unitary Operator with the Help of Auxiliary System

    International Nuclear Information System (INIS)

    Hu Baolin; Di Yaomin

    2007-01-01

    The entanglement capacity of general two-qubit unitary operators is studied when auxiliary systems are allowed, and the analytical results based on linear entropy when input states are disentangled are given. From the results the condition for perfect entangler, α 1 = α 2 = π/4, is obtained. Contrary to the case without auxiliary system, the parameter α 3 may play active role to the entanglement capacity when auxiliary systems are allowed.

  14. Aiming of Kirkpatrick-Baez microscope based on auxiliary optical system

    International Nuclear Information System (INIS)

    Huang Shengling; Mu Baozhong; Yi Shengzhen; Wang Xin; Wang Zhanshan; Ding Yongkun; Miao Wenyong; Dong Jianjun

    2009-01-01

    An auxiliary optical system has been designed, which can provide precise positioning for aiming Kirkpatrick-Baez (KB) microscope object location. An 8 keV X-ray imaging system by KB microscope with periodic multilayer films has been designed. The field of view and depth of field in the resolution of 5 μm are got, and then the corresponding point and depth of field in diagnostic experiments are calculated. Based on the object-image relations and precision of the KB microscope, an auxiliary visible light imaging system is designed and X-ray imaging experiments are performed, which can achieve equivalent aiming between the visible imaging system and the KB microscope. The results show that ±20 μm vertical axis plane and ±300 μm axial accuracy are achieved through the auxiliary optical path, which can meet the object point positioning requirements of the KB microscope. (authors)

  15. PWR auxiliary systems, safety and emergency systems, accident analysis, operation

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1976-01-01

    The author presents a description of PWR auxiliary systems like volume control, boric acid control, coolant purification, -degassing, -storage and -treatment system and waste processing systems. Residual heat removal systems, emergency systems and containment designs are discussed. As an accident analysis the author gives a survey over malfunctions and disturbances in the field of reactor operations. (TK) [de

  16. Auxiliary reactor for a hydrocarbon reforming system

    Science.gov (United States)

    Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

    2006-01-17

    An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

  17. 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)

  18. 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)

  19. 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

  20. 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

  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. CADRIGS--computer aided design reliability interactive graphics system

    International Nuclear Information System (INIS)

    Kwik, R.J.; Polizzi, L.M.; Sticco, S.; Gerrard, P.B.; Yeater, M.L.; Hockenbury, R.W.; Phillips, M.A.

    1982-01-01

    An integrated reliability analysis program combining graphic representation of fault trees, automated data base loadings and reference, and automated construction of reliability code input files was developed. The functional specifications for CADRIGS, the computer aided design reliability interactive graphics system, are presented. Previously developed fault tree segments used in auxiliary feedwater system safety analysis were constructed on CADRIGS and, when combined, yielded results identical to those resulting from manual input to the same reliability codes

  3. 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

  4. 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

  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. 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

  7. Progress in the integration of the ITER plant systems in auxiliary buildings

    International Nuclear Information System (INIS)

    Kotamäki, M.; Cordier, J.-J.; Kuehn, I.; Perrin, J.-L.; Sweeney, S.; Villedary, B.

    2016-01-01

    Highlights: • Usage of 3D CAD model in ITER configuration management presented. • 3D CAD models efficient in configuration and interface management. • Costly and schedule delaying changes avoided with proper interface management. • ITER buildings construction progressing. - Abstract: The ITER Tokamak machine is located in the center of Tokamak complex buildings consisting of Tokamak, Diagnostic, and Tritium buildings. Around the Tokamak complex there are over 30 auxiliary buildings housing various plant systems serving the Tokamak machine either directly or indirectly. The layout and space allocation of each auxiliary building and plant systems housed by the building are represented in the so-called Configuration Management Models (CMM). These are light 3D CAD models that define the required space envelope and the physical interfaces between the systems and the buildings and in-between the systems. The paper describes the CMM and interface management processes of the ITER auxiliary buildings and plant systems, and discusses the preparations for the plant installation phase. In addition, the current baseline configuration of the ITER plant systems in auxiliary buildings is described together with the recent developments in the configuration of different systems, as well as the current status of the construction of the buildings.

  8. Progress in the integration of the ITER plant systems in auxiliary buildings

    Energy Technology Data Exchange (ETDEWEB)

    Kotamäki, M., E-mail: miikka.kotamaki@iter.org; Cordier, J.-J.; Kuehn, I.; Perrin, J.-L.; Sweeney, S.; Villedary, B.

    2016-11-01

    Highlights: • Usage of 3D CAD model in ITER configuration management presented. • 3D CAD models efficient in configuration and interface management. • Costly and schedule delaying changes avoided with proper interface management. • ITER buildings construction progressing. - Abstract: The ITER Tokamak machine is located in the center of Tokamak complex buildings consisting of Tokamak, Diagnostic, and Tritium buildings. Around the Tokamak complex there are over 30 auxiliary buildings housing various plant systems serving the Tokamak machine either directly or indirectly. The layout and space allocation of each auxiliary building and plant systems housed by the building are represented in the so-called Configuration Management Models (CMM). These are light 3D CAD models that define the required space envelope and the physical interfaces between the systems and the buildings and in-between the systems. The paper describes the CMM and interface management processes of the ITER auxiliary buildings and plant systems, and discusses the preparations for the plant installation phase. In addition, the current baseline configuration of the ITER plant systems in auxiliary buildings is described together with the recent developments in the configuration of different systems, as well as the current status of the construction of the buildings.

  9. The Acquisition of Auxiliary Syntax: A Longitudinal Elicitation Study. Part 1: Auxiliary BE

    Science.gov (United States)

    Theakston, Anna L.; Rowland, Caroline F.

    2009-01-01

    Purpose: The question of how and when English-speaking children acquire auxiliaries is the subject of extensive debate. Some researchers posit the existence of innately given Universal Grammar principles to guide acquisition, although some aspects of the auxiliary system must be learned from the input. Others suggest that auxiliaries can be…

  10. 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%

  11. Data report of ROSA/LSTF experiment TR-LF-07. Loss-of-feedwater transient with primary feed-and-bleed operation

    International Nuclear Information System (INIS)

    Takeda, Takeshi

    2016-07-01

    An experiment TR-LF-07 was conducted on June 23, 1992 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-IV (ROSA-IV) Program. The ROSA/LSTF experiment TR-LF-07 simulated a loss-of-feedwater transient in a pressurized water reactor (PWR) under assumptions of primary feed-and-bleed operation and total failure of auxiliary feedwater system. A safety injection (SI) signal was generated when steam generator (SG) secondary-side collapsed liquid level decreased to 3 m. Primary depressurization was initiated by fully opening a power-operated relief valve (PORV) of pressurizer (PZR) 30 min after the SI signal. High pressure injection (HPI) system was started in loop with PZR 12 s after the SI signal, while it was initiated in loop without PZR when the primary pressure decreased to 10.7 MPa. The primary and SG secondary pressures were kept almost constant because of cycle opening of the PZR PORV and SG relief valves. The PZR liquid level began to drop steeply following the PORV full opening, which caused liquid level formation at the hot leg. The PZR and hot leg liquid levels recovered due to the HPI actuation in both loops. The primary pressure became lower than the SG secondary pressure, which resulted in the actuation of accumulator (ACC) system in both loops. The PZR and hot legs became full of liquid again after the ACC actuation. The primary feed-and-bleed operation by use of the PORV, HPI and ACC systems was effective to core cooling because of no core uncovery. The experiment was terminated when the continuous core cooling was confirmed due to the successive coolant injection from the HPI system even after the ACC termination. The obtained data would be useful to study operator actions and procedures in the PWR multiple fault events which behaviors in the PZR affect. This report summarizes the test procedures, conditions and major observation in the ROSA/LSTF experiment TR-LF-07. (author)

  12. 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

  13. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

    Sabek, M.; Gaafar, M.; Poucet, A.

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author)

  14. Use of computer codes for system reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Sabek, M.; Gaafar, M. (Nuclear Regulatory and Safety Centre, Atomic Energy Authority, Cairo (Egypt)); Poucet, A. (Commission of the European Communities, Ispra (Italy). Joint Research Centre)

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author).

  15. Multireference Density Functional Theory with Generalized Auxiliary Systems for Ground and Excited States.

    Science.gov (United States)

    Chen, Zehua; Zhang, Du; Jin, Ye; Yang, Yang; Su, Neil Qiang; Yang, Weitao

    2017-09-21

    To describe static correlation, we develop a new approach to density functional theory (DFT), which uses a generalized auxiliary system that is of a different symmetry, such as particle number or spin, from that of the physical system. The total energy of the physical system consists of two parts: the energy of the auxiliary system, which is determined with a chosen density functional approximation (DFA), and the excitation energy from an approximate linear response theory that restores the symmetry to that of the physical system, thus rigorously leading to a multideterminant description of the physical system. The electron density of the physical system is different from that of the auxiliary system and is uniquely determined from the functional derivative of the total energy with respect to the external potential. Our energy functional is thus an implicit functional of the physical system density, but an explicit functional of the auxiliary system density. We show that the total energy minimum and stationary states, describing the ground and excited states of the physical system, can be obtained by a self-consistent optimization with respect to the explicit variable, the generalized Kohn-Sham noninteracting density matrix. We have developed the generalized optimized effective potential method for the self-consistent optimization. Among options of the auxiliary system and the associated linear response theory, reformulated versions of the particle-particle random phase approximation (pp-RPA) and the spin-flip time-dependent density functional theory (SF-TDDFT) are selected for illustration of principle. Numerical results show that our multireference DFT successfully describes static correlation in bond dissociation and double bond rotation.

  16. 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

  17. Auxiliary mine ventilation manual

    International Nuclear Information System (INIS)

    Workplace Safety North

    2010-01-01

    An adequate ventilation system is needed for air quality and handling in a mine and is comprised of many different pieces of equipment for removing contaminated air and supplying fresh air and thereby provide a satisfactory working environment. This manual highlights auxiliary ventilation systems made up of small fans, ducts, tubes, air movers, deflectors and additional air flow controls which distribute fresh air delivered by the primary system to all areas. A review of auxiliary ventilation is provided. Design, operation and management issues are discussed and guidelines are furnished. This manual is limited to underground hard rock operations and does not address directly other, specific auxiliary systems, either in underground coal mines or uranium mines.

  18. Auxiliary mine ventilation manual

    Energy Technology Data Exchange (ETDEWEB)

    Workplace Safety North

    2010-07-01

    An adequate ventilation system is needed for air quality and handling in a mine and is comprised of many different pieces of equipment for removing contaminated air and supplying fresh air and thereby provide a satisfactory working environment. This manual highlights auxiliary ventilation systems made up of small fans, ducts, tubes, air movers, deflectors and additional air flow controls which distribute fresh air delivered by the primary system to all areas. A review of auxiliary ventilation is provided. Design, operation and management issues are discussed and guidelines are furnished. This manual is limited to underground hard rock operations and does not address directly other, specific auxiliary systems, either in underground coal mines or uranium mines.

  19. 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

  20. Progress on radio frequency auxiliary heating system designs in ITER

    International Nuclear Information System (INIS)

    Makowski, M.; Bosia, G.; Elio, F.

    1996-09-01

    ITER will require over 100 MW of auxiliary power for heating, on- and off-axis current drive, accessing the H-mode, and plasma shut-down. The Electron Cyclotron Range of Frequencies (ECRF) and Ion Cyclotron Range of Frequencies (ICRF) are two forms of Radio Frequency (RF) auxiliary power being developed for these applications. Design concepts for both the ECRF and ICRF systems are presented, key features and critical design issues are discussed, and projected performances outlined

  1. 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)

  2. 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.)

  3. Mechanical (turbines and auxiliary equipment)

    CERN Document Server

    Sherry, A; Cruddace, AE

    2013-01-01

    Modern Power Station Practice, Volume 3: Mechanical (Turbines and Auxiliary Equipment) focuses on the development of turbines and auxiliary equipment used in power stations in Great Britain. Topics covered include thermodynamics and steam turbine theory; turbine auxiliary systems such as lubrication systems, feed water heating systems, and the condenser and cooling water plants. Miscellaneous station services, and pipework in power plants are also described. This book is comprised of five chapters and begins with an overview of thermodynamics and steam turbine theory, paying particular attenti

  4. 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

  5. Experimental Results of A1.1 Test for OECD-ATLAS Project

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung-Ho; Bae, Byoung-Uhn; Park, Yu-Sun; Kim, Jong-Rok; Choi, Nam-Hyun; Choi, Ki-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    KAERI (Korea Atomic Energy Research Institute) is operating an OECD/NEA project (hereafter, OECD-ATLAS project) by utilizing a thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). Considering the importance of the SBO scenario and the related accident mitigation measures, a prolonged SBO scenario was selected as the first test subject worthy of investigation in the OECD-ATLAS project as summarized in Table 1. After the Fukushima accident, design extension conditions (DECs) such as an SBO and a total loss of feed water (TLOFW) attracted wide international attention in that such high-risk multiple failure accidents should be revisited from the viewpoint of the reinforcement of the 'defense in depth' concept. In particular, an SBO is one of the most important DECs because a total loss of heat sink can lead to a core melt-down scenario under high pressure without any proper operator action. As for a prolonged SBO transient of the OECD-ATLAS project, two tests, named A1.1 and A1.2, were determined to be performed. In most nuclear power plants (NPPs), a turbine-driven auxiliary feedwater system was designed to remove the decay heat during the early period of an SBO transient. From a conservative point of view, however, it is necessary to investigate the thermal-hydraulic behaviors of the NPP when a turbine-driven auxiliary feedwater supply is not available during the initial period of an SBO transient and moreover a mobile pump-driven auxiliary feedwater supply can only become realized in the later period of the scenario. In particular, asymmetric heat removal characteristic through the supply of auxiliary feedwater only to one steam generator has its own peculiar importance in terms of safety analysis code validation. With an aim of considering these safety importance, in the A1.1 test, a prolonged SBO transient was simulated with two temporal phases: Phase (I) for a conservative SBO transient

  6. Cooling system for auxiliary reactor component

    International Nuclear Information System (INIS)

    Fujihira, Tomoko.

    1991-01-01

    A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

  7. Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

    2003-01-01

    The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

  8. Report of the Bulletins and Orders Task Force. Volume II. Appendices

    International Nuclear Information System (INIS)

    1980-01-01

    Appendices include: Office of Inspection and Enforcement bulletins; NRR status report on feedwater transients in BWR plants; orders on Babcock and Wilcox Company plants; letters lifting orders; letters issuing auxiliary feedwater system requirements; letter to licensees of all operating reactors, dated October 30, 1979 concerning short-term lessons learned requirements; and letters approving guidelines for preparation of small-break LOCA operating procedures

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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

  14. STARTER-GENERATOR SYSTEM FOR AUXILIARY POWER UNIT

    Directory of Open Access Journals (Sweden)

    A. V. Levin

    2017-01-01

    Full Text Available The article presents a starter-generator system for an auxiliary power unit of an aircraft. A feature of the presented system is the use of a synchronous generator with excitation from permanent magnets and a semiconductor converter. The main problem of the system is the generation of electric energy of an aircraft on the basis of a synchronous generator with excitation from permanent magnets is the absence of the possibility of regulating the voltage and frequency of electrical energy, in this connection, a semiconductor converter that ensures the conversion of generated electric energy with significant mass-dimensions characteristics.The article proposes an approach to designing a starter-generator system with a parallel connection of a synchronous generator with excitation from permanent magnets and a semiconductor converter. This approach makes it possible to significantly reduce the part of the electrical energy that needs to be converted, as a consequence, the semiconductor converter has significantly smaller mass-and-batch characteristics.In the article the modes of generation of electric energy and the starter mode of operation of the starter-generator system are considered in detail, the circuit realization of the semiconductor converter is shown. A scheme for replacing one phase of the system for generating electric energy and calculating electric parameters is presented.The possibility of creating a highly efficient starter-generator system based on a synchronous generator with excitation from permanent magnets and a semiconductor converter for an auxiliary power plant of aircrafts is shown. Structural and basic schemes for constructing a system for generating electrical energy are proposed. The approach to the choice of rational circuit solutions is substantiated, basic estimates of the electrical parameters of the system are obtained. The possibility of achieving a specific mass of a semiconductor converter for synchronous

  15. 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.

  16. 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

  17. 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.

  18. 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.)

  19. 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.

  20. 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

  1. Energic, Exergic, Exergo‐economic investigation and optimization of auxiliary cooling system (ACS equipped with compression refrigerating system (CRS

    Directory of Open Access Journals (Sweden)

    Omid Karimi Sadaghiyani

    2017-09-01

    Full Text Available Heller main cooling tower as air-cooled heat exchanger is used in the combined cycle power plants (CCPP to reduce the temperature of condenser. In extreme summer heat, the efficiency of the cooling tower is reduced and it lessens performance of Steam Turbine Generation (STG unit of Combined Cycle Power Plant (CCPP. Thus, the auxiliary cooling system (ACS is equipped with compression refrigerating system (CRS. This auxiliary system is linked with the Heller main cooling tower and improves the performance of power plant. In other words, this auxiliary system increases the generated power of STG unit of CCPP by decreasing the temperature of returning water from cooling tower Therefore, in the first step, the mentioned auxiliary cooling system (ACS as a heat exchanger and compression refrigerating system (CRS have been designed via ASPEN HTFS and EES code respectively. In order to validate their results, these two systems have been built and theirs experimentally obtained data have been compared with ASPEN and EES results. There are good agreements between results. After that, exergic and exergo-economic analysis of designed systems have been carried out. Finally, the compression refrigerating system (CRS has been optimized via Genetic Algorithm (GA. Increasing in exergy efficiency (ε from 14.23% up to 36.12% and decreasing the total cost rate (ĊSystem from 378.2 ($/h to 308.2 ($/h are as results of multi-objective optimization.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  3. 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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  5. Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system

    International Nuclear Information System (INIS)

    1986-11-01

    This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport [sr

  6. 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

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

    International Nuclear Information System (INIS)

    Ji Duan; Cao Xuewu

    2006-01-01

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

  8. 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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  10. 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.

  11. Sea water take-up facility for cooling reactor auxiliary

    International Nuclear Information System (INIS)

    Numata, Noriko; Mizutani, Akira; Hirako, Shizuka; Uchiyama, Yuichi; Oda, Atsushi.

    1997-01-01

    The present invention provides an improvement of a cooling sea water take-up facility for cooling auxiliary equipments of nuclear power plant. Namely, an existent sea water take-up facility for cooling reactor auxiliary equipments has at least two circulation water systems and three independent sea water systems for cooling reactor auxiliary equipments. In this case, a communication water channel is disposed, which connects the three independent sea water systems for cooling reactor auxiliary equipments mutually by an opening/closing operation of a flow channel partitioning device. With such a constitution, even when any combination of two systems among the three circulation water systems is in inspection at the same time, one system for cooling the reactor auxiliary equipments can be kept operated, and one system is kept in a stand-by state by the communication water channel upon periodical inspection of water take-up facility for cooling the auxiliary equipments. As a result, the sea water take-up facility for cooling auxiliary equipments of the present invention have operation efficiency higher than that of a conventional case while keeping the function and safety at the same level as in the conventional case. (I.S.)

  12. Analysis of design of auxiliary system of Booshehr Nuclear Power Plant

    International Nuclear Information System (INIS)

    Naseh Hasanzadeh, M.

    1999-01-01

    Power plant's internal auxiliary system has an important role in its safety operation. Because of the decay heat and safety aspects in the nuclear power plants, this role is more important. In this thesis, operation of the nuclear power plant with PWR reactor is studied and deferent nuclear systems described. In the next section all electrical loads in the Booshehr Nuclear Power Plant identified and feeding methods of each load is determined. by use of the single line diagram of the internal auxiliary system, the nominal rating of all electrical devices as transformers, inverters, Ups, diesel generators and etc. is determined. In the following, short circuit calculations performed and by above conclusion, rating values of circuit breakers is determined. At last the starting problems of electrical motors is studied and the results of motor's behavior at starting moment is discussed

  13. 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

  14. Coolant clean up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tajima, Fumio; Iwami, Hiroshi.

    1981-01-01

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

  15. 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.

  16. 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.

  17. 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

  18. A Source Term Calculation for the APR1400 NSSS Auxiliary System Components Using the Modified SHIELD Code

    International Nuclear Information System (INIS)

    Park, Hong Sik; Kim, Min; Park, Seong Chan; Seo, Jong Tae; Kim, Eun Kee

    2005-01-01

    The SHIELD code has been used to calculate the source terms of NSSS Auxiliary System (comprising CVCS, SIS, and SCS) components of the OPR1000. Because the code had been developed based upon the SYSTEM80 design and the APR1400 NSSS Auxiliary System design is considerably changed from that of SYSTEM80 or OPR1000, the SHIELD code cannot be used directly for APR1400 radiation design. Thus the hand-calculation is needed for the portion of design changes using the results of the SHIELD code calculation. In this study, the SHIELD code is modified to incorporate the APR1400 design changes and the source term calculation is performed for the APR1400 NSSS Auxiliary System components

  19. Computer determination of event maps with application to auxiliary supply systems

    International Nuclear Information System (INIS)

    Wredenberg, L.; Billinton, R.

    1975-01-01

    A method of evaluating the reliability of sequential operations in systems containing standby and alternate supply facilities is presented. The method is based upon the use of a digital computer for automatic development of event maps. The technique is illustrated by application to a nuclear power plant auxiliary supply system. (author)

  20. 46 CFR 58.25-10 - Main and auxiliary steering gear.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Main and auxiliary steering gear. 58.25-10 Section 58.25... AUXILIARY MACHINERY AND RELATED SYSTEMS Steering Gear § 58.25-10 Main and auxiliary steering gear. (a) Power-operated main and auxiliary steering gear must be separate systems that are independent throughout their...

  1. Chemistry and radiochemistry strategies supported by FA3-EPRTM and UK-EPRTM auxiliary systems: performances and control

    International Nuclear Information System (INIS)

    Tigeras, Arancha; Fourment, Pierre; Elgallaf; Anas; Chupin, Antoine; Fauvel, Nicola

    2012-09-01

    The design and the operation of auxiliary systems play an essential role in: - the preservation of the primary circuit integrity, - the prevention of hydrogen risk, - the control of the boron concentration and radioactivity, - the application of pH and zinc programmes. While the source term generation mainly depends on the primary circuit material and primary coolant chemistry conditioning, the source term spreading is directly linked to the auxiliary systems treatment and performances. Indeed, the auxiliary systems regulate the boron, hydrogen, lithium and zinc injection as well as the countermeasures to ensure the reactivity control and the hazardous H 2 /O 2 mixture prevention. The main principles governing the chemistry and radiochemistry in the auxiliary systems are based on the application of: - Design features for hydrogen and boron management. - Criteria for selecting the appropriate material of each system considering the functional requirements and the source term build up reduction. - Measures for minimizing the activity deposition on the surfaces of components and pipings. - Adequate and reliable systems of purification for reducing the accumulation of liquid/gas radioactivity and impurities in the circuits and for optimizing the waste production. - Chemistry program for limiting the material corrosion of auxiliary systems and preventing the source term transfer to the core. - Appropriate sampling locations and equipment to monitor the chemistry and radiochemistry parameters. This paper describes the operation of the main auxiliary systems of FLAMANVILLE3-EPR TM and UK-EPR- TM participating in the chemistry/radiochemistry management such as Chemical and Volume Control System (CVCS), Reactor Borated Water Make-up System (RBWMS), Coolant Treatment System (CTS), Gaseous Waste Processing System (GWPS), Fuel Pool Purification System (PTR [FPPS/FPCS]) also. The performances requested to these systems and the chemistry programs applied to them are discussed

  2. 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)

  3. Auxiliary accelerating system for TRIUMF cyclotron

    International Nuclear Information System (INIS)

    Zach, M.; Fong, K.; Laxdal, R.; Mackenzie, G.H.; Pacak, V.; Pearson, J.; Richardson, J.R.; Stanford, G.; Worsham, R.

    1990-06-01

    A 92 MHz auxiliary accelerating cavity has been designed and manufactured for installation in the TRIUMF cyclotron. Operating at the fourth harmonic of the RF with a peak voltage of 150 kV, it almost doubles the present energy gain per turn in the 400-500 MeV range, and reduces by ∼50% the stripping loss of the H - beam. This significant improvement will allow a substantial increase in the extracted current above the present routine level of 150μA while maintaining the same levels of residual radioactivity. The system is completed and being commissioned. A description of the design and commissioning procedures is presented, and results of beam tests given. (Author) 7 refs., 5 figs

  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. An estimation of reactor thermal power uncertainty using UFM-based feedwater flow rate in nuclear power plants

    International Nuclear Information System (INIS)

    Byung Ryul Jung; Ho Cheol Jang; Byung Jin Lee; Se Jin Baik; Woo Hyun Jang

    2005-01-01

    Most of Pressurized Water Reactors (PWRs) utilize the venturi meters (VMs) to measure the feedwater (FW) flow rate to the steam generator in the calorimetric measurement, which is used in the reactor thermal power (RTP) estimation. However, measurement drifts have been experienced due to some anomalies on the venturi meter (generally called the venturi meter fouling). The VM's fouling tends to increase the measured pressure drop across the meter, which results in indication of increased feedwater flow rate. Finally, the reactor thermal power is overestimated and the actual reactor power is to be reduced to remain within the regulatory limits. To overcome this VM's fouling problem, the Ultrasonic Flow Meter (UFM) has recently been gaining attention in the measurement of the feedwater flow rate. This paper presents the applicability of a UFM based feedwater flow rate in the estimation of reactor thermal power uncertainty. The FW and RTP uncertainties are compared in terms of sensitivities between the VM- and UFM-based feedwater flow rates. Data from typical Optimized Power Reactor 1000 (OPR1000) plants are used to estimate the uncertainty. (authors)

  6. 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

  7. 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

  8. 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)

  9. Corrosion by galvanic coupling on the steam generator auxiliary feedwater pumps at the level of the steam-tight boxes

    International Nuclear Information System (INIS)

    Dordonat, M.; Huet, M.

    1994-01-01

    Corrosion by galvanic coupling in steam generator auxiliary pump is coming from electroplated chromium cracks for the rotor steel, and from chemical KANIGEN nickel cracks for the steam-tight boxes black steel. To avoid galvanic coupling between Cr coating and the rotor steel, first an electrolytic Ni coating is done followed by an electrolytic Cr coating. To avoid galvanic coupling between black steel and graphite rings, black steel is replaced by 316L steel. (A.B.). 1 ref., 7 figs

  10. Common cause evaluations in applied risk analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Taniguchi, T.; Ligon, D.; Stamatelatos, M.

    1983-04-01

    Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights in the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system

  11. 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.)

  12. 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.

  13. Auxiliary-field quantum Monte Carlo calculations of molecular systems with a Gaussian basis

    International Nuclear Information System (INIS)

    Al-Saidi, W.A.; Zhang Shiwei; Krakauer, Henry

    2006-01-01

    We extend the recently introduced phaseless auxiliary-field quantum Monte Carlo (QMC) approach to any single-particle basis and apply it to molecular systems with Gaussian basis sets. QMC methods in general scale favorably with the system size as a low power. A QMC approach with auxiliary fields, in principle, allows an exact solution of the Schroedinger equation in the chosen basis. However, the well-known sign/phase problem causes the statistical noise to increase exponentially. The phaseless method controls this problem by constraining the paths in the auxiliary-field path integrals with an approximate phase condition that depends on a trial wave function. In the present calculations, the trial wave function is a single Slater determinant from a Hartree-Fock calculation. The calculated all-electron total energies show typical systematic errors of no more than a few millihartrees compared to exact results. At equilibrium geometries in the molecules we studied, this accuracy is roughly comparable to that of coupled cluster with single and double excitations and with noniterative triples [CCSD(T)]. For stretched bonds in H 2 O, our method exhibits a better overall accuracy and a more uniform behavior than CCSD(T)

  14. '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)

  15. Hydrothermal carbonization (HTC) of wheat straw: influence of feedwater pH prepared by acetic acid and potassium hydroxide.

    Science.gov (United States)

    Reza, M Toufiq; Rottler, Erwin; Herklotz, Laureen; Wirth, Benjamin

    2015-04-01

    In this study, influence of feedwater pH (2-12) was studied for hydrothermal carbonization (HTC) of wheat straw at 200 and 260°C. Acetic acid and KOH were used as acidic and basic medium, respectively. Hydrochars were characterized by elemental and fiber analyses, SEM, surface area, pore volume and size, and ATR-FTIR, while HTC process liquids were analyzed by HPLC and GC. Both hydrochar and HTC process liquid qualities vary with feedwater pH. At acidic pH, cellulose and elemental carbon increase in hydrochar, while hemicellulose and pseudo-lignin decrease. Hydrochars produced at pH 2 feedwater has 2.7 times larger surface area than that produced at pH 12. It also has the largest pore volume (1.1 × 10(-1) ml g(-1)) and pore size (20.2 nm). Organic acids were increasing, while sugars were decreasing in case of basic feedwater, however, phenolic compounds were present only at 260°C and their concentrations were increasing in basic feedwater. Copyright © 2015 Elsevier Ltd. All rights reserved.

  16. 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)

  17. Lessons learned from the SONGS Unit 1 water hammer event

    International Nuclear Information System (INIS)

    Chiu, C.

    1987-01-01

    On November 21, 1985, a water hammer event occurred in horizontal feedwater line B at San Onofre Nuclear Generation Site (SONGS) Unit 1. The SONGS Unit 1 is a three-loop pressurized water reactor designed by Westinghouse Electric Corp. The event was initiated by a differential current trip on the bus of auxiliary transformer C. The root cause of the event was a simultaneous failure of five check valves in the feedwater system. Two of them are located downstream of the feedwater pump, and three of them are located further downstream and on the lines to the steam generators. The failure mechanism was determined to be flow-induced vibration, which caused repeated impact between the disk stud and the disk stop. The water hammer occurred in feedwater line B during the refilling of feedwater lines A, B, and C with auxiliary feedwater. The thermal-hydraulic process to initiate the water hammer and the reason that the water hammer only happened in line B have been fully investigated and explained. A root cause analysis after the event was prompted to answer the following two questions: (1) why did these five check valves fail at that time and not in the preceding 15 yr? (2) why did only these five check valves fail? The scope of the root cause analysis involves an investigation of the valve vibration characteristics, plant operation history, and the maintenance history of the valves. The paper answers these two questions, after a brief study of the vibration characteristics of a check valve

  18. LOBI test BT-15/BT-16 pre-test calculation

    International Nuclear Information System (INIS)

    Holmes, B.J.

    1991-03-01

    LOBI is a higher pressure, electrically heated integral system facility simulating a KWU 1300 MW PWR scaled 1:712 by volume, although full scale has been maintained in the vertical direction. In order to complete the test programme before the facility is closed in 1991 a number of tests have been performed in tandem, where the test procedures were compatible. BT-15/BT-16 is one such composite test. In their original form both tests BT-15 and BT-16 simulated a loss of main feedwater transient with delayed auxiliary feedwater injection, with the pumps running in BT-15 and tripped in BT-16. The aim of each test was to investigate the loss of primary/secondary heat transfer as the steam generator secondary sides boiled down, and the subsequent recovery of heat transfer as the auxiliary feedwater was tripped on. Due to a re-scheduling of the test programme there was insufficient time to perform sensitivity studies and so only one, base case, calculation is presented. (author)

  19. Comparative LCA of methanol-fuelled SOFCs as auxiliary power systems on-board ships

    International Nuclear Information System (INIS)

    Strazza, C.; Del Borghi, A.; Costamagna, P.; Traverso, A.; Santin, M.

    2010-01-01

    Fuel cells own the potential for significant environmental improvements both in terms of air quality and climate protection. Through the use of renewable primary energies, local pollutant and greenhouse gas emissions can be significantly minimized over the full life cycle of the electricity generation process, so that marine industry accounts renewable energy as its future energy source. The aim of this paper is to evaluate the use of methanol in Solid Oxide Fuel Cells (SOFC), as auxiliary power systems for commercial vessels, through Life Cycle Assessment (LCA). The LCA methodology allows the assessment of the potential environmental impact along the whole life cycle of the process. The unit considered is a 20 kWel fuel cell system. In a first part of the study different fuel options have been compared (methanol, bio-methanol, natural gas, hydrogen from cracking, electrolysis and reforming), then the operation of the cell fed with methanol has been compared with the traditional auxiliary power system, i.e. a diesel engine. The environmental benefits of the use of fuel cells have been assessed considering different impact categories. The results of the analysis show that fuel production phase has a strong influence on the life cycle impacts and highlight that feeding with bio-methanol represents a highly attractive solution from a life cycle point of view. The comparison with the conventional auxiliary power system shows extremely lower impacts for SOFCs.

  20. 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)

  1. Effects analysis fuzzy inference system in nuclear problems using approximate reasoning

    International Nuclear Information System (INIS)

    Guimaraes, Antonio C.F.; Franklin Lapa, Celso Marcelo

    2004-01-01

    In this paper a fuzzy inference system modeling technique applied on failure mode and effects analysis (FMEA) is introduced in reactor nuclear problems. This method uses the concept of a pure fuzzy logic system to treat the traditional FMEA parameters: probabilities of occurrence, severity and detection. The auxiliary feed-water system of a typical two-loop pressurized water reactor (PWR) was used as practical example in this analysis. The kernel result is the conceptual confrontation among the traditional risk priority number (RPN) and the fuzzy risk priority number (FRPN) obtained from experts opinion. The set of results demonstrated the great potential of the inference system and advantage of the gray approach in this class of problems

  2. Introduction to deaerator in auxiliary water supply system of nuclear power plant

    International Nuclear Information System (INIS)

    Dong Jianguo; Zhou Xia; Lei Yongxia

    2015-01-01

    The paper introduces the operation theory and thermal calculation and verification requirements for the deaerator in the auxiliary water supply system of nuclear power plant. In addition, it describes the key factors in terms of function, structure, design and fabrication of equipment. (authors)

  3. 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

  4. 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

  5. Power-feedwater temperature operating domain for Sbwr applying Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar M, L. A.; Quezada G, S.; Espinosa M, E. G.; Vazquez R, A.; Varela H, J. R.; Cazares R, R. I.; Espinosa P, G., E-mail: sequega@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2014-10-15

    In this work the analyses of the feedwater temperature effects on reactor power in a simplified boiling water reactor (Sbwr) applying a methodology based on Monte Carlo simulation is presented. The Monte Carlo methodology was applied systematically to establish operating domain, due that the Sbwr are not yet in operation, the analysis of the nuclear and thermal-hydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. The results show that the reactor power is inversely proportional to the temperature of the feedwater, reactor power changes at 8% when the feed water temperature changes in 8%. (Author)

  6. Power-feedwater temperature operating domain for Sbwr applying Monte Carlo simulation

    International Nuclear Information System (INIS)

    Aguilar M, L. A.; Quezada G, S.; Espinosa M, E. G.; Vazquez R, A.; Varela H, J. R.; Cazares R, R. I.; Espinosa P, G.

    2014-10-01

    In this work the analyses of the feedwater temperature effects on reactor power in a simplified boiling water reactor (Sbwr) applying a methodology based on Monte Carlo simulation is presented. The Monte Carlo methodology was applied systematically to establish operating domain, due that the Sbwr are not yet in operation, the analysis of the nuclear and thermal-hydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. The results show that the reactor power is inversely proportional to the temperature of the feedwater, reactor power changes at 8% when the feed water temperature changes in 8%. (Author)

  7. 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

  8. Optimization of feed water control for auxiliary boiler

    International Nuclear Information System (INIS)

    Li Lingmao

    2004-01-01

    This paper described the feed water control system of the auxiliary boiler steam drum in Qinshan Phase III Nuclear Power Plant, analyzed the deficiency of the original configuration, and proposed the optimized configuration. The optimized feed water control system can ensure the stable and safe operation of the auxiliary boiler, and the normal operation of the users. (author)

  9. Extensive feedwater quality control and monitoring concept for preventing chemistry-related failures of boiler tubes in a subcritical thermal power plant

    International Nuclear Information System (INIS)

    Vidojkovic, Sonja; Onjia, Antonije; Matovic, Branko; Grahovac, Nebojsa; Maksimovic, Vesna; Nastasovic, Aleksandra

    2013-01-01

    Prevention and minimizing corrosion processes on steam generating equipment is highly important in the thermal power industry. The maintenance of feedwater quality at a level corresponding to the standards of technological designing, followed by timely respond to the fluctuation of measured parameters, has a decisive role in corrosion prevention. In this study, the comprehensive chemical control of feedwater quality in 210 MW Thermal Power Plant (TPP) was carried out in order to evaluate its potentiality to assure reliable function of the boiler and discover possible irregularity that might be responsible for frequent boiler tube failures. Sensitive on-line and off-line analytical instruments were used for measuring key and diagnostic parameters considered to be crucial for boiler safety and performances. Obtained results provided evidences for exceeded levels of oxygen, silica, sodium, chloride, sulfate, copper, and conductivity what distinctly demonstrated necessity of feedwater control improvement. Consequently, more effective feedwater quality monitoring concept was recommended. In this paper, the explanation of presumable root causes of corrosive contaminants was given including basic directions for their maintenance in proscribed limits. -- Highlights: • Feedwater quality monitoring practice in a thermal power plant has been evaluated. • The more efficient feedwater quality control have been applied. • Analysis of feedwater quality parameters has been performed. • Exceeded levels of corrosive contaminants were found. • Recommendations for their maintenance at proscribed values were given

  10. The integrated design of the ITER magnets and their auxiliary systems

    International Nuclear Information System (INIS)

    Huget, M.

    1999-01-01

    The magnet system design for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration to meet performance and operation requirements, including reliability and maintainability, in a cost effective manner. This paper identifies the requirements of long inductive burn time, large number of tokamak pulses, operational flexibility for the poloidal field (PF) system, magnet reliability and the cost constraints as the main design drivers. Key features of the magnet system which stem from these design drivers are described, together with interfaces and integration aspects of certain auxiliary systems. (author)

  11. The integrated design of the ITER magnets and their auxiliary systems

    International Nuclear Information System (INIS)

    Huguet, M.

    2001-01-01

    The magnet system design for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration to meet performance and operation requirements, including reliability and maintainability, in a cost effective manner. This paper identifies the requirements of long inductive burn time, large number of tokamak pulses, operational flexibility for the poloidal field (PF) system, magnet reliability and the cost constraints as the main design drivers. Key features of the magnet system which stem from these design drivers are described, together with interfaces and integration aspects of certain auxiliary systems. (author)

  12. 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

  13. Nuclear reactors with auxiliary boiler circuit

    International Nuclear Information System (INIS)

    George, B.V.; Cook, R.K.

    1976-01-01

    A gas-cooled nuclear reactor has a main circulatory system for the gaseous coolant incorporating one or more main energy converting units, such as gas turbines, and an auxiliary circulatory system for the gaseous coolant incorporating at least one steam generating boiler arranged to be heated by the coolant after its passage through the reactor core to provide steam for driving an auxiliary steam turbine, such an arrangement providing a simplified start-up procedure also providing emergency duties associated with long term heat removal on reactor shut down

  14. 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.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  16. 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)

  17. Interim status report on the revision of ASME PTC 12.1 -- closed feedwater heaters

    International Nuclear Information System (INIS)

    Stellern, J.L.; Hoobler, J.V.; Milton, J.W.; Welch, T.; Kona, C.; Thompson, H.N.; Tsou, J.L.

    1993-01-01

    The ASME Performance Test Code (PTC) 12.1-1978 for the performance testing of feedwater heaters is being revised extensively and updated. The committee anticipates that the final draft of the proposed Code will be ready for industry review in 1993. This Code revision will greatly enhance the usefulness and cost effectiveness of feedwater heater performance testing. This paper has been prepared to report on the progress of the committee and to disseminate information on the nature of the revision. Included in this paper are some of the notable changes intended for the Code. The most extensive change is the calculation method, which is described in step-by-step detail. An approach is also described for using ultrasonic flow techniques to test individual or split-string feedwater heaters, when flow nozzles are not available. Additionally some educational information on the use and limitations of ultrasonic measurement instrumentation is included. Discussion is also included on the required uncertainty analysis. 3 refs., 2 figs., 2 tabs

  18. Test system design for Hardware-in-Loop evaluation of PEM fuel cells and auxiliaries

    Energy Technology Data Exchange (ETDEWEB)

    Randolf, Guenter; Moore, Robert M. [Hawaii Natural Energy Institute, University of Hawaii, Honolulu, HI (United States)

    2006-07-14

    In order to evaluate the dynamic behavior of proton exchange membrane (PEM) fuel cells and their auxiliaries, the dynamic capability of the test system must exceed the dynamics of the fastest component within the fuel cell or auxiliary component under test. This criterion is even more critical when a simulated component of the fuel cell system (e.g., the fuel cell stack) is replaced by hardware and Hardware-in-Loop (HiL) methodology is employed. This paper describes the design of a very fast dynamic test system for fuel cell transient research and HiL evaluation. The integration of the real time target (which runs the simulation), the test stand PC (that controls the operation of the test stand), and the programmable logic controller (PLC), for safety and low-level control tasks, into one single integrated unit is successfully completed. (author)

  19. Reduction of residual gas in a sputtering system by auxiliary sputter of rare-earth metal

    International Nuclear Information System (INIS)

    Li Dejie

    2002-01-01

    In film deposition by sputtering, the oxidation and nitrification of the sputtered material lead to degradation of film quality, particularly with respect to metal sulfide films. We propose to use auxiliary sputtering as a method to produce a fresh film of rare-earth metal, usually dysprosium (Dy), that absorbs the active gases in a sputtering system, greatly reducing the background pressure and protecting the film from oxidation and nitrification effectively. The influence of the auxiliary sputtering power consumption, sputtering time, and medium gas pressure on the background pressure in the vacuum chamber is investigated in detail. If the auxiliary sputtering power exceeds 120 W and the sputtering time is more than 4 min, the background pressure is only one fourth of the ultimate pressure pumped by an oil diffusion pump. The absorption activity of the sputtered Dy film continues at least an hour after completion of the auxiliary sputter. Applied to film deposition of Ti and ZnS, this technique has been proven to be effective. For the Ti film, the total content of N and O is reduced from 45% to 20% when the auxiliary sputtering power of Dy is 120 W, and the sputtering time is 20 min. In the case of ZnS, the content of O is reduced from 8% to 2%

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  1. Sensitivity Analysis of Core Damage by Loss of Auxiliary Feed Water during the Extended Loss of All AC Power

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Woo Jae; Chung, Soon Il; Hwang, Su Hyun; Lee, Kyung Jin; Lee, Byung Chul [FNC Tech., Yongin (Korea, Republic of); Yun, Duk Joo; Lee, Seung Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the reactor core damage time for OPR1000 type Nuclear Power Plant (NPP) was analyzed to develop a strategy to handle ELAP and to apply to the EOP. The reactor core damage time in the ELAP condition was calculated according to the time of Auxiliary Feedwater (AFW) loss. Fukushima accident was caused by long hours of Station Black Out (SBO) caused by natural disaster beyond Design Based Accident (DBA) criteria. It led to the reactor core damage. After the accident, the regulatory authorities of each country (Japan, US, EU, IAEA, and etc.) recommended developing the necessary systems and strategies in order to cover up the Extended Loss of All AC Power (ELAP) such as one occurred in the Fukushima accident. And the need of procedure or guideline to cope with ELAP has been raised through the stress test for Wolsong Nuclear Power Plant unit 1. Current Emergency Operating Procedures (EOP) used in domestic nuclear power plant are seemed to be insufficient to cope with ELAP. Therefore, it has been required to be improved. As the result, the time of AFW loss in the ELAP condition influences greatly on core damage time.

  2. 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.

  3. 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.)

  4. Window-mounted auxiliary solar heater

    Science.gov (United States)

    Anthony, K. G.; Herndon, E. P.

    1977-01-01

    System uses hot-air collectors, no thermal storage, and fan with thermostat switches. At cost of heating efficiency, unit could be manufactured and sold at price allowing immediate entry to market as auxiliary heating system. Its simplicity allows homeowner installation, and maintenance is minimal.

  5. Loss of preferred power events in German BWRs and PWRs

    International Nuclear Information System (INIS)

    Frisch, W.

    1984-01-01

    This chapter examines the variety of courses which the loss of auxiliary power case may take in different plants, depending on the initiation of the event, the switching to the alternate auxiliary power supply and the automatic start-up sequences of auxiliaries (e.g. feedwater pumps, condensate pumps) after a short term interruption of auxiliary power. Topics considered include main station transformer failure in a boiling water reactor (BWR), main plant features, event description, evaluation of the event, deviations from expected behavior, operator actions, a comparison with start-up tests and other events, a comparison with postcalculations, and auxiliary transformer failure in a pressurized water reactor (PWR). Loss of preferred (or auxiliary) power events are standard design cases taken into account in the design of many reactor and plant systems

  6. Auxiliary buildings

    International Nuclear Information System (INIS)

    Lakner, I.; Lestyan, E.

    1979-01-01

    The nuclear power station represents a complicated and a particular industrial project. Consequently, the design of the auxiliary buildings serving the power station (offices, kitchen, refreshment room, workshops, depots, water treatment plant building, boiler houses, etc.) requires more attention than usual. This chapter gives a short survey of the auxiliary buildings already completed and discusses the problems of their design, location and structure. (author)

  7. Preliminary report on the development of rf auxiliary heating systems for TEPR-1

    International Nuclear Information System (INIS)

    Reed, B.W.; Bowen, O.N.; Hill, H.M.; Lawson, J.Q.; Newman, W.G.; Sivo, A.J.

    1977-12-01

    Conceptual designs for 50 MW (expandable to 100 MW) ICRF and LHRF heating systems suitable for auxiliary heating of the TEPR-1 plasma to ignition temperatures are presented. Engineering milestones are enumerated and the extensions of current technology required for successful completion of the project are identified

  8. Auxiliary verbs in Dinka

    DEFF Research Database (Denmark)

    Andersen, Torben

    2007-01-01

    Dinka, a Western Nilotic language, has a class of auxiliary verbs which is remarkable in the following four respects: (i) It is unusually large, comprising some 20 members; (ii) it is grammatically homogeneous in terms of both morphology and syntax; (iii) most of the auxiliary verbs correspond...... to adverbs in languages like English, while the rest are tense-aspect markers; and (iv) it is possible to combine several auxiliary verbs in a single clause. For some of the auxiliary verbs there are extant full verbs from which they have evolved. To some extent, therefore, it is possible to observe what...

  9. Auxiliary bearing design and rotor dynamics analysis of blower fan for HTR-10

    International Nuclear Information System (INIS)

    Gao Mingshan; Yang Guojun; Xu Yang; Zhao Lei; Yu Suyuan

    2005-01-01

    The electromagnetic bearing instead of ordinary mechanical bearing was chosen to support the rotor in the blower fan system with helium of 10 MW high temperature gas-cooled test reactor (HTR-10), and the auxiliary bearing was applied in the HTR-10 as the backup protector. When the electromagnetic bearing doesn't work suddenly for the power broken, the auxiliary bearing is used to support the falling rotor with high rotating speed. The rotor system will be protected by the auxiliary bearing. The design of auxiliary bearing is the ultimate safeguard for the system. This rotor is vertically mounted to hold the blower fan. The rotor's length is about 1.5 m, its weight is about 240 kg and the rotating speed is about 5400 r/min. Auxiliary bearing design and rotor dynamics analysis are very important for the design of blower fan to make success. The research status of the auxiliary bearing was summarized in the paper. A sort of auxiliary bearing scheme was proposed. MSC.Marc was selected to analyze the vibration mode and the natural frequency of the rotor. The scheme design of auxiliary bearing and analysis result of rotor dynamics offer the important theoretical base for the protector design and control system of electromagnetic bearing of the blower fan. (authors)

  10. Comparison of Severe Accident Results Among SCDAP/RELAP5, MAAP, and MELCOR Codes

    International Nuclear Information System (INIS)

    Wang, T.-C.; Wang, S.-J.; Teng, J.-T.

    2005-01-01

    This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequence of the Maanshan NPP

  11. An auxiliary frequency tracking system for general purpose lock-in amplifiers

    Science.gov (United States)

    Xie, Kai; Chen, Liuhao; Huang, Anfeng; Zhao, Kai; Zhang, Hanlu

    2018-04-01

    Lock-in amplifiers (LIAs) are designed to measure weak signals submerged by noise. This is achieved with a signal modulator to avoid low-frequency noise and a narrow-band filter to suppress out-of-band noise. In asynchronous measurement, even a slight frequency deviation between the modulator and the reference may lead to measurement error because the filter’s passband is not flat. Because many commercial LIAs are unable to track frequency deviations, in this paper we propose an auxiliary frequency tracking system. We analyze the measurement error caused by the frequency deviation and propose both a tracking method and an auto-tracking system. This approach requires only three basic parameters, which can be obtained from any general purpose LIA via its communications interface, to calculate the frequency deviation from the phase difference. The proposed auxiliary tracking system is designed as a peripheral connected to the LIA’s serial port, removing the need for an additional power supply. The test results verified the effectiveness of the proposed system; the modified commercial LIA (model SR-850) was able to track the frequency deviation and continuous drift. For step frequency deviations, a steady tracking error of less than 0.001% was achieved within three adjustments, and the worst tracking accuracy was still better than 0.1% for a continuous frequency drift. The tracking system can be used to expand the application scope of commercial LIAs, especially for remote measurements in which the modulation clock and the local reference are separated.

  12. PS auxiliary magnet

    CERN Multimedia

    CERN PhotoLab

    1974-01-01

    Units of the PS auxiliary magnet system. The picture shows how the new dipoles, used for vertical and horizontal high-energy beam manipulation, are split for installation and removal so that it is not necessary to break the accelerator vacuum. On the right, adjacent to the sector valve and the windings of the main magnet, is an octupole of the set.

  13. 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

  14. Research of Tai-chi-chuan auxiliary training system based on Kinect

    Directory of Open Access Journals (Sweden)

    Zhihong XUE

    2017-04-01

    Full Text Available In order to promote the scientific standardization of Tai-chi-chuan teaching and training, a Tai-chi-chuan auxiliary training system based on Kinect motion capture technology and extraction of the angle feature and speed feature is proposed and designed. The dynamic time planning algorithm (DTW sampling corresponding frame is applied to replace the traditional method of sampling the key frame. The auxiliary training system consists of learning module, action acquisition module and action scoring module. Learning module with stan-dard action teaching video, to meet the requirements of students learning; action acquisition module uses Microsoft developed Kinect equipment to realize data acquisition of the spatial coordinates of the human skeleton in Tai-chi-chuan action, and through filtered noise reduction and the occlusion point data processing the skeleton data is saved. The scoring module constructs the test sequence based on the 16 angle features of eight key joint nodes and the instantaneous velocity characteristics of 20 joint nodes. By comparing with the standard sequence in real time, the original scoring rules is used to achieve the scoring function, which provides students with intuitive and reliable training recommendations, so as to improve the efficiency of learning. By increasing the number of sensors, further designing and optimizing data fusion and filtering algorithm, combining with Microsoft’s Kinect2.0 version, the system can also be applied to medical rehabilitation, physical training analysis and evaluation, virtual reality and human-computer interaction, etc.

  15. Development of 8 MW Power Supply Based on Pulse Step Modulation Technique for Auxiliary Heating System on HL-2A

    International Nuclear Information System (INIS)

    Xu Weidong; Xuan Weimin; Yao Lieying; Wang Yingqiao

    2012-01-01

    The high voltage power supply (HVPS) based on pulse step modulation (PSM) has already been developed for the auxiliary heating system on HL-2A. This power supply consists of many switch power supplies, and its output voltage can be obtained by modulating their delay time and pulse widths. The PSM topology and control principle are presented in this paper. The simple algorithms for the control system are explained clearly. The switch power supply (SPS) module has been built and the test results show it can meet the requirements of the auxiliary heating system. Now, 112 SPS modules and the whole system have already been developed. Its maximum output is about 72 kV/93 A. The protection time is less than 5 μs. The different outputs of this power supply are used for the electron cyclotron resonant heating (ECRH) system with different duty ratios. The experimental results of the entire system are presented. The results indicate that the whole system can meet the requirements of the auxiliary heating system on HL-2A.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

  18. An initial examination of aging related degradation in turbine drives and governors for safety related pumps

    International Nuclear Information System (INIS)

    Cox, D.F.

    1991-01-01

    This study is being performed to examine the relationship between time dependent degradation, and current industry practices in the areas of maintenance, surveillance, and operation of steam turbine drives for safety related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized water reactor (PWR) plants, and the Reactor Core Isolation Cooling (RCIC) and High Pressure Coolant Injection (HPCI) systems for Boiling Water Reactor (BWR) facilities. This research has been conducted by examining current information in NPRDS, reviewing Licensee Event Reports, and thoroughly investigating contacts with operating plant personnel, and by personal observation. The reported information was reviewed to determine the cause of the event and the method of discovery. From this data attempts have been made at determining the predictability of events and possible preventive measures that may be implemented. Findings in a recent study on the Auxiliary Feedwater System (NUREG/CR-5404) indicate that the turbine drive is the single largest contributor to AFW system degradation. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design seem to indicate that this equipment can be a reliable component in safety systems

  19. Auxiliary basis expansions for large-scale electronic structure calculations.

    Science.gov (United States)

    Jung, Yousung; Sodt, Alex; Gill, Peter M W; Head-Gordon, Martin

    2005-05-10

    One way to reduce the computational cost of electronic structure calculations is to use auxiliary basis expansions to approximate four-center integrals in terms of two- and three-center integrals, usually by using the variationally optimum Coulomb metric to determine the expansion coefficients. However, the long-range decay behavior of the auxiliary basis expansion coefficients has not been characterized. We find that this decay can be surprisingly slow. Numerical experiments on linear alkanes and a toy model both show that the decay can be as slow as 1/r in the distance between the auxiliary function and the fitted charge distribution. The Coulomb metric fitting equations also involve divergent matrix elements for extended systems treated with periodic boundary conditions. An attenuated Coulomb metric that is short-range can eliminate these oddities without substantially degrading calculated relative energies. The sparsity of the fit coefficients is assessed on simple hydrocarbon molecules and shows quite early onset of linear growth in the number of significant coefficients with system size using the attenuated Coulomb metric. Hence it is possible to design linear scaling auxiliary basis methods without additional approximations to treat large systems.

  20. Remote Visual Testing (RVT) for the diagnostic inspection of feedwater heaters

    International Nuclear Information System (INIS)

    Nugent, M.J.; Pellegrino, B.A.

    1993-01-01

    Feedwater heaters are an important component in the overall plant heat rate, reliability, availability, performance and maintenance considerations at power stations. The ability to diagnose heater problems in-situ properly can lead to: (1) Preventative plugging of damaged, but unfailed tubes; (2) In-place repair procedures; (3) Incorporation of corrective actions into replacement designs or heater/unit operations. The benefits and limitations of Non-Destructive Testing (NDT) on feedwater heaters are briefly reviewed. All Remote Visual Testing (RVT) including borescopes, fiberscopes, videoborescopes and Closed Circuit Television (CCTV) cameras are discussed along with currently accepted formats for documentation. The benefits of a comprehensive in-place inspection involving Remote Visual Testing are discussed in relationship to its diagnostic capabilities. The results of eight post-service heater inspections are discussed along with the root cause of failure of seven unique failure mechanisms. These inspections, including FWH access, RVT tool and data analysis, are detailed. 13 figs

  1. 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)

  2. 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

  3. Detection of a regulating valve closure failure during review of recorded data after an automatic reactor shut down. Incident at the NPP Beznau-1, 27 April 1995

    International Nuclear Information System (INIS)

    Deutschmann, H.

    1996-01-01

    After recognizing a leak in the oil system of the running main feedwater pump 1 during rated power operation of the plant the operator changed feedwater supply manually to the stand-by pump 2. A short time later pump 2 was automatically tripped by the signal ''low oil pressure''. Immediate reduction of the reactor power by the operator was not successful because the scram signal ''low steam generator level and mismatch of steam/feedwater flow'' occurred and scram was actuated. In this plant a special operating feature, actuated by the scram signal, is implemented to reduce steam release to atmosphere in case of scram. The signal ''scram and average primary Temperature >287 deg. C opens the feedwater regulating valves, and later, if the average primary temperature decreases to <287 deg. C, they reclose by a redundant signal. In the experienced event, after the scram actuation, in the steam generator A a feedwater overfill occurred. The overfill protection tripped the operating feedwater pumps (main feedwater pump 3 and two auxiliary feedwater pumps). The large injection of water produced an overcooling of the primary with isolation of the volume control system outlet of the primary. The operator repaired the defective oil coolers of the feedwater pumps and restarted the plant. At that time, he had not recognized, that the plant response, which caused the steam generator overfill, was wrong. One day later, as all the recorded data were reviewed in more detail, it was found that the closure time of the feedwater regulating valve to steam generator A was much longer than designed (19 s instead 7 s). The operator requested an LCO for continued operation in spite of the fact, that the closure time was not fixed in the Technical specification. 3 figs

  4. 76 FR 28795 - Privacy Act of 1974; Department of Homeland Security United States Coast Guard-024 Auxiliary...

    Science.gov (United States)

    2011-05-18

    ... 1974; Department of Homeland Security United States Coast Guard-024 Auxiliary Database System of... Security/United States Coast Guard-024 Auxiliary Database (AUXDATA) System of Records.'' This system of...: United States Coast Guard Auxiliary Database (AUXDATA). Security classification: Unclassified. System...

  5. Auxiliary bearing design considerations for gas cooled reactors

    International Nuclear Information System (INIS)

    Penfield, S.R. Jr.; Rodwell, E.

    2001-01-01

    The need to avoid contamination of the primary system, along with other perceived advantages, has led to the selection of electromagnetic bearings (EMBs) in most ongoing commercial-scale gas cooled reactor (GCR) designs. However, one implication of magnetic bearings is the requirement to provide backup support to mitigate the effects of failures or overload conditions. The demands on these auxiliary or 'catcher' bearings have been substantially escalated by the recent development of direct Brayton cycle GCR concepts. Conversely, there has been only limited directed research in the area of auxiliary bearings, particularly for vertically oriented turbomachines. This paper explores the current state-of-the-art for auxiliary bearings and the implications for current GCR designs. (author)

  6. 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)

  7. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

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

  8. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

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

  9. 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

  10. 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.

  11. Auxiliary resonant DC tank converter

    Science.gov (United States)

    Peng, Fang Z.

    2000-01-01

    An auxiliary resonant dc tank (ARDCT) converter is provided for achieving soft-switching in a power converter. An ARDCT circuit is coupled directly across a dc bus to the inverter to generate a resonant dc bus voltage, including upper and lower resonant capacitors connected in series as a resonant leg, first and second dc tank capacitors connected in series as a tank leg, and an auxiliary resonant circuit comprising a series combination of a resonant inductor and a pair of auxiliary switching devices. The ARDCT circuit further includes first clamping means for holding the resonant dc bus voltage to the dc tank voltage of the tank leg, and second clamping means for clamping the resonant dc bus voltage to zero during a resonant period. The ARDCT circuit resonantly brings the dc bus voltage to zero in order to provide a zero-voltage switching opportunity for the inverter, then quickly rebounds the dc bus voltage back to the dc tank voltage after the inverter changes state. The auxiliary switching devices are turned on and off under zero-current conditions. The ARDCT circuit only absorbs ripples of the inverter dc bus current, thus having less current stress. In addition, since the ARDCT circuit is coupled in parallel with the dc power supply and the inverter for merely assisting soft-switching of the inverter without participating in real dc power transmission and power conversion, malfunction and failure of the tank circuit will not affect the functional operation of the inverter; thus a highly reliable converter system is expected.

  12. Fuel handling machine and auxiliary systems for a fuel handling cell

    International Nuclear Information System (INIS)

    Suikki, M.

    2013-10-01

    This working report is an update for as well as a supplement to an earlier fuel handling machine design (Kukkola and Roennqvist 2006). A focus in the earlier design proposal was primarily on the selection of a mechanical structure and operating principle for the fuel handling machine. This report introduces not only a fuel handling machine design but also auxiliary fuel handling cell equipment and its operation. An objective of the design work was to verify the operating principles of and space allocations for fuel handling cell equipment. The fuel handling machine is a remote controlled apparatus capable of handling intensely radiating fuel assemblies in the fuel handling cell of an encapsulation plant. The fuel handling cell is air tight space radiation-shielded with massive concrete walls. The fuel handling machine is based on a bridge crane capable of traveling in the handling cell along wall tracks. The bridge crane has its carriage provided with a carousel type turntable having mounted thereon both fixed and telescopic masts. The fixed mast has a gripper movable on linear guides for the transfer of fuel assemblies. The telescopic mast has a manipulator arm capable of maneuvering equipment present in the fuel handling cell, as well as conducting necessary maintenance and cleaning operations or rectifying possible fault conditions. The auxiliary fuel handling cell systems consist of several subsystems. The subsystems include a service manipulator, a tool carrier for manipulators, a material hatch, assisting winches, a vacuum cleaner, as well as a hose reel. With the exception of the vacuum cleaner, the devices included in the fuel handling cell's auxiliary system are only used when the actual encapsulation process is not ongoing. The malfunctions of mechanisms or actuators responsible for the motion actions of a fuel handling machine preclude in a worst case scenario the bringing of the fuel handling cell and related systems to a condition appropriate for

  13. Fuel handling machine and auxiliary systems for a fuel handling cell

    Energy Technology Data Exchange (ETDEWEB)

    Suikki, M. [Optimik Oy, Turku (Finland)

    2013-10-15

    This working report is an update for as well as a supplement to an earlier fuel handling machine design (Kukkola and Roennqvist 2006). A focus in the earlier design proposal was primarily on the selection of a mechanical structure and operating principle for the fuel handling machine. This report introduces not only a fuel handling machine design but also auxiliary fuel handling cell equipment and its operation. An objective of the design work was to verify the operating principles of and space allocations for fuel handling cell equipment. The fuel handling machine is a remote controlled apparatus capable of handling intensely radiating fuel assemblies in the fuel handling cell of an encapsulation plant. The fuel handling cell is air tight space radiation-shielded with massive concrete walls. The fuel handling machine is based on a bridge crane capable of traveling in the handling cell along wall tracks. The bridge crane has its carriage provided with a carousel type turntable having mounted thereon both fixed and telescopic masts. The fixed mast has a gripper movable on linear guides for the transfer of fuel assemblies. The telescopic mast has a manipulator arm capable of maneuvering equipment present in the fuel handling cell, as well as conducting necessary maintenance and cleaning operations or rectifying possible fault conditions. The auxiliary fuel handling cell systems consist of several subsystems. The subsystems include a service manipulator, a tool carrier for manipulators, a material hatch, assisting winches, a vacuum cleaner, as well as a hose reel. With the exception of the vacuum cleaner, the devices included in the fuel handling cell's auxiliary system are only used when the actual encapsulation process is not ongoing. The malfunctions of mechanisms or actuators responsible for the motion actions of a fuel handling machine preclude in a worst case scenario the bringing of the fuel handling cell and related systems to a condition appropriate for

  14. Space shuttle auxiliary propulsion system design study. Phase C report: Oxygen-hydrogen RCS/OMS integration study

    Science.gov (United States)

    Bruns, A. E.; Regnier, W. W.

    1972-01-01

    A comparison of the concepts of auxiliary propulsion systems proposed for the space shuttle vehicle is discussed. An evaluation of the potential of integration between the reaction control system and the orbit maneuvering system was conducted. Numerous methods of implementing the various levels of integration were evaluated. Preferred methods were selected and design points were developed for two fully integrated systems, one partially integrated system, and one separate system.

  15. An optimal power management system for a regenerative auxiliary power system for delivery refrigerator trucks

    International Nuclear Information System (INIS)

    Mohagheghi Fard, Soheil; Khajepour, Amir

    2016-01-01

    Highlights: • A new anti-idling system for refrigerator trucks is proposed. • This system enables regenerative braking. • An innovative two-level controller is proposed for the power management system. • A fast dynamic programming technique to find real-time SOC trajectory is proposed. • In addition to idling elimination, this system reduces fuel consumption. - Abstract: Engine idling of refrigerator trucks during loading and unloading contributes to greenhouse gas emissions due to their increased fuel consumption. This paper proposes a new anti-idling system that uses two sources of power, battery and engine-driven generator, to run the compressor of the refrigeration system. Therefore, idling can be eliminated because the engine is turned OFF and the battery supplies auxiliary power when the vehicle is stopped for loading or unloading. This system also takes advantage of regenerative braking for increased fuel savings. The power management of this system needs to satisfy two requirements: it must minimize fuel consumption in the whole cycle and must ensure that the battery has enough energy for powering the refrigeration system when the engine is OFF. To meet these objectives, a two-level controller is proposed. In the higher level of this controller, a fast dynamic programming technique that utilizes extracted statistical features of drive and duty cycles of a refrigerator truck is used to find suboptimal values of the initial and final SOC of any two consecutive loading/unloading stops. The lower level of the controller employs an adaptive equivalent fuel consumption minimization (A-ECMS) to determine the split ratio of auxiliary power between the generator and battery for each segment with initial and final SOC obtained by the high-level controller. The simulation results confirm that this new system can eliminate idling of refrigerator trucks and reduce their fuel consumption noticeably such that the cost of replacing components is recouped in a

  16. 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.).

  17. 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.).

  18. Test Specification of A1-1 Test for OECD-ATLAS Project

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Moon, Sang-Ki; Lee, Seung-Wook; Choi, Ki-Yong; Song, Chul-Hwa

    2014-01-01

    In the OECD-ATLAS project, design extension conditions (DECs) such as a station blackout (SBO) and a total loss of feed water (TLOFW) will be experimentally investigated to meet the international interests in the multiple high-risk DECs raised after the Fukushima accident. The proposed test matrix for the OECD-ATLAS project is summarized in Table 1.. In this study, detailed specification of the first test named as A1-1 in the OECD-ATLAS project was described. The target scenario of the A1-1 test is a prolonged SBO with delayed supply of turbine-driven auxiliary feedwater to only SG number 2 (SG-2). A SBO is one of the most important DECs in that without any proper operator actions, a total loss of heat sink leads to core uncover, to core damage, and ultimately a core melt-down scenario under high pressure. Due to this safety importance, a SBO is considered to be a base test item of the OECD-ATLAS project. A detailed specification of the first test named as A1-1 in the OECD-ATLAS project was described. The target scenario of the A1-1 test is a prolonged SBO with delayed supply of turbine-driven auxiliary feedwater to only SG-2 in order to consider an accident mitigation measure. The pre-test analysis using MARS code was performed with an aim of setting up the detailed test procedures for A1-1 test and also gaining the physical insights for a prolonged SBO transient. In the A1-1 test, a prolonged SBO transient will be simulated with two temporal phases: Phase (I) for conservative SBO transient without supply of turbine-driven auxiliary feedwater and Phase (II) for asymmetric cooling via single trained supply of turbine-driven auxiliary feedwater

  19. Auxiliary Deep Generative Models

    DEFF Research Database (Denmark)

    Maaløe, Lars; Sønderby, Casper Kaae; Sønderby, Søren Kaae

    2016-01-01

    Deep generative models parameterized by neural networks have recently achieved state-of-the-art performance in unsupervised and semi-supervised learning. We extend deep generative models with auxiliary variables which improves the variational approximation. The auxiliary variables leave...... the generative model unchanged but make the variational distribution more expressive. Inspired by the structure of the auxiliary variable we also propose a model with two stochastic layers and skip connections. Our findings suggest that more expressive and properly specified deep generative models converge...... faster with better results. We show state-of-the-art performance within semi-supervised learning on MNIST (0.96%), SVHN (16.61%) and NORB (9.40%) datasets....

  20. 30 CFR 57.6161 - Auxiliary facilities.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Auxiliary facilities. 57.6161 Section 57.6161...-Underground Only § 57.6161 Auxiliary facilities. (a) Auxiliary facilities used to store explosive material near work places shall be wooden, box-type containers equipped with covers or doors, or facilities...

  1. 45 CFR 707.10 - Auxiliary aids.

    Science.gov (United States)

    2010-10-01

    ... 45 Public Welfare 3 2010-10-01 2010-10-01 false Auxiliary aids. 707.10 Section 707.10 Public Welfare Regulations Relating to Public Welfare (Continued) COMMISSION ON CIVIL RIGHTS ENFORCEMENT OF... § 707.10 Auxiliary aids. (a) The Agency shall furnish appropriate auxiliary aids where necessary to...

  2. Auxiliary power unit for moving a vehicle

    Science.gov (United States)

    Akasam, Sivaprasad [Peoria, IL; Johnson, Kris W [Peoria, IL; Johnson, Matthew D [Peoria, IL; Slone, Larry M [Washington, IL; Welter, James Milton [Chillicothe, IL

    2009-02-03

    A power system is provided having at least one traction device and a primary power source configured to power the at least one traction device. In addition, the power system includes an auxiliary power source also configured to power the at least one traction device.

  3. 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)

  4. Life cycle management, design review, and condition assessment of feedwater heaters

    Energy Technology Data Exchange (ETDEWEB)

    Gammage, D.; Idvorian, N. [Babcock & Wilcox Canada Ltd., Cambridge, Ontario (Canada)

    2012-07-01

    OPEX from both the Nuclear and Fossil Power Generation Industries shows that Feedwater Heaters (FWHs) are subject to several degradation mechanisms and that this degradation commonly leads to replacement of these vessels in order to ensure reliable, efficient operation of the plants. Loss of feedwater heating will impact plant thermal performance. In response to inspection results showing on-going degradation as well as other factors, B&W Canada completed a project in conjunction with a US PWR utility to review the design, condition, and Life Cycle Management of their FWHs. This project involved a multi-disciplinary approach in order to consider all aspects of the FWHs in order to provide insight into the Life Cycle Management Plan (LCMP) so that the FWHs can be operated reliably into the future and so that adequate inspections can be conducted in order to produce a detailed condition assessment. The utility was interested in evaluating their FWH LCMP to determine if it was adequate in its requirements to enable reliable, leak-free operation of their FWH equipment. As inputs to this evaluation, it was required that B&W Canada evaluate both confirmed and plausible degradation mechanisms. They also required that the thermal hydraulic and functional design be evaluated for their particular FWHs. It was important to also incorporate industry OPEX in order to provide proper trending information for tube plugging. Out of this evaluation there were several findings and recommendations that could be used to update the utilities’ LCMP as it was apparent that the current version may not be truly reflective of the current condition of the equipment or of current industry OPEX of such FWHs. Several recommendations came from this evaluation, the most significant were: • Performing thermal/hydraulic, FIV (flow-induced vibration), and tube/shell interaction calculations to determine how the FWHs operate and how their performance can change over time as a function of tube

  5. Life cycle management, design review, and condition assessment of feedwater heaters

    International Nuclear Information System (INIS)

    Gammage, D.; Idvorian, N.

    2012-01-01

    OPEX from both the Nuclear and Fossil Power Generation Industries shows that Feedwater Heaters (FWHs) are subject to several degradation mechanisms and that this degradation commonly leads to replacement of these vessels in order to ensure reliable, efficient operation of the plants. Loss of feedwater heating will impact plant thermal performance. In response to inspection results showing on-going degradation as well as other factors, B&W Canada completed a project in conjunction with a US PWR utility to review the design, condition, and Life Cycle Management of their FWHs. This project involved a multi-disciplinary approach in order to consider all aspects of the FWHs in order to provide insight into the Life Cycle Management Plan (LCMP) so that the FWHs can be operated reliably into the future and so that adequate inspections can be conducted in order to produce a detailed condition assessment. The utility was interested in evaluating their FWH LCMP to determine if it was adequate in its requirements to enable reliable, leak-free operation of their FWH equipment. As inputs to this evaluation, it was required that B&W Canada evaluate both confirmed and plausible degradation mechanisms. They also required that the thermal hydraulic and functional design be evaluated for their particular FWHs. It was important to also incorporate industry OPEX in order to provide proper trending information for tube plugging. Out of this evaluation there were several findings and recommendations that could be used to update the utilities’ LCMP as it was apparent that the current version may not be truly reflective of the current condition of the equipment or of current industry OPEX of such FWHs. Several recommendations came from this evaluation, the most significant were: • Performing thermal/hydraulic, FIV (flow-induced vibration), and tube/shell interaction calculations to determine how the FWHs operate and how their performance can change over time as a function of tube

  6. Linearized curvatures for auxiliary fields in the de Sitter space

    Energy Technology Data Exchange (ETDEWEB)

    Vasiliev, M A

    1988-09-19

    New consistent linearized curvatures in the de Sitter space are constructed. The sequence of actions, describing bosonic and fermionic gauge auxiliary fields, is found based on these curvatures. The proposed actions are parametrized by two integer parameters, n greater than or equal to 0 and m greater than or equal to 0. The simplest case n=m=0 corresponds in the flat limit to the auxiliary fields of 'new minimal' supergravity. The hamiltonian formulation is developed for the auxiliary fields suggested; hamiltonians and first- and second-class constraints are constructed. Using these results, it is shown that the systems of fields proposed possess no dynamical degrees of freedom in de Sitter and flat spaces. In addition the hamiltonian formalism is analysed for some free dynamical systems based on linearized higher-spin curvatures introduced previously.

  7. Loss Model and Efficiency Analysis of Tram Auxiliary Converter Based on a SiC Device

    Directory of Open Access Journals (Sweden)

    Hao Liu

    2017-12-01

    Full Text Available Currently, the auxiliary converter in the auxiliary power supply system of a modern tram adopts Si IGBT as its switching device and with the 1700 V/225 A SiC MOSFET module commercially available from Cree, an auxiliary converter using all SiC devices is now possible. A SiC auxiliary converter prototype is developed during this study. The author(s derive the loss calculation formula of the SiC auxiliary converter according to the system topology and principle and each part loss in this system can be calculated based on the device datasheet. Then, the static and dynamic characteristics of the SiC MOSFET module used in the system are tested, which aids in fully understanding the performance of the SiC devices and provides data support for the establishment of the PLECS loss simulation model. Additionally, according to the actual circuit parameters, the PLECS loss simulation model is set up. This simulation model can simulate the actual operating conditions of the auxiliary converter system and calculate the loss of each switching device. Finally, the loss of the SiC auxiliary converter prototype is measured and through comparison it is found that the loss calculation theory and PLECS loss simulation model is valuable. Furthermore, the thermal images of the system can prove the conclusion about loss distribution to some extent. Moreover, these two methods have the advantages of less variables and fast calculation for high power applications. The loss models may aid in optimizing the switching frequency and improving the efficiency of the system.

  8. High voltage bus and auxiliary heater control system for an electric or hybrid vehicle

    Science.gov (United States)

    Murty, Balarama Vempaty

    2000-01-01

    A control system for an electric or hybrid electric vehicle includes a vehicle system controller and a control circuit having an electric immersion heater. The heater is electrically connected to the vehicle's high voltage bus and is thermally coupled to a coolant loop containing a heater core for the vehicle's climate control system. The system controller responds to cabin heat requests from the climate control system by generating a pulse width modulated signal that is used by the control circuit to operate the heater at a duty cycle appropriate for the amount of cabin heating requested. The control system also uses the heater to dissipate excess energy produced by an auxiliary power unit and to provide electric braking when regenerative braking is not desirable and manual braking is not necessary. The control system further utilizes the heater to provide a safe discharge of a bank of energy storage capacitors following disconnection of the battery or one of the high voltage connectors used to transmit high voltage operating power to the various vehicle systems. The control circuit includes a high voltage clamping circuit that monitors the voltage on the bus and operates the heater to clamp down the bus voltage when it exceeds a pre-selected maximum voltage. The control system can also be used to phase in operation of the heater when the bus voltage exceeds a lower threshold voltage and can be used to phase out the auxiliary power unit charging and regenerative braking when the battery becomes fully charged.

  9. Auxiliary/Master microprocessor CAMAC Crate Controller applications

    International Nuclear Information System (INIS)

    Barsotti, E.

    1975-01-01

    The need for further sophistication of an already complex serial CAMAC control system at Fermilab led to the development of an Auxilary/Master CAMAC Crate Controller. The controller contains a Motorola 6800 microprocessor, 2K bytes of RAM, and 8K bytes of PROM memory. Bussed dataway lines are time shared with CAMAC signals to provide memory expansion and direct addressing of peripheral devices without the need of external cabling. The Auxiliary/Master Crate Controller (A/MCC) can function as either a Master, i.e., stand alone, crate controller or as an Auxiliary controller to Fermilab's Serial Crate Controller (SCC). Two modules, one single- and one double-width, make up an A/MCC. The microprocessor has one nonmaskable and one maskable vectored interrupt. Time sharing the dataway between SCC programmed and block transfer generated dataway cycles and A/MCC operations still allows a 99 percent microprocessor CPU busy time. Since the conception of the A/MCC, there has been an increasing number of control system-related projects proposed which would not have been possible or would have been very difficult to implement without such a device. The first such application now in use at Fermilab is a stand-alone control system for a mass spectrometer experiment in the Main Ring Internal Target Area. This application in addition to other proposed A/MCC applications, both stand-alone and auxiliary, is discussed

  10. Intrinsically safe electrical installations, auxiliary circuits and electric communication equipment

    Energy Technology Data Exchange (ETDEWEB)

    Herms, C D

    1981-11-19

    Technical progress has not stopped short of electrical systems in mining, so that three new chapters are new included in the VDE regulations leaflet No. 0118 on 'Installation of electrical systems in underground coal mining'. The regulations on intrinsically safe electric systems, auxiliary circuits and communication systems are briefly described, and grounds for the regulations are presented. The regulations already take account of European regulations on intrinsic safety which will soon be published in a European Regulation on Mine Explosions. In the chapters on auxiliary circuits and communication systems, protection against direct contact, fires, and explosions is discussed as well as the further goal of reliable signal transmission.

  11. 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.

  12. Auxiliary Sensor-Based Borehole Transient Electromagnetic System for the Nondestructive Inspection of Multipipe Strings

    Directory of Open Access Journals (Sweden)

    Bo Dang

    2017-08-01

    Full Text Available Transient electromagnetic (TEM techniques are widely used in the field of geophysical prospecting. In borehole detection, the nondestructive inspection (NDI of a metal pipe can be performed efficiently using the properties of eddy currents. However, with increasing concern for safety in oil and gas production, more than one string of pipe is used to protect wellbores, which complicates data interpretation. In this paper, an auxiliary sensor-based borehole TEM system for the NDI of multipipe strings is presented. On the basis of the characteristics of the borehole TEM model, we investigate the principle behind the NDI of multipipe strings using multiple time slices of induced electromotive force (EMF in a single sensor. The results show that the detection performance of NDI is strongly influenced by eddy-current diffusion in the longitudinal direction. To solve this problem, we used time slices of the induced EMF in both the main and auxiliary sensors. The performance of the proposed system was verified by applying it to an oil well with a production casing and liner. Moreover, field experiments were conducted, and the results demonstrate the effectiveness of the proposed method.

  13. Component Data Base for Space Station Resistojet Auxiliary Propulsion

    Science.gov (United States)

    Bader, Clayton H.

    1988-01-01

    The resistojet was baselined for Space Station auxiliary propulsion because of its operational versatility, efficiency, and durability. This report was conceived as a guide to designers and planners of the Space Station auxiliary propulsion system. It is directed to the low thrust resistojet concept, though it should have application to other station concepts or systems such as the Environmental Control and Life Support System (ECLSS), Manufacturing and Technology Laboratory (MTL), and the Waste Fluid Management System (WFMS). The information will likely be quite useful in the same capacity for other non-Space Station systems including satellite, freeflyers, explorers, and maneuvering vehicles. The report is a catalog of the most useful information for the most significant feed system components and is organized for the greatest convenience of the user.

  14. A Polyethylene Moderator Design for Auxiliary Ex-core Neutron Detector

    International Nuclear Information System (INIS)

    Lee, Hwan Soo; Shin, Ho Cheol; Bae, Seong Man

    2012-01-01

    The moderator of detector assembly in ENFMS (Excore Neutron Flux Monitoring System) plays a key role for slowing down from fast neutron to thermal neutron at outside of reactor vessel. Since neutron monitoring detector such as BF3, fission chamber detectors mostly responds to thermal neutron, moderator should be included to neutron detector assembly to detect more efficiently. Generally, resin has been used for moderator of detector in ENFMS of OPR1000 and APR1400, because resin has stable thermal resistance, availability and high neutron moderation characteristics due to the light atomic materials. In case of an auxiliary ex-core neutron detector, the polyethylene is suggested that polyethylene has a better moderator rather than resin, then, the amounts of moderator are reduced. This is important thing for auxiliary ex-core detector equipment at reactor, because the auxiliary equipment should affect minimally to another system. In this study, polyethylene moderator is designed for auxiliary ex-core neutron detector. To find out the optimal thickness of polyethylene moderator, preliminary simulation and experiments are performed. And sensitivity simulation for detector moderator at actual reactor is performed by DORT code

  15. Simulation of the KAERI PASCAL Test with MARS-KS and TRACE Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Won; Cheong, Aeju; Shin, Andong; Cho, Min Ki [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    In order to validate the operational performance of the PAFS, KAERI has performed the experimental investigation using the PASCAL (PAFS Condensing heat removal Assessment Loop) facility. In this study, we simulated the KAERI PASCAL SS-540-P1 test with MARS-KS V1.4 and TRACE V5.0 p4 codes to assess the code predictability for the condensation heat transfer inside the passive auxiliary feedwater system. We simulated the KAERI PASCAL SS-540-P1 test with MARS-KS V1.4 and TRACE V5.0 p4 codes to assess the code predictability for the condensation heat transfer inside the passive auxiliary feedwater system. The calculated results of heat flux, inner wall surface temperature of the condensing tube, fluid temperature, and steam mass flow rate are compared with the experimental data. The result shows that the MARS-KS generally under-predict the heat fluxes. The TRACE over-predicts the heat flux at tube inlet region and under-predicts it at tube outlet region. The TRACE prediction shows larger amount of steam condensation by about 3% than the MARS-KS prediction.

  16. Condensation heat transfer of a feed-water heater and improvement of its performance

    International Nuclear Information System (INIS)

    Takamori, Kazuhide; Murase, Michio; Baba, Yoshikazu; Aihara, Tsuyoshi

    1995-01-01

    In this study, a condensation heat transfer model, coupled with a three-dimensional two-phase flow analysis, was developed. In the heat transfer model, the liquid film flow rate on the heat transfer tubes was calculated by a mass balance equation and the liquid film thickness was calculated from the liquid film flow rate using Nusselt's laminar flow model and Fujii's equation for the steam velocity effect. The model was verified by condensation heat transfer experiments. In the experiments, 112 horizontal, staggered tubes with an outer diameter of 16mm and length of 0.55m were used. The calculated over-all heat transfer coefficients agreed with the data within ±5% under the inlet quality conditions of 13-100%. Based on a three-dimensional two-phase flow analysis, an improved feed-water heater with support plates, which have flow holes between the upper and lower tube bundles, was designed. The total heat exchange capacity of the improved feed-water heater increased about 6%. (author)

  17. Inteligent control system for a CANDU 600 type reactor process

    International Nuclear Information System (INIS)

    Venescu, B.; Zevedei, D.; Jurian, M.; Venescu, R.

    2013-01-01

    The present paper is set on presenting a highly intelligent configuration, capable of controlling, without the need of the human factor, a complete nuclear power plant type of system, giving it the status of an autonomous system. The urge for such a controlling system is justified by the amount of drawbacks that appear in real life as disadvantages, loses and sometimes even inefficiency in the current controlling and comanding systems of the nuclear reactors. The application stands in the comand sent from the auxiliary feedwater flow control valves to the steam generators. As an environment fit for development I chose Matlab Simulink to simulate the behaviour of the process and the adjusted system. Comparing the results obtained after the fuzzy regulation with those obtained after the classical regulation, we can demonstrate the necessity of implementing artificial intelligence techniques in nuclear power plants and we can agree to the advantages of being able to control everything automatically. (authors)

  18. Qualitative and Quantitative Analysis of Organic Impurities in Feedwater of a Heat-Recovery Steam Generator

    Science.gov (United States)

    Chichirov, A. A.; Chichirova, N. D.; Filimonova, A. A.; Gafiatullina, A. A.

    2018-03-01

    In recent years, combined-cycle units with heat-recovery steam generators have been constructed and commissioned extensively in the European part of Russia. By the example of the Kazan Cogeneration Power Station no. 3 (TETs-3), an affiliate of JSC TGK-16, the specific problems for most power stations with combined-cycle power units that stem from an elevated content of organic impurities in the feedwater of the heat-recovery steam generator (HRSG) are examined. The HRSG is fed with highly demineralized water in which the content of organic carbon is also standardized. It is assumed that the demineralized water coming from the chemical water treatment department of TETs-3 will be used. Natural water from the Volga River is treated to produce demineralized water. The results of a preliminary analysis of the feedwater demonstrate that certain quality indices, principally, the total organic carbon, are above the standard values. Hence, a comprehensive investigation of the feedwater for organic impurities was performed, which included determination of their structure using IR and UV spectroscopy techniques, potentiometric measurements, and element analysis; determination of physical and chemical properties of organic impurities; and prediction of their behavior in the HRSG. The estimation of the total organic carbon revealed that it exceeded the standard values in all sources of water comprising the feedwater for the HRSG. The extracted impurities were humic substances, namely, a mixture of humic and fulvic acids in a 20 : 80 ratio, respectively. In addition, an analysis was performed of water samples taken at all intermediate stages of water treatment to study the behavior of organic substances in different water treatment processes. An analysis of removal of the humus substances in sections of the water treatment plant yielded the concentration of organic substances on the HRSG condensate. This was from 100 to 150 μg/dm3. Organic impurities in boiler water can induce

  19. 30 CFR 75.331 - Auxiliary fans and tubing.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Auxiliary fans and tubing. 75.331 Section 75... HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Ventilation § 75.331 Auxiliary fans and tubing. (a) When auxiliary fans and tubing are used for face ventilation, each auxiliary fan shall be— (1...

  20. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

    Energy Technology Data Exchange (ETDEWEB)

    Shah, V.N.; Ware, A.G.; Porter, A.M.

    1997-03-01

    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

  1. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

    International Nuclear Information System (INIS)

    Shah, V.N.; Ware, A.G.; Porter, A.M.

    1997-03-01

    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage

  2. Unavailability modeling and analysis of redundant safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Vaurio, J.K.; Sciaudone, D.

    1979-10-01

    Analytical expressions have been developed to estimate the average unavailability of an m-out-of-n (m/n, 1 less than or equal to m less than or equal to n less than or equal to 4) standby safety system of a nuclear power plant. The expressions take into account contributions made by testing, repair, equipment failure, human error, and different testing schemes. A computer code, ICARUS, has been written to incorporate these analytical equations. The code is capable of calculating the average unavailability, optimum test interval, and relative contributions of testing, repair, and random failures for any of three testing schemes. After verification of the methodology and coding in ICARUS, a typical auxiliary feedwater system of a nuclear power plant was analyzed. The results show that the failure modes associated with testing and true demands contribute considerably to the unavailability and that diesel generators are the most critical components contributing to the overall unavailability of the system.

  3. Unavailability modeling and analysis of redundant safety systems

    International Nuclear Information System (INIS)

    Vaurio, J.K.; Sciaudone, D.

    1979-10-01

    Analytical expressions have been developed to estimate the average unavailability of an m-out-of-n (m/n, 1 less than or equal to m less than or equal to n less than or equal to 4) standby safety system of a nuclear power plant. The expressions take into account contributions made by testing, repair, equipment failure, human error, and different testing schemes. A computer code, ICARUS, has been written to incorporate these analytical equations. The code is capable of calculating the average unavailability, optimum test interval, and relative contributions of testing, repair, and random failures for any of three testing schemes. After verification of the methodology and coding in ICARUS, a typical auxiliary feedwater system of a nuclear power plant was analyzed. The results show that the failure modes associated with testing and true demands contribute considerably to the unavailability and that diesel generators are the most critical components contributing to the overall unavailability of the system

  4. A study on the shell wall thinning causes identified through experiment, numerical analysis and ultrasonic test of high-pressure feedwater heater

    International Nuclear Information System (INIS)

    Hwang, Kyeong Mo; Woo, Lee; Jin, Tae Eun; Kim, Kyung Hoon

    2008-01-01

    Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which accelerates as the operation progresses. Several nuclear power plants in Korea have undergone this damage around the impingement baffle - installed downstream of the high-pressure turbine extraction steam line - inside numbers 5A and 5B feedwater heaters. At that point, the extracted steam from the high-pressure turbine consists in the form of two-phase fluid at high temperature, high pressure and high velocity. Since it flows in reverse direction after impinging the impingement baffle, the shell wall of number 5 high-pressure feedwater heater may be affected by flow-accelerated corrosion. This paper describes the comparisons between the numerical analysis results using the FLUENT code and the downscaled experimental data in an effort to determine root causes of the shell wall thinning of the high-pressure feedwater heaters. The numerical analysis and experimental data were also confirmed by the actual wall thickness measured by ultrasonic tests. From the comparison of the results for the local velocity profiles and the wall thinning measurements, the local velocity component only in the y-direction flowing vertically to the shell wall, and not in the x- and z-directions, was analogous to the wall thinning data

  5. 46 CFR 63.25-1 - Small automatic auxiliary boilers.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Small automatic auxiliary boilers. 63.25-1 Section 63.25... AUXILIARY BOILERS Requirements for Specific Types of Automatic Auxiliary Boilers § 63.25-1 Small automatic auxiliary boilers. Small automatic auxiliary boilers defined as having heat-input ratings of 400,000 Btu/hr...

  6. Developing the optimum boiler water and feedwater treatment for fossil plants

    Energy Technology Data Exchange (ETDEWEB)

    Dooley, B [Electric Power Research Inst., Palo Alto, California (United States)

    1996-12-01

    Over the last two years a new set of cycle chemistry guidelines has been developed for each of the treatments used in fossil plants. These revisions have been based on research conducted over the last ten years, much at the international collaborative level. By careful selection and optimization of the boiler water and feedwater treatments, it will be possible to accrue large financial, maintenance, availability and performance improvements. (au) 14 refs.

  7. 7 CFR 15b.37 - Auxiliary aids.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 1 2010-01-01 2010-01-01 false Auxiliary aids. 15b.37 Section 15b.37 Agriculture... ACTIVITIES RECEIVING FEDERAL FINANCIAL ASSISTANCE Other Aid, Benefits, or Services § 15b.37 Auxiliary aids... appropriate auxiliary aids to persons with impaired sensory, manual, or speaking skills, where necessary to...

  8. 46 CFR 182.620 - Auxiliary means of steering.

    Science.gov (United States)

    2010-10-01

    ... TONS) MACHINERY INSTALLATION Steering Systems § 182.620 Auxiliary means of steering. (a) Except as... personnel hazards during normal or heavy weather operation. (b) A suitable hand tiller may be acceptable as...

  9. Analysis of ultrasound propagation in high-temperature nuclear reactor feedwater to investigate a clamp-on ultrasonic pulse doppler flowmeter

    International Nuclear Information System (INIS)

    Tezuka, Kenichi; Mori, Michitsugu; Wada, Sanehiro; Aritomi, Masanori; Kikura, Hiroshige; Sakai, Yukihiro

    2008-01-01

    The flow rate of nuclear reactor feedwater is an important factor in the operation of a nuclear power reactor. Venturi nozzles are widely used to measure the flow rate. Other types of flowmeters have been proposed to improve measurement accuracy and permit the flow rate and reactor power to be increased. The ultrasonic pulse Doppler system is expected to be a candidate method because it can measure the flow profile across the pipe cross section, which changes with time. For accurate estimation of the flow velocity, the incidence angle of ultrasound entering the fluid should be estimated using Snell's law. However, evaluation of the ultrasound propagation is not straightforward, especially for a high-temperature pipe with a clamp-on ultrasonic Doppler flowmeter. The ultrasound beam path may differ from what is expected from Snell's law due to the temperature gradient in the wedge and variation in the acoustic impedance between interfaces. Recently, simulation code for ultrasound propagation has come into use in the nuclear field for nondestructive testing. This article analyzes and discusses ultrasound propagation, using 3D-FEM simulation code plus the Kirchhoff method, as it relates to flow profile measurement in nuclear reactor feedwater with the ultrasonic pulse Doppler system. (author)

  10. Research needs for risk-informed, performance-based regulation

    International Nuclear Information System (INIS)

    Bailey, J.A.

    1997-01-01

    Palo Verde Nuclear Generating Station has used PRA-derived risk insights for about 10 years now. The plant originally started applying PRA modeling to an auxiliary feedwater system during the initial licensing phases of the plant, and as a result of that, they were able to work with the NRC and apply some graded quality requirements to that particular system. There was a third redundant auxiliary feedwater pump, and they now can treat that system as partially safety related and partially non-safety related. So it was an advance for Palo Verde at that time to be able to make decisions with a PRA and they began learning how to use those techniques. After completing the IPE it became natural for the plant to make a transition into other areas at the plant to look for areas where the insights gained from PRA could be applied into their decision-making processes. Those that the plant embarked upon initially were areas where they could gain operational risk assessment insights. The author goes on to discuss experiences gained in using these techniques to better assess the safety of operations within the plant. In addition he offers comments on areas which need further development and research to make them more applicable to a plant by plant basis

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  12. ASCERTAINMENT OF ELECTRIC-SUPPLY SCHEMES RELIABILITY FOR THE ATOMIC POWER PLANT AUXILIARIES

    Directory of Open Access Journals (Sweden)

    A. L. Starzhinskij

    2015-01-01

    Full Text Available The paper completes ascertainment of electrical-supply scheme reliability for the auxiliaries of a nuclear power plant. Thereat the author considers the system behavior during the block normal operation, carrying out current maintenance, and capital repairs in combination with initiating events. The initiating events for reactors include complete blackout, i.e. the loss of outside power supply (normal and reserve; emergency switching one of the working turbogenerators; momentary dumping the normal rating to the level of auxiliaries with seating the cutout valve of one turbo-generator. The combination of any initiating event with the repairing mode in case of one of the system elements failure should not lead to blackout occurrence of more than one system of the reliable power supply. This requirement rests content with the help of the reliable power supply system self-dependence (electrical and functional and the emergency power-supply operational autonomy (diesel generator and accumulator batteries.The reliability indicators of the power supply system for the nuclear power plant auxiliaries are the conditional probabilities of conjoined blackout of one, two, and three sections of the reliable power supply conditional upon an initiating event emerging and the blackout of one, two, and three reliable power-supply sections under the normal operational mode. Furthermore, they also are the blackout periodicity of one and conjointly two, three, and four sections of normal operation under the block normal operational mode. It is established that the blackout of one bus section of normal operation and one section of reliable power-supply system of the auxiliaries that does not lead to complete blackout of the plant auxiliaries may occur once in three years. The probability of simultaneous power failure of two or three normal-operation sections and of two reliable power-supply sections during the power plant service life is unlikely.

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

    International Nuclear Information System (INIS)

    Bennert, J.; Becher, L.

    1977-01-01

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

  14. Development of a multi-path ultrasonic flow meter for the application to feedwater flow measurement in nuclear power plants

    International Nuclear Information System (INIS)

    Jong, J. C.; Ha, J. H.; Kim, Y. H.; Jang, W. H.; Park, K. S.; Park, M. S.; Park, M. H.

    2002-01-01

    In this work, we propose a method to measure the feedwater flow using multi-path ultrasonic flow meter (UFM). Since the UFM measures a path velocity at which the ultrasonic wave is propagated, the flow profile may be important to convey the path velocity to the velocity averaged over the entire cross section of the flowing medium. The conventional UFM has used the smooth-wall circular pipe model presented by Nikurades. However, this model covers a lower range which is less than 3.2 million while the Reynolds number of the feedwater flow in operating nuclear power plants (NPPs) is about 20 million. Therefore, we feedwater flow in operating nuclear power plants (NPPs) is about 20 million. Therefore, we proposed the non-linear correlation model that combines the ratio between the DP output and proposed the non-linear correlation model that combines the ratio between the DP output and UFM output. Experiments were performed using both computer simulation and newly constructed NPPs' test data. The uncertainty analysis result shows that the proposed method has reasonably lower uncertainty than conventional UFM

  15. Investigation into sensitivity of Darlington boiler 2 feedwater flow calibration factor to boiler level control valve configuration

    Energy Technology Data Exchange (ETDEWEB)

    Coppens, D. [Darlington Nuclear Generating Station, Ontario Power Generation, Bowmanville, Ontario (Canada); Gurevich, Y. [Daystar Technologies Inc., Toronto, Ontario (Canada); Ton, V. [Inspection and Maintenance Services Div., Ontario Power Generation, Ajax, Ontario (Canada); Zobin, D. [AMEC NSS Ltd., Toronto, Ontario (Canada)

    2009-07-01

    The Ultrasonic Cross-Correlation Flow Meter (USCCFM) has been used for regular feedwater flow calibration at Darlington NGS since the early nineties. Typical measurement repeatability over the duration of a calibration run (normally several weeks long) is within {+-}0.2%. However, it was recently noticed that BO2 calibration factor experienced sudden changes of close to 1%. The paper will describe several different approaches used for identifying the reason for the observed effect. The investigation has revealed that changes in USCCFM readings are due to the complicated geometry of BO2 feedwater piping and that its accuracy can be as high as a fraction of percent if several readings are averaged around the pipe. (author)

  16. Accuracy Solution of Boundary Problems in Electrostatics for Systems "Conductors-Dielectrics" by Means of Auxiliary Charges

    CERN Document Server

    Topuriya, T P

    2004-01-01

    The analysis has been carried out on checking the influence of auxiliary charges on solution accuracy of boundary problems of electrostatics for systems "conductors-dielectrics". This accuracy depends on the number of charges and configuration of their allocation. The extended round dielectric in the electric field of a parallel-plate capacitor was taken as a physical model.

  17. Use of the GO methodology to assess plant operating data

    International Nuclear Information System (INIS)

    Wenger, L.O.; Danner, M.A.

    1987-01-01

    In September 1985, the Institute of Nuclear Power Operations (INPO) published The Operational Performance of Auxiliary Feedwater Systems in US PWRs, 1980-1984, in which many of the problems experienced by the nuclear industry with auxiliary feedwater systems (AFWSs) were summarized. The principal conclusions of the INPO report were that pressurized water reactor AFWS equipment failure rates had nearly doubled industry-wide since 1980. Calvert Cliffs Units 1 and 2 were highlighted by INPO as being among some of the poorer performers identified in the report. The study objective involved estimating the historical AFWS unavailability so that it could be compared with a predetermined acceptable availability goal. From this comparison, the relative importance of individual contributions to overall system unavailability resulting from various equipment failure modes could be assessed. The GO methodology, a computerized reliability analyses technique, was selected as the means for performing additional analyses. A GO system model was developed to represent the current Calvert Cliffs AFWS configuration. In summary, it was shown that integration of plant data with a comprehensive availability/reliability model can provide both the necessary insights to focus attention on the most important contributions to system unavailability and a pragmatic way for developing a cost effective implementation strategy. The GO methodology was useful and provided the requisite results with a minimal outlay of valuable resources

  18. ERP-IV-A program for transient thermal-hydraulic analysis of PWR plant

    International Nuclear Information System (INIS)

    Dai Anguo; Tang Jiahuan; Qian Huifu; Gao Zhikang

    1987-12-01

    The author deal with the descriptions of physical model of transient process in PWR plant and the function of ERP-IV (ERR-IV Transient Thermo-Hydraulic Analysis Code). The code has been developed for safety analysis and design transient. The code is characterized by the multi-loop long-term, short term, wide-range plant simulation with the capability to analyze natural circulation condition. The description of ERP-IV includes following parts: reactor, primary coolant loops, pressurizer, steam generators, main steam system, turbine, feedwater system, steam dump, relive valves, and safety valves in secondary side, etc.. The code can use for accident analysis, such as loss of all A.C. power to power plant auxiliaries (a station blackout), loss of normal feedwater, loss of load, loss of condenser vacuum and other events causing a turbine trip, complete loss of forced reactor coolant flow, uncontrolled rod cluster control assembly bank withdrawal. It can also be used for accident analysis of the emergency and limiting conditions, such as feedwater line break and main steam line rupture. It can also be utilized as a tool for system design studies, component design, setpoint studies and design transition studies, etc

  19. Quick look report for semiscale MOD-2C Test S-FS-11

    International Nuclear Information System (INIS)

    Plessinger, M.P.

    1985-11-01

    Results of a preliminary analysis of the fifth test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-11 simulated a pressurized water reactor transient initiated by a 50% break in a steam generator bottom feedwater line downstream of the check valve. With the exception of primary pressure, the initial conditions represented the initial conditions used for the C-E System 80 Final Safety Analysis Report (FSAR) Appendix 15B calculations. The transient included an initial 600 s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant followed by break isolation and affected loop steam generator refill with auxiliary feedwater. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overpressurization and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overpressurization and primary-to-secondary heat transfer. 64 figs

  20. 33 CFR 5.47 - Auxiliary ensign.

    Science.gov (United States)

    2010-07-01

    ... matching blue Coast Guard Auxiliary emblem is centered. The white slash shall be at a 70 degree angle, rising away from the hoist. (c) The Auxiliary emblem consists of a disk with the shield of the Coat of...

  1. Device for the analysis of feedwater and condensation samples from power plants

    International Nuclear Information System (INIS)

    Mostofin, A.A.; Sorokina, N.S.

    1978-01-01

    An improved version of a device for automatic measurement of the salt and NH 3 contents of feedwater and condensate samples from nuclear power plants is described. Only one sample is required for determining both values. The invention proposes on the one hand to change the dimensions of a throttle opening and on the other to install a second measuring instrument (conductivity measuring instrument). (UWI) [de

  2. Auxiliary equipment cooling circuit in nuclear reactors

    International Nuclear Information System (INIS)

    Yanagisawa, Ko.

    1986-01-01

    Purpose: To prevent the propagation of bacterias that transform NO 2 into NO 3 in auxiliary equipment coolants using corrosion inhibitors of nitrite type in BWR type reactors. Method: In auxiliary equipments coolant systems, water quality is controlled by using purified water as supplement water and nitrite such as Na 2 NO 2 as the corrosion inhibitors. However, in the circumstance where dissolved oxygen is present, bacteria propagate to oxidize NO 2 into NO 3 . Thus, NO 2 at 200 ppm is reduced to 20 ppm. In view of the above, a surge tank supplied from water supplement line is connected in series and a deaeration device is disposed thereto. Since the presence of dissolved oxygen causes the bacteria to propagate it is desired that the dissolved oxygen density in the supplement water is less than 5 ppm. Deaeration and pressure reduction in the surge tank can remove the dissolved oxygen, prevent NO 3 increase and also prevent stress corrosion cracks in the system pipeways. (Horiuchi, T.)

  3. Systems for Nuclear Auxiliary Power annual report, government fiscal year 1976/TQ

    International Nuclear Information System (INIS)

    1976-01-01

    The overall objective of the Systems for Nuclear Auxiliary Power (SNAP) Program is to continue system and component engineering activities relating to the zirconium hydride (ZrH) reactor. The specific objectives for FY 1976/TQ were to: (1) study standardized ZrH reactor space power systems and components, (2) perform preconceptual analysis and design of ZrH reactor--organic Rankine power systems for subsea applications, (3) conduct fuel and hydrogen barrier investigations, (4) perform system studies in support of the Department of Defense and their contractors as directed by ERDA, (5) test components, and (6) provide for material disposal and facility surveillance. In the study, representative systems which utilize Brayton, Rankine, and Stirling cycle power conversion units as well as thermoelectric modules, are analyzed at power levels of 10, 25, 50, and 75 kWe. Waste heat rejection is accomplished by concentric, cylindrical space radiators which can be nested during launch for space shuttle integration. Subsequent studies, which supported this effort, were completed and provided useful information on system reliability and survivability

  4. Water-hammer in the feed-water pipes for PWR steam generators

    International Nuclear Information System (INIS)

    Gonnet, Bernard; Leroy, Claude; Oullion, Jean; Yazidjian, J.-C.

    1979-01-01

    PWR boiler water feed pipes have been known for several years to be affected by violent water-hammer during start-ups and operation of the plant. In view of the varying results of corrective design modifications in America and Europe, FRAMATOME undertook an experimental research programme which resulted in the adoption of cruciform tubes on the feed-water distributor as the most reliable solution. Subsequent tests at Fessenheim I confirmed the effectiveness of this device [fr

  5. Manufacture and installation of reactor auxiliary facilities for advanced thermal prototype reactor 'Fugen'

    International Nuclear Information System (INIS)

    Kawahara, Toshio; Matsushita, Tadashi

    1977-01-01

    The facilities of reactor auxiliary systems for the advanced thermal prtotype reactor ''Fugen'' were manufactured in factories since 1972, and the installation at the site began in November, 1974. It was almost completed in March, 1977, except a part of the tests and inspections, therefore the outline of the works is reported. The ATR ''Fugen'' is a heavy water-moderated, boiling light water reactor, and its reactor auxiliary systems comprise mainly the facilities for handling heavy water, such as heavy water cooling system, heavy water cleaning system, poison supplying system, helium circulating system, helium cleaning system, and carbon dioxide system. The poison supplying system supplies liquid poison to the heavy water cooling system to absorb excess reactivity in the initial reactor core. The helium circulating system covers heavy water surface with helium to prevent the deterioration of heavy water and maintains heavy water level by pressure difference. The carbon dioxide system flows highly pure CO 2 gas in the space of pressure tubes and carandria tubes, and provides thermal shielding. The design, manufacture and installation of the facilities of reactor auxiliary systems, and the helium leak test, synthetic pressure test and total cleaning are explained. (Kako, I.)

  6. 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.

  7. The Acquisition of Auxiliary Syntax: A Longitudinal Elicitation Study. Part 2: The Modals and Auxiliary DO

    Science.gov (United States)

    Rowland, Caroline F.; Theakston, Anna L.

    2009-01-01

    Purpose: The study of auxiliary acquisition is central to work on language development and has attracted theoretical work from both nativist and constructivist approaches. This study is part of a 2-part companion set that represents a unique attempt to trace the development of auxiliary syntax by using a longitudinal elicitation methodology. The…

  8. Operation and Control of a Direct-Driven PMSG-Based Wind Turbine System with an Auxiliary Parallel Grid-Side Converter

    Directory of Open Access Journals (Sweden)

    Jiawei Chu

    2013-07-01

    Full Text Available In this paper, based on the similarity, in structure and principle, between a grid-connected converter for a direct-driven permanent magnet synchronous generator (D-PMSG and an active power filter (APF, a new D-PMSG-based wind turbine (WT system configuration that includes not only an auxiliary converter in parallel with the grid-side converter, but also a coordinated control strategy, is proposed to enhance the low voltage ride through (LVRT capability and improve power quality. During normal operation, the main grid-side converter maintains the DC-link voltage constant, whereas the auxiliary grid-side converter functions as an APF with harmonic suppression and reactive power compensation to improve the power quality. During grid faults, a hierarchical coordinated control scheme for the generator-side converter, main grid-side converter and auxiliary grid-side converter, depending on the grid voltage sags, is presented to enhance the LVRT capability of the direct-driven PMSG WT. The feasibility and the effectiveness of the proposed system’s topology and hierarchical coordinated control strategy were verified using MATLAB/Simulink simulations.

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

    International Nuclear Information System (INIS)

    Sun Shuhai; Li Huasheng; Zhang Hao

    2012-01-01

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

  10. Performance evaluation of a state-of-the-art solar air-heating system with auxiliary heat pump

    Science.gov (United States)

    1980-01-01

    The system in Solar House 2 consists of 57.9 sq. m. of Solaron Series 300 Collectors, 10.3 cu. m. of pebble bed storage, domestic water preheating capability and a Carrier air-to-air heat pump as an auxiliary heater. Although the control subsystem was specially constructed to facilitate experimental changes and data reduction, the balance of the solar system was assembled with off-the-shelf components. Since all components of the system are commercially available the system is considered to be a state of the art solar air-heating system. The system design is one that is recommended for residential and small office buildings.

  11. Numerical analysis of magnetically suspended rotor in HTR-10 helium circulator being dropped into auxiliary bearings

    International Nuclear Information System (INIS)

    Zhao Jingxiong; Yang Guojun; Li Yue; Yu Suyuan

    2012-01-01

    Active magnetic bearings (AMB) have been selected to support the rotor of primary helium circulator in commercial 10 Mega-Walt High Temperature Gas-cooled Reactor (HTR-10). In an AMB system, the auxiliary bearings are necessary to protect the AMB components in case of losing power. This paper performs the impact simulation of Magnetically Suspended Rotor in HTR-10 Helium Circulator being dropped into the auxiliary bearings using the finite element program ABAQUS. The dynamic response and the strain field of auxiliary bearings are analyzed. The results achieved by the numerical analysis are in agreement with the experiment results. Therefore, the feasibility of the design of auxiliary bearing and the possibility of using the AMB system in the HTR are proved. (authors)

  12. A multivariate statistical study on a diversified data gathering system for nuclear power plants

    International Nuclear Information System (INIS)

    Samanta, P.K.; Teichmann, T.; Levine, M.M.; Kato, W.Y.

    1989-02-01

    In this report, multivariate statistical methods are presented and applied to demonstrate their use in analyzing nuclear power plant operational data. For analyses of nuclear power plant events, approaches are presented for detecting malfunctions and degradations within the course of the event. At the system level, approaches are investigated as a means of diagnosis of system level performance. This involves the detection of deviations from normal performance of the system. The input data analyzed are the measurable physical parameters, such as steam generator level, pressurizer water level, auxiliary feedwater flow, etc. The study provides the methodology and illustrative examples based on data gathered from simulation of nuclear power plant transients and computer simulation of a plant system performance (due to lack of easily accessible operational data). Such an approach, once fully developed, can be used to explore statistically the detection of failure trends and patterns and prevention of conditions with serious safety implications. 33 refs., 18 figs., 9 tabs

  13. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  14. Simulative technology for auxiliary fuel tank separation in a wind tunnel

    Directory of Open Access Journals (Sweden)

    Ma Xin

    2016-06-01

    Full Text Available In this paper, we propose a simulative experimental system in wind tunnel conditions for the separation of auxiliary fuel tanks from an aircraft. The experimental system consists of a simulative release mechanism, a scaled model and a pose measuring system. A new release mechanism was designed to ensure stability of the separation. Scaled models of the auxiliary fuel tank were designed and their moment of inertia was adjusted by installing counterweights inside the model. Pose parameters of the scaled model were measured and calculated by a binocular vision system. Additionally, in order to achieve high brightness and high signal-to-noise ratio of the images in the dark enclosed wind tunnel, a new high-speed image acquisition method based on miniature self-emitting units was presented. Accuracy of the pose measurement system and repeatability of the separation mechanism were verified in the laboratory. Results show that the position precision of the pose measurement system can reach 0.1 mm, the precision of the pitch and yaw angles is less than 0.1° and that of the roll angle can be up to 0.3°. Besides, repeatability errors of models’ velocity and angular velocity controlled by the release mechanism remain small, satisfying the measurement requirements. Finally, experiments for the separation of auxiliary fuel tanks were conducted in the laboratory.

  15. Basic information about development and construction of a PWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1977-01-01

    1.0) Plant layout of a PWR; 2.0) principle design of a PWR and the reactor coolant system; 3.0) reactor auxiliary and ancillary systems; 3.1) volume control system; 3.2) boric acid control and chemical feeding system; 3.3) coolant purification and degassing system; 3.4) coolant storage and treatment system; 3.5) nuclear component cooling system; 3.6) liquid waste processing system; 3.7) gaseous waste processing system; 4.0) residual heat removal system; 5.0) emergency feedwater system; 6.0) containment design; 7.0) fuel handling, storage and transport system in a PWR. (orig.) [de

  16. Safety analysis of high temperature reactor cooled and moderated by supercritical light water

    International Nuclear Information System (INIS)

    Ishiwatari, Yuki; Oka, Yoshiaki; Koshizuka, Seiichi

    2003-01-01

    This paper describes 'Safety' of a high temperature supercritical light water cooled and moderated reactor (SCRLWR-H) with descending flow water rods. The safety system of the SCLWR-H is similar to that of a BWR. It consists of reactor scram, high pressure auxiliary feedwater system (AFS), low pressure core injection system (LPCI), safety relief valves (SRV), automatic depressurization system (ADS), and main steam isolation valves (MSIV). Ten types of transients and five types of accidents are analyzed using a plant transient analysis code SPRAT-DOWN. The sequences are determined referring to LWRs. At the 'Loss of load without turbine bypass' transient, the coolant density and the core power are increased by the over-pressurization, and at the same time the core flow rate is decreased by the closure of the turbine control valves. The peak cladding temperature increases to 727degC. The high temperature at this type of transient is one of the characteristics of the SCLWR-H. Conversely at 'feedwater-loss' events, the core power decrease to some extend by density feedback before the reactor scram. The peak cladding temperatures at the 'Partial loss of feedwater' transient and the 'Total loss of feedwater' accident are only 702degC and 833degC, respectively. The cladding temperature does not increase so much at the transients 'Loss of feedwater heating' and 'CR withdrawal' because of the operation of the plant control system. All the transients and accidents satisfy the satisfy criteria with good margins. The highest cladding temperatures of the transients and the accidents are 727degC and 833degC at the 'Loss of load without turbine bypass' and 'Total loss of feedwater', respectively. The duration of the high cladding temperature is very short at the transients. According to the parametric survey, the peak cladding temperature are sensitive to the parameters such as the pump coast-down time, delay of pump trip, AFS capacity, AFS delay, CR worth, and SRV setpoint

  17. Computation within the auxiliary field approach

    International Nuclear Information System (INIS)

    Baeurle, S.A.

    2003-01-01

    Recently, the classical auxiliary field methodology has been developed as a new simulation technique for performing calculations within the framework of classical statistical mechanics. Since the approach suffers from a sign problem, a judicious choice of the sampling algorithm, allowing a fast statistical convergence and an efficient generation of field configurations, is of fundamental importance for a successful simulation. In this paper we focus on the computational aspects of this simulation methodology. We introduce two different types of algorithms, the single-move auxiliary field Metropolis Monte Carlo algorithm and two new classes of force-based algorithms, which enable multiple-move propagation. In addition, to further optimize the sampling, we describe a preconditioning scheme, which permits to treat each field degree of freedom individually with regard to the evolution through the auxiliary field configuration space. Finally, we demonstrate the validity and assess the competitiveness of these algorithms on a representative practical example. We believe that they may also provide an interesting possibility for enhancing the computational efficiency of other auxiliary field methodologies

  18. Bayesian Analysis of Geostatistical Models With an Auxiliary Lattice

    KAUST Repository

    Park, Jincheol; Liang, Faming

    2012-01-01

    of observations is large. In this article, we propose an auxiliary lattice-based approach for tackling this difficulty. By introducing an auxiliary lattice to the space of observations and defining a Gaussian Markov random field on the auxiliary lattice, our model

  19. 47 CFR 73.1675 - Auxiliary antennas.

    Science.gov (United States)

    2010-10-01

    ... Class A TV licensees may request a decrease from the authorized facility's ERP in the license application. An FM, TV or Class A TV licensee may also increase the ERP of the auxiliary facility in a license... licensed main facility as an auxiliary facility with an ERP less than or equal to the ERP specified on the...

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

    International Nuclear Information System (INIS)

    Finicle, D.P.

    1977-01-01

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

  1. 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.

  2. Power-feedwater enthalpy operating domain for SBWR applying Monte Carlo simulation

    International Nuclear Information System (INIS)

    Quezada-Garcia, S.; Espinosa-Martinez, E.-G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G.

    2014-01-01

    In this work the analyses of the feedwater enthalpy effects on reactor power in a simplified boiling water reactor (SBWR) applying a methodology based on Monte Carlo's simulation (MCS), is presented. The MCS methodology was applied systematically to establish operating domain, due that the SBWR are not yet in operation, the analysis of the nuclear and thermalhydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. (author)

  3. Design and setting up of a system for remote monitoring and control on auxiliary machines in electric vehicles

    Directory of Open Access Journals (Sweden)

    Dimitrov Vasil

    2017-01-01

    Full Text Available Systems for remote monitoring and control of the proper operation, energy consumption, and efficiency of the controlled objects are very often used in different spheres of industry, in the electricity distribution network, etc. Various types of intelligent energy meters, PLCs and other control devices are involved in such systems. Proper operation of the auxiliary machines in electric vehicles is of great importance and implementation of a system for their remote monitoring and control is useful and ensures reliability and increased efficiency. A system has been designed and built using contemporary devices. An asynchronous motor is controlled by a soft starter and opportunities for remote monitoring (by an intelligent energy meter and control (by a PLC and Touch panel have been provided. Soft starters are widely used in industry for control on asynchronous drives when speed regulation is not a mandatory requirement. They are cheaper than inverters and frequency converters and allow for temporal reduction of the torque and current surge during start-up, as well as smooth deceleration. Therefore they can also be used in electric vehicles to control auxiliary machines (pumps, fans, air coolers, compressors, etc.. The present paper presents a methodology for their design and setting up.

  4. 14 CFR 29.757 - Hull and auxiliary float strength.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Hull and auxiliary float strength. 29.757... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Design and Construction Floats and Hulls § 29.757 Hull and auxiliary float strength. The hull, and auxiliary floats if used, must withstand the...

  5. Space shuttle auxiliary propulsion system design study. Phase C and E report: Storable propellants, RCS/OMS/APU integration study

    Science.gov (United States)

    Anglim, D. D.; Bruns, A. E.; Perryman, D. C.; Wieland, D. L.

    1972-01-01

    Auxiliary propulsion concepts for application to the space shuttle are compared. Both monopropellant and bipropellant earth storable reaction control systems were evaluated. The fundamental concepts evaluated were: (1) monopropellant and bipropellant systems installed integrally within the vehicle, (2) fuel systems installed modularly in nose and wing tip pods, and (3) fuel systems installed modularly in nose and fuselage pods. Numerous design variations within these three concepts were evaluated. The system design analysis and methods for implementing each of the concepts are reported.

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

    International Nuclear Information System (INIS)

    Tijerina S, F.

    2008-01-01

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

  7. 78 FR 27321 - Revision of Auxiliary Regulations

    Science.gov (United States)

    2013-05-10

    ... Auxiliary organizational structure, extending civil liability protection to Auxiliary units and members, and... entities. C. Assistance for Small Entities If you think that your business, organization, or governmental..., explain why you think your business or organization qualifies, how and to what degree this rule would...

  8. Vacuum switchgear for power station auxiliary switchboards

    International Nuclear Information System (INIS)

    Coombs, P.E.

    1992-01-01

    Sizewell B is the first UK power station in which vacuum switchgear is used for the auxiliary switchboards. Previously the 3.3kV, 6.6kV or 11kV switchgear has used air-break circuit breakers and fused air-break contactors, known as motor starting devices or fused switching devices (FSD). The use of vacuum interrupters is therefore a new technology in this application, although it has been established in the UK distribution network and in industrial installations from the mid 1970s. Vacuum switchgear was already in use in the USA for power station auxiliary switchgear at the time that it was proposed for Sizewell B. The Sizewell B high voltage auxiliary switchgear comprises eight Unit and Station Auxiliary Switchboards at 3.3kV and 11kV, and four 3.3kV Essential Switchboards for the essential safety related circuits, making a total of 65 circuit breakers plus FSD panels. (Author)

  9. Test facility for auxiliary cooling system (ACS) of fast breeder reactor for Power Reactor and Nuclear Fuel Development Corporation (PNC)

    International Nuclear Information System (INIS)

    1983-01-01

    In preparation of constructing ''Monju'', a prototype fast breeder reactor, PNC has been pushing forward its research and development projects and the ACS was constructed under these projects. The auxiliary cooling system is an important engineered safety feature, and is used for safe removal of heat from the reactor at the shutdown. The ACS serves as a means of testing and assessing the auxiliary cooling system for the ''Monju'' and is designed and manufactured to have one fifth capacity of the Monju. The air heat exchanger and the ACS system was designed to withstand higher temperature range of the conventional design code (MITI-501), and finned tubes were applied for effective heat removal. Preheating system was designed to heat up the whole system over 200 0 C within 20 hours to prevent sodium from freezing. Basic performance of ACS was verified satisfactorily by a series of performance tests, such as start up test, flow rate measurement and preheating test before delivery. The experience from designing and construction of ACS and data obtained by these tests will be very instructive for designing and construction of the ''Monju''. (author)

  10. Study on decay heat removal capability of reactor vessel auxiliary cooling system

    International Nuclear Information System (INIS)

    Nishi, Y.; Kinoshita, I.

    1991-01-01

    The reactor vessel auxiliary cooling system (RVACS) is a simple, Passive decay heat removal system for an LMFBR. However, the heat removal capacity of this system is small compared to that of an immersed type of decay heat exchanger. In this study, a high-porosity porous body is proposed to enhance the RVACS's heat transfer performance to improve its applicability. The objectives of this study are to propose a new method which is able to use thermal radiation effectively, to confirm its heat removal capability and to estimate its applicability limit of RVACS for an LMFBR. Heat transfer tests were conducted in an experimental facility with a 3.5 m heat transfer height to evaluate the heat transfer performance of the high-porosity porous body. Using the experimental results, plant transient analyses were performed for a 300 MWe pool type LMFBR under a Total Black Out (TBO) condition to confirm the heat removal capability. Furthermore, the relationship between heat removal capability and thermal output of a reactor were evaluated using a simple parameter model

  11. 30 CFR 57.22209 - Auxiliary fans (I-C mines).

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Auxiliary fans (I-C mines). 57.22209 Section 57... Standards for Methane in Metal and Nonmetal Mines Ventilation § 57.22209 Auxiliary fans (I-C mines). Electric auxiliary fans shall be approved by MSHA under the applicable requirements of 30 CFR part 18...

  12. Improving Energy Efficiency of Auxiliaries

    International Nuclear Information System (INIS)

    Carl T. Vuk

    2001-01-01

    The summaries of this report are: Economics Ultimately Dictates Direction; Electric Auxiliaries Provide Solid Benefits. The Impact on Vehicle Architecture Will be Important; Integrated Generators With Combined With Turbo Generators Can Meet the Electrical Demands of Electric Auxiliaries; Implementation Will Follow Automotive 42V Transition; Availability of Low Cost Hardware Will Slow Implementation; Industry Leadership and Cooperation Needed; Standards and Safety Protocols Will be Important. Government Can Play an Important Role in Expediting: Funding Technical Development; Incentives for Improving Fuel Economy; Developing Standards, Allowing Economy of Scale; and Providing Safety Guidelines

  13. Precision improvement of frequency-modulated continuous-wave laser ranging system with two auxiliary interferometers

    Science.gov (United States)

    Shi, Guang; Wang, Wen; Zhang, Fumin

    2018-03-01

    The measurement precision of frequency-modulated continuous-wave (FMCW) laser distance measurement should be proportional to the scanning range of the tunable laser. However, the commercial external cavity diode laser (ECDL) is not an ideal tunable laser source in practical applications. Due to the unavoidable mode hopping and scanning nonlinearity of the ECDL, the measurement precision of FMCW laser distance measurements can be substantially affected. Therefore, an FMCW laser ranging system with two auxiliary interferometers is proposed in this paper. Moreover, to eliminate the effects of ECDL, the frequency-sampling method and mode hopping influence suppression method are employed. Compared with a fringe counting interferometer, this FMCW laser ranging system has a measuring error of ± 20 μm at the distance of 5.8 m.

  14. Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Carroll, D.G.; Chen, C.; Crane, C.; Dalton, R.; Taylor, J.R.; Tosunoglu, S.; Weymouth, T.

    1993-01-01

    One of the most important safety systems in General Electric's (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE/NE Robotics for Advanced Reactors program formed a team to evaluate the ALMR design for remote inspection of the RVACS. Conceptual designs for robots to perform the required inspection tasks were developed by the team. Design criteria for these remote systems included robot deployment, power supply, navigation, environmental hardening of components, tether management, communication with an operator, sensing, and failure recovery. The operation of the remote inspection concepts were tested using 3-D simulation models of the ALMR. In addition, the team performed an extensive technology review of robot components that could survive the environmental conditions in the RVACS

  15. Studying the dynamical characteristics of pumps in NPP unit auxiliary water system under operational conditions

    International Nuclear Information System (INIS)

    Belyaev, S.G.; Puzanov, A.I.; Belikov, V.P.; Dizik, B.S.

    1990-01-01

    Hydrodynamic loads appearing in the flow-through part of pump aggregates of the system of auxiliary water supply in NPP with variations in the operation modes are investigated. It is shown that during operation of centrifugal pumps the position of the mode on the pump characteristic plot must be controlled. When the mode point exceeds the limits of the working zone it results in a considerable increase of dynamic loads: pressure pulsation and vibration. As the flow rate increase the decrease in dynamic loads is recorded at low frequencies of about 2-4 Hz

  16. Remote visual testing (RVT) for the diagnostic inspection of feedwater heaters

    International Nuclear Information System (INIS)

    Nugent, M.J.; Pellegrino, B.A.

    1991-01-01

    In this paper the benefits and limitations of Non-Destructive Testing (NDT) on feedwater heaters will be briefly reviewed. All Remote Visual Testing (RVT) devices including borescopes, fiberscopes, videoborescopes and Closed Circuit Television (CCTV) cameras will be discussed along with currently accepted formats for documentation. The benefits of a comprehensive in-place inspection involving Remote Visual Testing will be discussed in relationship to its diagnostic capabilities. The results of eight post-service heater inspections will be discussed along with the root cause of failure of seven unique failure mechanisms. These inspections, including FWH access, RVT tool and data analysis, will be detailed

  17. Analysis of loss of offsite power transient using RELAP5/MOD1/NSC

    International Nuclear Information System (INIS)

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

    1986-01-01

    System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAP5/MOD1/NSC), based upon the sequence of events for the KNU1( Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximate the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV setting etc. are recognized to be important in the transient analyses on a bestestimate basis. (Author)

  18. Energy Recovery for the Main and Auxiliary Sources of Electric Vehicles

    Directory of Open Access Journals (Sweden)

    Binggang Cao

    2010-10-01

    Full Text Available Based on the traditional regenerative braking electrical circuit, a novel energy recovery system for the main and auxiliary sources of electric vehicles (EVs has been developed to improve their energy efficiency. The electrical circuit topology is presented in detail. During regenerative braking, the recovered mechanical energy is stored in both the main source and the auxiliary source at the same time. The mathematical model of the proposed system is derived step by step. Combining the merits and defects of H2 optimal control and H∞ robust control, a H2/H∞ controller is designed to guarantee both the system performance and robust stability. The perfect match between the simulated and experimental results validates the notion that the proposed novel energy recovery system is both feasible and effective, as more energy is recovered than that with the traditional energy recovery systems, in which recovered energy is stored only in the main source.

  19. Energy Recovery for the Main and Auxiliary Sources of Electric Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Min Ye [Key Laboratory for Highway Construction Technology and Equipment of Ministry of Education, Chang’an University, Xi’an (China); Sengjie Jiao [Key Laboratory for Highway Construction Technology and Equipment of Ministry of Education, Chang’an University, Xi’an (China); Binggang Cao [School of Mechanical Engineering, Xi’an Jiaotong University, Xi’an (China)

    2010-09-15

    Based on the traditional regenerative braking electrical circuit, a novel energy recovery system for the main and auxiliary sources of electric vehicles (EVs) has been developed to improve their energy efficiency. The electrical circuit topology is presented in detail. During regenerative braking, the recovered mechanical energy is stored in both the main source and the auxiliary source at the same time. The mathematical model of the proposed system is derived step by step. Combining the merits and defects of H2 optimal control and H-infinity robust control, a H2/H-infinity controller is designed to guarantee both the system performance and robust stability. The perfect match between the simulated and experimental results validates the notion that the proposed novel energy recovery system is both feasible and effective, as more energy is recovered than that with the traditional energy recovery systems, in which recovered energy is stored only in the main source.

  20. Energy Recovery for the Main and Auxiliary Sources of Electric Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Ye, M.; Jiao, S. [Key Laboratory for Highway Construction Technology and Equipment of Ministry of Education, Chang' an University, Xi' an 710064 (China); Cao, B. [School of Mechanical Engineering, Xi' an Jiaotong University, Xi' an 710049 (China)

    2010-10-15

    Based on the traditional regenerative braking electrical circuit, a novel energy recovery system for the main and auxiliary sources of electric vehicles (EVs) has been developed to improve their energy efficiency. The electrical circuit topology is presented in detail. During regenerative braking, the recovered mechanical energy is stored in both the main source and the auxiliary source at the same time. The mathematical model of the proposed system is derived step by step. Combining the merits and defects of H{sub 2} optimal control and H{sub {infinity}} robust control, a H{sub 2}/H{sub {infinity}} controller is designed to guarantee both the system performance and robust stability. The perfect match between the simulated and experimental results validates the notion that the proposed novel energy recovery system is both feasible and effective, as more energy is recovered than that with the traditional energy recovery systems, in which recovered energy is stored only in the main source. (authors)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  2. Evaluation of heat exchange performance for the auxiliary component cooling water system cooling tower in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Kameyama, Yasuhiko; Shimizu, Atsushi; Inoi, Hiroyuki; Yamazaki, Kazunori; Shimizu, Yasunori; Aragaki, Etsushi; Ota, Yukimaru; Fujimoto, Nozomu

    2006-09-01

    The auxiliary component cooling water system (ACCWS) is one of the cooling system in High Temperature Engineering Test Reactor (HTTR). The ACCWS has main two features, many facilities cooling, and heat sink of the vessel cooling system which is one of the engineering safety features. Therefore, the ACCWS is required to satisfy the design criteria of heat removal performance. In this report, heat exchange performance data of the rise-to-power-up test and the in-service operation for the ACCWS cooling tower was evaluated. Moreover, the evaluated values were compared with the design values, and it is confirmed that ACCWS cooling tower has the required heat exchange performance in the design. (author)

  3. Evaluation of heatup and recovery in a loss of feedwater accident with multiple failure

    International Nuclear Information System (INIS)

    Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung

    1991-01-01

    A loss of feedwater accident with multiple failure has been studied in order to identify the potential severity of the accident when compared with the design basis accident in PWR. The PCS heatup and recovery mode in a LOFA with multiple failure was evaluated using the LOFT L9-1/L3-3 experiment. From experimental result, 4 separable subphase were identified and the associated phenomena were also addressed

  4. 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

  5. Tracer test method and process data reconciliation based on VDI 2048. Comparison of two methods for highly accurate determination of feedwater massflow at NPP Beznau

    International Nuclear Information System (INIS)

    Hungerbuehler, T.; Langenstein, M.

    2007-01-01

    The feedwater mass flow is the key measured variable used to determine the thermal reactor output in a nuclear power plant. Usually this parameter is recorded via venturi nozzles of orifice plates. The problem with both principles of measurement, however, is that an accuracy of below 1% cannot be reached. In order to make more accurate statements about the feedwater amounts recirculated in the water-steam cycle, tracer measurements that offer an accuracy of up to 0.2% are used. In the NPP Beznau both methods have been used in parallel to determine the feedwater flow rates in 2004 (unit 1) and 2005 (unit 2). Comparison of the results shows that a high level of agreement is obtained between the results of the reconciliation and the results of the tracer measurements. As a result of the findings of this comparison, a high level of acceptance of process data reconciliation based on VDI 2048 was achieved. (orig.)

  6. Feed-water heaters alternative design comparison; Comparacion de disenos alternativos de calentadores

    Energy Technology Data Exchange (ETDEWEB)

    Torres Toledano, Gerardo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1988-12-31

    A procedure is presented for the alternative design comparison of feed water heaters, based in the failure records of damaged tubes during operation. The procedure is used for cases in which non-continuous or random inspections are made to the feed-water heaters. [Espanol] Se presenta un procedimiento para comparar disenos alternativos de calentadores, basandose en los registros de fallas de los tubos rotos acumuladas durante su operacion. El procedimiento se emplea para casos en los que se realizan inspecciones a los calentadores no continuas, ya sea periodicas o al azar.

  7. Data report for ROSA-IV LSTF 10% hot leg break experiment Run SB-HL-02

    International Nuclear Information System (INIS)

    Kukita, Yutaka; Hirata, Kazuo; Gotou, Hiroki

    1990-03-01

    Experimental data for the 10% hot leg break test, Run SB-HL-02, conducted at the ROSA-IV Large Scale Test Facility (LSTF) on June 30, 1987, are presented. This test assumed total failure of both high pressure injection (HPI) and auxiliary feedwater (AFW) systems. The test results were characterized by asymmetric loop responses, flashing in the cold legs and upper downcomer, as well as condensation depressurization in the cold legs following injection of emergency core coolant (ECC) from accumulators. (author)

  8. 47 CFR 74.601 - Classes of TV broadcast auxiliary stations.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 4 2010-10-01 2010-10-01 false Classes of TV broadcast auxiliary stations. 74... Television Broadcast Auxiliary Stations § 74.601 Classes of TV broadcast auxiliary stations. (a) TV pickup stations. A land mobile station used for the transmission of TV program material and related communications...

  9. Effects of Auxiliary-Source Connection in Multichip Power Module

    DEFF Research Database (Denmark)

    Li, Helong; Munk-Nielsen, Stig; Wang, Xiongfei

    2017-01-01

    the power loop and the gate loop like how the Kelvin-source connection does, owing to their involvement in the loop of the power source current. Three effects of the auxiliary-source connections are then analyzed, which are 1) the common source stray inductance reduction, 2) the transient drain......Auxiliary-source bond wires and connections are widely used in power modules with paralleled MOSFETs or IGBTs. This paper investigates the operation mechanism of the auxiliary-source connections in multichip power modules. It reveals that the auxiliary-source connections cannot fully decouple......-source current imbalance mitigation, and 3) the influence on the steady-state current distribution. Lastly, simulations and experimental results validate the theoretical analysis....

  10. Simulation of loss of feedwater transient of MASLWR test facility by MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Juyeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is being current developed domestically also adopts helical coil steam generator, KINS has joined this ICSP to evaluate performance of domestic regulatory audit thermal-hydraulic code (MARS-KS code) in various respects including wall-to-fluid heat transfer model modification implemented in the code by independent international experiment database. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3). In the present study, KINS simulation results by the MARS-KS code (KS-002 version) for the SP-2 experiment are presented in detail and conclusions on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the loss of feedwater transient of the MASLWR test facility. Steady state run shows helical coil specific heat transfer models implemented in the code is reasonable. However, through the transient run, it is also found that three-dimensional effect within the HPC and axial conduction effect through the HTP are not well reproduced by the code.

  11. Evaluation method for two-phase flow and heat transfer in a feed-water heater

    International Nuclear Information System (INIS)

    Takamori, Kazuhide; Minato, Akihiko

    1993-01-01

    A multidimensional analysis code for two-phase flow using a two-fluid model was improved by taking into consideration the condensation heat transfer, film thickness, and film velocity, in order to develop an evaluation method for two-phase flow and heat transfer in a feed-water heater. The following results were obtained by a two-dimensional analysis of a feed-water heater for a power plant. (1) In the model, the film flowed downward in laminar flow due to gravity, with droplet entrainment and deposition. For evaluation of the film thickness, Fujii's equation was used in order to account for forced convection of steam flow. (2) Based on the former experimental data, the droplet deposition coefficient and droplet entrainment rate of liquid film were determined. When the ratio at which the liquid film directly flowed from an upper heat transfer tube to a lower heat transfer tube was 0.7, the calculated total heat transfer rate agreed with the measured value of 130 MW. (3) At the upper region of a heat transfer tube bundle where film thickness was thin, and at the outer region of a heat transfer tube bundle where steam velocity was high, the heat transfer rate was large. (author)

  12. 30 CFR 57.8534 - Shutdown or failure of auxiliary fans.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Shutdown or failure of auxiliary fans. 57.8534... Ventilation Underground Only § 57.8534 Shutdown or failure of auxiliary fans. (a) Auxiliary fans installed and... fan maintenance or fan adjustments where air quality is maintained in compliance with the applicable...

  13. Auxiliary controller for time-to-digital converter module readout

    International Nuclear Information System (INIS)

    Ermolin, Yu.V.

    1992-01-01

    The KD-225 auxiliary controller for time-to-digital converter module readout in the SUMMA crate is described. After readout and preliminary processing the data are written in the P-140 buffer memory module. The controller is used in the FODS-2 experimental setup data acquisition system. 12 refs.; 1 fig

  14. Builtin vs. auxiliary detection of extrapolation risk.

    Energy Technology Data Exchange (ETDEWEB)

    Munson, Miles Arthur; Kegelmeyer, W. Philip,

    2013-02-01

    A key assumption in supervised machine learning is that future data will be similar to historical data. This assumption is often false in real world applications, and as a result, prediction models often return predictions that are extrapolations. We compare four approaches to estimating extrapolation risk for machine learning predictions. Two builtin methods use information available from the classification model to decide if the model would be extrapolating for an input data point. The other two build auxiliary models to supplement the classification model and explicitly model extrapolation risk. Experiments with synthetic and real data sets show that the auxiliary models are more reliable risk detectors. To best safeguard against extrapolating predictions, however, we recommend combining builtin and auxiliary diagnostics.

  15. National Ignition Facility subsystem design requirements target area auxiliary subsystem SSDR 1.8.6

    International Nuclear Information System (INIS)

    Reitz, T.

    1996-01-01

    This Subsystem Design Requirement (SSDR) establishes the performance, design, development, and test requirements for the Target Area Auxiliary Subsystems (WBS 1.8.6), which is part of the NIF Target Experimental System (WBS 1.8). This document responds directly to the requirements detailed in NIF Target Experimental System SDR 003 document. Key elements of the Target Area Auxiliary Subsystems include: WBS 1.8.6.1 Local Utility Services; WBS 1.8.6.2 Cable Trays; WBS 1.8.6.3 Personnel, Safety, and Occupational Access; WBS 1.8.6.4 Assembly, Installation, and Maintenance Equipment; WBS 1.8.6.4.1 Target Chamber Service System; WBS 1.8.6.4.2 Target Bay Service Systems

  16. Design and Experiment of Auxiliary Bearing for Helium Blower of HTR-PM

    International Nuclear Information System (INIS)

    Yang Guojun; Shi Zhengang; Liu Xingnan; Zhao Jingjing

    2014-01-01

    The helium blower is the important equipment for HTR-PM. Active magnetic bearing (AMB) instead of mechanical bearing is selected to support the rotor of the helium blower. However, one implication of AMB is the requirement to provide the auxiliary bearing to mitigate the effects of failures or overload conditions. The auxiliary bearing is used to support the rotor when the AMB fails to work. It must support the dropping rotor and bear the great impact force and friction heat. The design of the auxiliary bearing is one of the challenging problems in the whole system. It is very important for the helium blower with AMB of HTR-PM to make success. The rotor’s length of helium blower of HTR-PM is about 3.3 m, its weight is about 4000 kg and the rotating speed is 4000 r/min. The axial load is 4500kg, and the radial load is 1950kg. The angular contact ball bearing was selected as the auxiliary bearing. The test rig has been finished. It is difficult to analyze the falling course of the rotor. The preliminary analysis of the dropping rotor was done in the special condition. The impact force of auxiliary bearing was computed for the axial and radial load. And the dropping test of the blower rotor for HTR-10 will be introduced also in this paper. Results offer the important theoretical base for the protector design of the helium blower with AMB for HTR-PM. (author)

  17. MTX [Microwave Tokamak Experiment] diagnostic and auxiliary systems for confinement, transport, and plasma physics studies

    International Nuclear Information System (INIS)

    Hooper, E.B.; Allen, S.L.; Casper, T.A.; Thomassen, K.I.

    1989-01-01

    This note describes the diagnostics and auxiliary systems on the Microwave Tokamak Experiment (MTX) for confinement, transport, and other plasma physics studies. It is intended as a reference on the installed and planned hardware on the machine for those who need more familiarity with this equipment. Combined with the tokamak itself, these systems define the opportunities and capabilities for experiments in the MTX facility. We also illustrate how these instruments and equipment are to be used in carrying out the MTX Operations Plan. Near term goals for MTX are focussed on the absorption and heating by the microwave beam from the FEL, but the Plan also includes using the facility to study fundamental phenomena in the plasma, to control MHD activity, and to drive current noninductively

  18. 14 CFR 25.1142 - Auxiliary power unit controls.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Auxiliary power unit controls. 25.1142... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Powerplant Controls and Accessories § 25.1142 Auxiliary power unit controls. Means must be provided on the flight deck for starting...

  19. 14 CFR 23.1142 - Auxiliary power unit controls.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Auxiliary power unit controls. 23.1142... Powerplant Controls and Accessories § 23.1142 Auxiliary power unit controls. Means must be provided on the... power unit. [Doc. No. 26344, 58 FR 18974, Apr. 9, 1993] ...

  20. 14 CFR 29.1142 - Auxiliary power unit controls.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Auxiliary power unit controls. 29.1142... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Powerplant Controls and Accessories § 29.1142 Auxiliary power unit controls. Means must be provided on the flight deck for starting...

  1. Computation of Ground-State Properties in Molecular Systems: Back-Propagation with Auxiliary-Field Quantum Monte Carlo.

    Science.gov (United States)

    Motta, Mario; Zhang, Shiwei

    2017-11-14

    We address the computation of ground-state properties of chemical systems and realistic materials within the auxiliary-field quantum Monte Carlo method. The phase constraint to control the Fermion phase problem requires the random walks in Slater determinant space to be open-ended with branching. This in turn makes it necessary to use back-propagation (BP) to compute averages and correlation functions of operators that do not commute with the Hamiltonian. Several BP schemes are investigated, and their optimization with respect to the phaseless constraint is considered. We propose a modified BP method for the computation of observables in electronic systems, discuss its numerical stability and computational complexity, and assess its performance by computing ground-state properties in several molecular systems, including small organic molecules.

  2. Auxiliary systems

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1981-01-01

    Systems included under the heading ''Reactor Auxillary Systems'' are those immediately involved with the reactor operation. These include the systems for dosing and letdown of reactor coolant, as well as for the chemical dosing, purification and treatment of the reactor coolant and the cooling system in the controlled area. The ancillary systems are mainly responsible for liquid and gaseous treatment and the waste treatment for final storage. (orig.)

  3. Comparative Study of Electric Energy Storages and Thermal Energy Auxiliaries for Improving Wind Power Integration in the Cogeneration System

    Directory of Open Access Journals (Sweden)

    Yanjuan Yu

    2018-01-01

    Full Text Available In regards to the cogeneration system in Northern China, mainly supported by combined heat and power (CHP plants, it usually offers limited operation flexibility due to the joint production of electric and thermal power. For that large-scale wind farms included in the cogeneration system, a large amount of wind energy may have to be wasted. To solve this issue, the utilization of the electric energy storages and the thermal energy auxiliaries are recommended, including pumped hydro storage (PHS, compressed air energy storage (CAES, hydrogen-based energy storage (HES, heat storage (HS, electric boilers (EB, and heat pumps (HP. This paper proposes a general evaluation method to compare the performance of these six different approaches for promoting wind power integration. In consideration of saving coal consumption, reducing CO2 emissions, and increasing investment cost, the comprehensive benefit is defined as the evaluation index. Specifically, a wind-thermal conflicting expression (WTCE is put forward to simplify the formulation of the comprehensive benefit. Further, according to the cogeneration system of the West Inner Mongolia (WIM power grid, a test system is modelled to perform the comparison of the six different approaches. The results show that introducing the electric energy storages and the thermal energy auxiliaries can both contribute to facilitating wind power integration, and the HP can provide the best comprehensive benefit.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  5. Design and operation of the LAMPF Auxiliary Controller. High-speed remote processing on the CAMAC dataway

    International Nuclear Information System (INIS)

    Machen, D.R.

    1979-02-01

    A CAMAC Auxiliary Controller has been developed to further the concepts of distributed processing in both process control and experiment data-acquisition systems. The Auxiliary Controller is built around a commercially available 16-bit microcomputer and a high-speed bit-sliced microprocessor capable of instruction execution times of 140 ns. The modular nature of the controller allows the user to tailor the controller capabilities to the system problem, while maintaining the interface techniques of the CAMAC Standard

  6. Dedicated auxiliary power units for Hybrid Electric Vehicles

    NARCIS (Netherlands)

    Mourad, S.; Weijer, C.J.T. van de

    1998-01-01

    The use of a dedicated auxiliary power unit is essential to utilize the potential that hybrid vehicles offer for efficient and ultra-clean transportation. An example of a hybrid project at the TNO Road-Vehicles Research Institute shows the development and the results of a dedicated auxiliary power

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  8. Solar combisystems with forecast control to increase the solar fraction and lower the auxiliary energy cost

    DEFF Research Database (Denmark)

    Perers, Bengt; Furbo, Simon; Fan, Jianhua

    2011-01-01

    Solar Combi systems still need quite a lot of auxiliary energy especially in small systems without seasonal storage possibilities. The control of the auxiliary energy input both in time and power is important to utilize as much as possible of the solar energy available from the collectors and also...... energy sources. It can be either direct electric heating elements or a heat pump upgrading ambient energy in the air, ground, solar collector or waste heat from the house. The paper describes system modeling and simulation results. Advanced laboratory experiments are also starting now with three...

  9. Terminal Sliding Mode Control with Unidirectional Auxiliary Surfaces for Hypersonic Vehicles Based on Adaptive Disturbance Observer

    Directory of Open Access Journals (Sweden)

    Naibao He

    2015-01-01

    Full Text Available A novel flight control scheme is proposed using the terminal sliding mode technique, unidirectional auxiliary surfaces and the disturbance observer model. These proposed dynamic attitude control systems can improve control performance of hypersonic vehicles despite uncertainties and external disturbances. The terminal attractor is employed to improve the convergence rate associated with the critical damping characteristics problem noted in short-period motions of hypersonic vehicles. The proposed robust attitude control scheme uses a dynamic terminal sliding mode with unidirectional auxiliary surfaces. The nonlinear disturbance observer is designed to estimate system uncertainties and external disturbances. The output of the disturbance observer aids the robust adaptive control scheme and improves robust attitude control performance. Finally, simulation results are presented to illustrate the effectiveness of the proposed terminal sliding mode with unidirectional auxiliary surfaces.

  10. Transport in Auxiliary Heated NSTX Discharges

    International Nuclear Information System (INIS)

    LeBlanc, B.P.; Bell, M.G.; Bell, R.E.; Bitte, M.L.; Bourdelle, C.; Gates, D.A.; Kaye, S.M.; Maingi, R.; Menard, J.E.; Mueller, D.; Ono, M.; Paul, S.F.; Redi, M.H.; Roquemore, A.L.; Rosenberg, A.; Sabbagh, S.A.; Stutman, D.; Synakowski, E.J.; Soukhanovskii, V.A.; Wilson, J.R.

    2003-01-01

    The NSTX spherical torus (ST) provides a unique platform to investigate magnetic confinement in auxiliary-heated plasmas at low aspect ratio. Auxiliary power is routinely coupled to ohmically heated plasmas by deuterium neutral-beam injection (NBI) and by high-harmonic fast waves (HHFW) launch. While theory predicts both techniques to preferentially heat electrons, experiment reveals the electron temperature is greater than the ion temperature during HHFW, but the electron temperature is less than the ion temperature during NBI. In the following we present the experimental data and the results of transport analyses

  11. Trojan Nuclear Plant. 1976 Annual operating report

    International Nuclear Information System (INIS)

    1977-01-01

    Net electrical energy generated was 2,102,986 MWH with the generator on line 3144.4 hours. Information is presented concerning operations, tests, corrective maintenance, auxiliary feedwater pump modifications, main generator fault, fuel performance, shutdowns and power outages, radiation exposure of personnel, radioactive and chemical effluents, and environmental surveillance

  12. Summary of ACSL Simulations of the MSRE Auxiliary Charcoal Bed Vacuum System

    Energy Technology Data Exchange (ETDEWEB)

    Damiano, B

    2000-10-26

    The simulation of the Auxiliary Charcoal Bed (ACB) Vacuum System was performed to evaluate the original vacuum system design, detect and identify design deficiencies, investigate the effects of proposed corrections on system performance, and generally aid in refining the system design before construction and mockup testing. The simulation was performed by using the Advanced Continuous Simulation Language (ACSL). The vacuum system design goals are to provide approximately 20 SCFM of both booster gas and purge gas through the system and maintain a flow of approximately 40 SCFM with a velocity of 50 to 75 f/sec at the entrance to the cyclone separator. The model results showed that the original system design was incapable of meeting the system performance goals. Further simulations showed that the following modifications to the original vacuum system design were required to make the system performance acceptable; (1) Remove valve PCV4. (2) Modify the flow controllers FTC3 and FTC4 from the original flow range of 0-17.6 SCFM (0-500 SLM) to 0-35.3 SCFM (0-1000 SLM). (3) Replace the bellows sealed valves SV-1, SV-3A, SV-3B, SV-4A, and SV-4B with less restrictive ball valves. The simulation results saved considerable time and effort by identifying flaws in the original system design. Early identification of these flaws and the use of the simulation model to investigate possible solutions allowed corrective modifications to be made before construction of the mock up test facility.

  13. Modeling and stability analysis for the upper atmosphere research satellite auxiliary array switch component

    Science.gov (United States)

    Wolfgang, R.; Natarajan, T.; Day, J.

    1987-01-01

    A feedback control system, called an auxiliary array switch, was designed to connect or disconnect auxiliary solar panel segments from a spacecraft electrical bus to meet fluctuating demand for power. A simulation of the control system was used to carry out a number of design and analysis tasks that could not economically be performed with a breadboard of the hardware. These tasks included: (1) the diagnosis of a stability problem, (2) identification of parameters to which the performance of the control system was particularly sensitive, (3) verification that the response of the control system to anticipated fluctuations in the electrical load of the spacecraft was satisfactory, and (4) specification of limitations on the frequency and amplitude of the load fluctuations.

  14. Natural circulation in reactor coolant system

    International Nuclear Information System (INIS)

    Han, J.T.

    1987-01-01

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

  15. Report of the independent Ad Hoc Group for the Davis-Besse incident

    International Nuclear Information System (INIS)

    1986-06-01

    The Nuclear Regulatory Commission established an independent Ad Hoc Group in January 1986 to review issues subsequent to a complete loss of feedwater event at Davis-Besse Nuclear Power Station on June 9, 1985, including the NRC Incident Investigation Team (IIT) investigation of that event. The Commission asked the Group to identify additional lessons that might be learned and from these to make recommendations to improve NRC oversight of reactor licensees. To fulfill its charter, the Ad Hoc Group examined the following: (1) pre-event interactions between the licensee and NRC concerning reliability of the auxiliary feedwater system and associated systems; (2) pre-event probabilistic assessments of the reliability of plant safety systems, NRC's review of them, and their use in regulatory decisionmaking; (3) licensee management, operation and maintenance programs as they may have contributed to equipment failures and NRC oversight of such programs; and (4) the mandate, capabilities of members, operation, and results of the NRC Davis-Besse IIT, and the use to which its report was put by the regulatory staff

  16. The English Primary Auxiliary Verbs: A Linguistic Theoretical Exercise

    African Journals Online (AJOL)

    Nekky Umera

    Abstract. Obviously, the fact remains that English Language is a sensitive Language ... Even though the English auxiliary verbs are of two kinds: Primary and Modal auxiliary ..... Therefore we are of the opinion that most speakers lack adequate.

  17. Specific features of auxiliary water supply at underground NPPs

    International Nuclear Information System (INIS)

    Pergamenshchik, B.K.; Pavlov, A.S.

    1991-01-01

    Specific features of auxiliary water supply systems for underground NPPs related to peculiarities of NPP basis equipment arrangement, are considered. Circulation water supply scheme, in which water cooling storage basin (cooling towers) with operational area corresponding to NPP power is on the surface and has traditional design, is proposed. Sufficiently high efficiency of the arrangement proposed is proved

  18. Operational experience on reduction of feedwater iron and liquid radwaste input for Kuosheng Nuclear Power Plant

    International Nuclear Information System (INIS)

    Wen, T.J.; Huang, Theresa Chen; Liu, Wen Tsung; Liu, T.C.; Shyur, Tzu Sheng; Shen, S.C.

    1998-01-01

    Other than cobalt alloys, or low cobalt materials, feedwater iron content plays an important role in crud activation and transport causing the growth of out-of-core radiation fields and associated with radwaste generation. Before installing prefilter in the upstream of condensate deep-bed demineralizer, increasing demand for suspended solid removal required new backwash and regeneration technique in Kuosheng Nuclear Power Plant. At steady state full power operation, the average iron concentration in condensate demineralizer influent was 8-15 ppb. Considering both the necessity of backwash and reduction of liquid radwaste input, several actions had been taken to promote the crud removal capabilities without using ultrasonic resin cleaner and controlled feedwater iron content between 0.5 and 2.0 ppb. This modified resin backwash technique would also generate minimum liquid radwaste. Meanwhile, significant efforts have been made to promote the quality of waste water by carefully control input streams as well as backwash modification to reduce liquid radwaste generation. The daily quantity of liquid radwaste has decreased dramatically in the past two years and is effectively controlled under the expected average daily input of design basis. (author)

  19. Natural convection in an asymmetrically heated vertical channel with an adiabatic auxiliary plate

    International Nuclear Information System (INIS)

    Taieb, Soumaya; Hatem, Laatar Ali; Balti, Jalloul

    2013-01-01

    The effect of an auxiliary plate on natural convection in an asymmetrically heated channel is studied numerically in laminar regime. The computational procedure is made by solving the unsteady two dimensional Navier-Stokes and energy equations. This nonlinear system is integrated by a finite volume approach and then solved in time using the projection method, allowing the decoupling pressure from velocity. More than hundred simulations are performed to determine the best positions of the auxiliary plate that enhance the induced mass flow and the heat transfer rate for modified Rayleigh numbers ranging from Ra m = 10 2 to Ra m = 10 5 . Contour maps are plotted and then used to precise the enhancement rates of the mass flow and the heat transfer for any position of the auxiliary plate in the channel. The numerical results (velocity, pressure and temperature fields) provide detailed information about the evolution of the flow structure according to the geometry considered in this study. In addition, they permit to explain why the mass flow rate and Nusselt number are enhanced for certain positions of the auxiliary plate and are on the contrary deteriorated for others. (authors)

  20. Pressurizer /Auxiliary Spray Piping Stress Analysis For Determination Of Lead Shielding Maximum Allow Able Load

    International Nuclear Information System (INIS)

    Setjo, Renaningsih

    2000-01-01

    Piping stress analysis for PZR/Auxiliary Spray Lines Nuclear Power Plant AV Unit I(PWR Type) has been carried out. The purpose of this analysis is to establish a maximum allowable load that is permitted at the time of need by placing lead shielding on the piping system on class 1 pipe, Pressurizer/Auxiliary Spray Lines (PZR/Aux.) Reactor Coolant Loop 1 and 4 for NPP AV Unit one in the mode 5 and 6 during outage. This analysis is intended to reduce the maximum amount of radiation dose for the operator during ISI ( In service Inspection) period.The result shown that the maximum allowable loads for 4 inches lines for PZR/Auxiliary Spray Lines is 123 lbs/feet

  1. Results of an aging-related failure survey of light water safety systems and components

    International Nuclear Information System (INIS)

    Meale, B.M.; Satterwhite, D.G.; MacDonald, P.E.

    1988-01-01

    The collection and evaluation of operating experience data are necessary in determining the effects of aging on the safety of operating nuclear plants. This paper presents the final results of a two-year research effort evaluating aging impacts on components in light water reactor systems. This research was performed as a part of the Nuclear Plant Aging Research program, sponsored by the US Nuclear Regulatory Commission. Two unique types of data analyses were performed. In the first, an aging-survey study, aging-related failure data for fifteen light water reactor systems were obtained from the Nuclear Plant Reliability Data System (NPRDS). These included safety, support, and power conversion systems. A computerized sort of these records classified each record into one of five generic categories, based on the utility's choice of the failure's NPRDS cause category. Systems and components within the systems that were most affected by aging were identified. In the second analysis, information on aging-related reported causes of failures was evaluated for component failures reported to NPRDS for auxiliary feedwater, high pressure injection, service water, and Class 1E electrical power distribution systems. 3 refs., 13 figs., 4 tabs

  2. Optimized auxiliary representation of non-Markovian impurity problems by a Lindblad equation

    International Nuclear Information System (INIS)

    Dorda, A; Sorantin, M; Linden, W von der; Arrigoni, E

    2017-01-01

    We present a general scheme to address correlated nonequilibrium quantum impurity problems based on a mapping onto an auxiliary open quantum system of small size. The infinite fermionic reservoirs of the original system are thereby replaced by a small number N B of noninteracting auxiliary bath sites whose dynamics are described by a Lindblad equation, which can then be exactly solved by numerical methods such as Lanczos or matrix-product states. The mapping becomes exponentially exact with increasing N B , and is already quite accurate for small N B . Due to the presence of the intermediate bath sites, the overall dynamics acting on the impurity site is non-Markovian. While in previous work we put the focus on the manybody solution of the associated Lindblad problem, here we discuss the mapping scheme itself, which is an essential part of the overall approach. On the one hand, we provide technical details together with an in-depth discussion of the employed algorithms, and on the other hand, we present a detailed convergence study. The latter clearly demonstrates the above-mentioned exponential convergence of the procedure with increasing N B . Furthermore, the influence of temperature and an external bias voltage on the reservoirs is investigated. The knowledge of the particular convergence behavior is of great value to assess the applicability of the scheme to certain physical situations. Moreover, we study different geometries for the auxiliary system. On the one hand, this is of importance for advanced manybody solution techniques such as matrix product states which work well for short-ranged couplings, and on the other hand, it allows us to gain more insights into the underlying mechanisms when mapping non-Markovian reservoirs onto Lindblad-type impurity problems. Finally, we present results for the spectral function of the Anderson impurity model in and out of equilibrium and discuss the accuracy obtained with the different geometries of the auxiliary system

  3. Weyl functions, the inverse problem and special solutions for the system auxiliary to the nonlinear optics equation

    International Nuclear Information System (INIS)

    Sakhnovich, Alexander

    2008-01-01

    A Borg–Marchenko-type uniqueness theorem (in terms of the Weyl function) is obtained here for the system auxiliary to the N-wave equation. A procedure to solve the inverse problem is used for this purpose. The asymptotic condition on the Weyl function, under which the inverse problem is uniquely solvable, is completed by a new and simple sufficient condition on the potential, which implies this asymptotic condition. The evolution of the Weyl function is discussed and the solution of an initial-boundary-value problem for the N-wave equation follows. Explicit solutions of an inverse problem are obtained. The system with a shifted argument is treated

  4. Some comments regarding the use of frantic

    International Nuclear Information System (INIS)

    Poern, K.

    1982-01-01

    The paper describes how the unavailabiblity of a system can be calculated using structure functions of independent subtrees and an organizing function for the entire fault tree. By the definition of an independent subtree the substructures will consist of disjunctive sets of indicator variables. If these variables are assumed to be stochastically independent the expected value operation to calculate the system unavailability can be factorized. In that way the system unavailability can explicitly be expressed in terms of subtree unavailabilities. The method is applied to the auxiliary feedwater system in the Oskarshamn 2 reactor, earlier studied in connection with the FRANTIC code (Studsvik/SD-81/82). (Author)

  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. Parallel Auxiliary Space AMG Solver for $H(div)$ Problems

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, Tzanio V. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Vassilevski, Panayot S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2012-12-18

    We present a family of scalable preconditioners for matrices arising in the discretization of $H(div)$ problems using the lowest order Raviart--Thomas finite elements. Our approach belongs to the class of “auxiliary space''--based methods and requires only the finite element stiffness matrix plus some minimal additional discretization information about the topology and orientation of mesh entities. Also, we provide a detailed algebraic description of the theory, parallel implementation, and different variants of this parallel auxiliary space divergence solver (ADS) and discuss its relations to the Hiptmair--Xu (HX) auxiliary space decomposition of $H(div)$ [SIAM J. Numer. Anal., 45 (2007), pp. 2483--2509] and to the auxiliary space Maxwell solver AMS [J. Comput. Math., 27 (2009), pp. 604--623]. Finally, an extensive set of numerical experiments demonstrates the robustness and scalability of our implementation on large-scale $H(div)$ problems with large jumps in the material coefficients.

  8. Digital control system of a steam generator water level by LQG optimal method

    International Nuclear Information System (INIS)

    Lee, Yoon Joon

    1993-01-01

    A digital control system for the steam generator water level control is developed using LQG optimal design method. To describe the more realistic situaton, a feedwater valve actuator is assumed to be of the first order lagger and is included in the overall control system. By composing the digital control circuit in such a way that the overall control system consists of two sub-systems of feedwater station and feedback loop digital controller, the design procedure is divided into two independent steps. The feedwater station system is described in the error dynamics of an ordinary regulator system. The optimal gains are obtained by LQ method which imposes the constraints of the feedwater valve motion as well as on the output deviations. Developed also is a Kalman observer on account of the flow measurement uncertainty at low power. Then a digital controller on the feedback loop is designed so that the system maintains the same stability margins for all power ranges. The simulation results show thst the optimal digital system has a good control characteristics despite the adverse dynamics of a steam generator at low power. (Author)

  9. Reliability of supply of switchgear for auxiliary low voltage in substations extra high voltage to high voltage

    Directory of Open Access Journals (Sweden)

    Perić Dragoslav M.

    2015-01-01

    Full Text Available Switchgear for auxiliary low voltage in substations (SS of extra high voltages (EHV to high voltage (HV - SS EHV/HV kV/kV is of special interest for the functioning of these important SS, as it provides a supply for system of protection and other vital functions of SS. The article addresses several characteristic examples involving MV lines with varying degrees of independence of their supply, and the possible application of direct transformation EHV/LV through special voltage transformers. Auxiliary sources such as inverters and diesel generators, which have limited power and expensive energy, are also used for the supply of switchgear for auxiliary low voltage. Corresponding reliability indices are calculated for all examples including mean expected annual engagement of diesel generators. The applicability of certain solutions of switchgear for auxiliary low voltage SS EHV/HV, taking into account their reliability, feasibility and cost-effectiveness is analyzed too. In particular, the analysis of applications of direct transformation EHV/LV for supply of switchgear for auxiliary low voltage, for both new and existing SS EHV/HV.

  10. Auxiliary partial liver transplantation

    NARCIS (Netherlands)

    C.B. Reuvers (Cornelis Bastiaan)

    1986-01-01

    textabstractIn this thesis studies on auxiliary partial liver transplantation in the dog and the pig are reported. The motive to perform this study was the fact that patients with acute hepatic failure or end-stage chronic liver disease are often considered to form too great a risk for successful

  11. Design and scope of impact of auxiliary lanes : technical report.

    Science.gov (United States)

    2014-06-01

    For decades, Texas Department of Transportation districts have constructed auxiliary lanes to support interchange : ramp operations and to resolve congestion proximate to freeway entrance and exit ramps. While auxiliary lanes are : built throughout T...

  12. A data acquisition and storage system for the ion auxiliary propulsion system cyclic thruster test

    Science.gov (United States)

    Hamley, John A.

    1989-01-01

    A nine-track tape drive interfaced to a standard personal computer was used to transport data from a remote test site to the NASA Lewis mainframe computer for analysis. The Cyclic Ground Test of the Ion Auxiliary Propulsion System (IAPS), which successfully achieved its goal of 2557 cycles and 7057 hr of thrusting beam on time generated several megabytes of test data over many months of continuous testing. A flight-like controller and power supply were used to control the thruster and acquire data. Thruster data was converted to RS232 format and transmitted to a personal computer, which stored the raw digital data on the nine-track tape. The tape format was such that with minor modifications, mainframe flight data analysis software could be used to analyze the Cyclic Ground Test data. The personal computer also converted the digital data to engineering units and displayed real time thruster parameters. Hardcopy data was printed at a rate dependent on thruster operating conditions. The tape drive provided a convenient means to transport the data to the mainframe for analysis, and avoided a development effort for new data analysis software for the Cyclic test. This paper describes the data system, interfacing and software requirements.

  13. Conceptual design of passive containment cooling system with air holdup tanks of improved APR+

    International Nuclear Information System (INIS)

    Jeon, Byong Guk; Cheon No, Hee

    2014-01-01

    In Korea, after the successful validation of passive auxiliary feedwater system (PAFS), a passive containment cooling system (PCCS) gets attention for future development. We suggested PCCS design based on APR+, an advanced PWR developed in Korea, and performed scoping analysis. On the extension of the simple scoping analysis, MARS simulation is performed to incorporate the behavior of water pool outside the containment as well as steam-air mixture inside the containment. Through the simulation we demonstrated the effectiveness of the air holdup tank (AHT). Also we investigated the effect of the models of heat transfer coefficients between steam-air mixture side and water side, and flow instability inside HX tubes. The presence of AHT enables us to reduce the number of required HX tubes more than half through an increase in the heat transfer coefficients due to the reduction of air fraction in the containment. Finally flow instability was observed and mitigated by putting orifice plates at the inlet of tubes, increasing height of return nozzle, and increasing a tube angle. (authors)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. A moving image system for cardiovascular nuclear medicine. A dedicated auxiliary device for the total capacity imaging system for multiple plane dynamic colour display

    International Nuclear Information System (INIS)

    Iio, M.; Toyama, H.; Murata, H.; Takaoka, S.

    1981-01-01

    The recent device of the authors, the dedicated multiplane dynamic colour image display system for nuclear medicine, is discussed. This new device is a hardware-based auxiliary moving image system (AMIS) attached to the total capacity image processing system of the authors' department. The major purpose of this study is to develop the dedicated device so that cardiovascular nuclear medicine and other dynamic studies will include the ability to assess the real time delicate processing of the colour selection, edge detection, phased analysis, etc. The auxiliary system consists of the interface for image transferring, four IC refresh memories of 64x64 matrix with 10 bit count depth, a digital 20-in colour TV monitor, a control keyboard and a control panel with potentiometers. This system has five major functions for colour display: (1) A microcomputer board can select any one of 40 different colour tables preset in the colour transformation RAM. This key also provides edge detection at a certain level of the count by leaving the optional colour and setting the rest of the levels at 0 (black); (2) The arithmetic processing circuit performs the operation of the fundamental rules, permitting arithmetic processes of the two images; (3) The colour level control circuit is operated independently by four potentiometers for four refresh image memories, so that the gain and offset of the colour level can be manually and visually controlled to the satisfaction of the operator; (4) The simultaneous CRT display of the maximum four images with or without cinematic motion is possible; (5) The real time movie interval is also adjustable by hardware, and certain frames can be freezed with overlapping of the dynamic frames. Since this system of AMIS is linked with the whole capacity image processing system of the CPU size of 128kW, etc., clinical applications are not limited to cardiovascular nuclear medicine. (author)

  16. TRAC-PF1 analysis of LOFT steam-generator feedwater transient test L9-1

    International Nuclear Information System (INIS)

    Meier, J.K.

    1983-01-01

    The Transient Reactor Analysis Code (TRAC-PF1) calculations were compared to test data from Loss-of-Fluid Test (LOFT) L9-1, which was a loss-of-feedwater transient. This paper includes descriptions of the test and the TRAC input and compares the TRAC-calculated results with the test data. We conclude that the code predicted the experiment well, given the uncertainties in the boundary conditions. The analysis indicates the need to model all the flow paths and heat structures, and to improve the TRAC wall condensation heat-transfer model

  17. Device for detecting the water leak within a feedwater nozzle in water cooled reactors

    International Nuclear Information System (INIS)

    Hattori, Tsunekazu.

    1984-01-01

    Purpose: To enable exact recognition and detection for the state of water leak. Constitution: The detection device comprises a thermocouple disposed to the outer surface of a feedwater nozzle, a distortion meter for detecting the change in the outer diameter of a nozzle and an acoustic emission generator disposed to the inside of the nozzle for generating a signal upon temperature change. These sensors previously monitor the states during normal operation, and thus detect the change in each of the states upon occurrence of water leakage to issue an alarm. (Kamimura, M.)

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  19. Exergoeconomic Evaluation of a Modern Ultra-Supercritical Power Plant

    Directory of Open Access Journals (Sweden)

    Lingnan Wu

    2012-09-01

    Full Text Available In this paper, the exergoeconomic analysis was conducted to an existing ultra-supercritical coal-fired power plant in China to understand the cost-formation process, to evaluate the economic performance of each component and to find possible solutions for more cost-effective designs. The total revenue requirement (TRR and the specific exergy costing (SPECO methods were applied for economic analysis and exergy costing, respectively. Quantitative balances of exergy and exergetic costs as well as necessary auxiliary equations for both individual component and the overall system were established. The results show that the exergoeconomic factors of the furnace and heat exchangers at low temperature levels, including air preheater and low-pressure feedwater preheaters, are rather small; while those of other components are relatively large. Moving more heat absorption into furnace to use the effective radiation heat transfer, increasing the air preheating temperature and adding more low pressure feedwater preheaters can be promising solutions for future design.

  20. Simulation of the fault transitory of the feedwater controller in a Boiling water reactor with the Ramona-3B code

    International Nuclear Information System (INIS)

    Hernandez M, J.L.; Ortiz V, J.

    2005-01-01

    The obtained results when carrying out the simulation of the fault transitory of the feedwater controller (FCAA) with the Ramona-3B code, happened in the Unit 2 of the Laguna Verde power plant (CNLV), in September of the year 2000 are presented. The transitory originates as consequence of the controller's fault of speed of a turbo pump of feedwater. The work includes a short description of the event, the suppositions considered for the simulation and the obtained results. Also, a discussion of the impact of the transitory event is presented on aspects of reactor safety. Although the carried out simulation is limited by the capacities of the code and for the lack of available information, it was found that even in a conservative situation, the power was incremented only in 12% above the nominal value, while that the thermal limit determined by the minimum reason of the critical power, MCPR, always stayed above the limit values of operation and safety. (Author)

  1. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  2. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    International Nuclear Information System (INIS)

    Lv, Quiping; Sun, Xiaodong; Chtistensen, Richard; Blue, Thomas; Yoder, Graydon; Wilson, Dane

    2015-01-01

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  3. Biased divertor performance under auxiliary heating conditions on the TdeV tokamak

    International Nuclear Information System (INIS)

    Decoste, R.; Lachambre, J.L.; Demers, Y.

    1994-01-01

    Plasma biasing has been shown on TdeV in the ohmic regime to be very promising for divertor applications. Negative biasing, with shortened SOL density gradients, improves the divertor performance, whereas positive biasing, with longer gradients, does not do much for the divertor. The next objectives were to extrapolate those results to auxiliary heated plasmas and optimize/simplify the biasing geometry for future upgrades. New results are now available with an improved divertor geometry and auxiliary heating/current drive provided by a new lower hybrid (LH) system. The new geometry, optimized for positive biasing with predictably acceptable negative biasing performances, allows for a fair comparison between the two polarities. (author) 4 refs., 5 figs

  4. Eigenstates with the auxiliary field method

    Energy Technology Data Exchange (ETDEWEB)

    Semay, Claude [Service de Physique Nucleaire et Subnucleaire, Universite de Mons-UMONS, 20 Place du Parc, 7000 Mons (Belgium); Silvestre-Brac, Bernard, E-mail: claude.semay@umons.ac.b, E-mail: silvestre@lpsc.in2p3.f [LPSC Universite Joseph Fourier, Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, Avenue des Martyrs 53, F-38026 Grenoble-Cedex (France)

    2010-07-02

    The auxiliary field method is a powerful technique to obtain approximate closed-form energy formulas for eigenequations in quantum mechanics. Very good results can be obtained for Schroedinger and semirelativistic Hamiltonians with various potentials, even in the case of many-body problems. This method can also provide approximate eigenstates in terms of well-known wavefunctions, for instance harmonic oscillator or hydrogen-like states, but with a characteristic size which depends on quantum numbers. In this paper, we consider two-body Schroedinger equations with linear, logarithmic and exponential potentials and show that analytical approximations of the corresponding eigenstates can be obtained with the auxiliary field method, with very good accuracy in some cases.

  5. Eigenstates with the auxiliary field method

    International Nuclear Information System (INIS)

    Semay, Claude; Silvestre-Brac, Bernard

    2010-01-01

    The auxiliary field method is a powerful technique to obtain approximate closed-form energy formulas for eigenequations in quantum mechanics. Very good results can be obtained for Schroedinger and semirelativistic Hamiltonians with various potentials, even in the case of many-body problems. This method can also provide approximate eigenstates in terms of well-known wavefunctions, for instance harmonic oscillator or hydrogen-like states, but with a characteristic size which depends on quantum numbers. In this paper, we consider two-body Schroedinger equations with linear, logarithmic and exponential potentials and show that analytical approximations of the corresponding eigenstates can be obtained with the auxiliary field method, with very good accuracy in some cases.

  6. S-matrix for the theories that admit closure of the algebra with the aid of auxiliary fields. Auxiliary fields in supergravity. [Word identities

    Energy Technology Data Exchange (ETDEWEB)

    Fradkin, E S; Vasiliev, M A [AN SSSR, Moscow. Fizicheskij Inst.

    1978-08-19

    A minimal set of auxiliary fields (scalarpseudoscalar and pseudovector) providing the closed algebra in supergravity is constructed. A compact scheme for the generating functional with closed gauge algebra is proposed. The S-matrix and the Ward identities for arbitrary theory that admits the closing of the algebra by introducing auxiliary fields is obtained.

  7. INFLUENCE OF FEEDING ELECTRIC ENERGY QUALITY ON HEATING OF THE AUXILIARY MA-CHINES OF AC ELECTRIC ROLLING STOCK

    Directory of Open Access Journals (Sweden)

    O. YU. Baliichuk

    2014-04-01

    Full Text Available Purpose. The article aims to study the problem of increase the reliability of auxiliary machines for AC electric trains during operation in real conditions. Methodology. The peculiarity of system construction of auxiliary machines for AC electric rolling stock is the use of asynchronous motors for general industrial purpose. An engineering method of influence determination on the feeding voltage asymmetry and its deviation from the nominal value on heating of auxiliary machines insulation was proposed. Findings. It is found out that in case when the auxiliary machines of AC electric trains work under asymmetry factor of the voltage 10% or more and feeding voltage deviation from the nominal order 0.6 relative unit then it is possible the overheat of their isolation, even if it has class H. Originality. For the first time the issue of the total insulation heating under such boundary parameters combinations of energy quality, when each of them contributes to the heating insulation increase as compared to the nominal regime of the "rotating phase splitter−auxiliary machinery" system was illuminated. Practical value. Conducted research allow us to establish the boundary parameter values of feeding energy quality (asymmetry factor, feeding voltage deviations from the nominal value, at which additional isolation overheating of this class under the effect of specified factors will not exceed the agreed value.

  8. Wigner Transport Simulation of Resonant Tunneling Diodes with Auxiliary Quantum Wells

    Science.gov (United States)

    Lee, Joon-Ho; Shin, Mincheol; Byun, Seok-Joo; Kim, Wangki

    2018-03-01

    Resonant-tunneling diodes (RTDs) with auxiliary quantum wells ( e.g., emitter prewell, subwell, and collector postwell) are studied using a Wigner transport equation (WTE) discretized by a thirdorder upwind differential scheme. A flat-band potential profile is used for the WTE simulation. Our calculations revealed functions of the auxiliary wells as follows: The prewell increases the current density ( J) and the peak voltage ( V p ) while decreasing the peak-to-valley current ratio (PVCR), and the postwell decreases J while increasing the PVCR. The subwell affects J and PVCR, but its main effect is to decrease V p . When multiple auxiliary wells are used, each auxiliary well contributes independently to the transport without producing side effects.

  9. Increasing the mode-locking efficiency of a cw solid-state laser with an auxiliary cavity

    International Nuclear Information System (INIS)

    Kalashnikov, V.L.; Kalosha, V.P.; Mikhailov, V.P.; Demchuk, M.I.

    1992-01-01

    It is predicted theoretically that the efficiency of self-mode locking can be raised by means of a bleachable shutter in the main cavity or an auxiliary cavity. The laser emits a stable train of ultrashort pulses under these conditions. The theory is based on a fluctuation model of the operation of a cw solid-state laser with a linear auxiliary cavity. The increase in efficiency involves a broadening of the region of parameter values of the system in which self-mode locking occurs, a significant decrease in the threshold pump intensity, and a reduced sensitivity of the operation to the phase mismatch of the lengths of the cavities. It is shown, for the first time, that a stable train of double ultrashort pulses can be generated by a system with a shutter in the auxiliary cavity. It is also shown that a self-mode locking is possible in the case in which there is a phase mismatch of the cavity lengths and there is no phase self-modulation in the main cavity. 15 refs., 8 figs

  10. Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely.

  11. Analysis of Multiple Spurious Operation Scenarios for Decay Heat Removal Function of CANDU Reactors

    International Nuclear Information System (INIS)

    Lee, Youngseung; Bae, Yeon-kyoung; Kim, Myungsu

    2016-01-01

    The worst fire broke out in the Browns Ferry Nuclear Power Plant on March 22, 1975. A fire occurrence in a nuclear power plant has recognized a latently serious incident. Nuclear power plants should achieve and maintain the safe shutdown conditions during and after the occurrence of a fire. Functions of the safe shutdown are five such as the shutdown function, the decay heat removal function, the containment function, monitoring and control function, and the supporting function for CANDU type reactors. The purpose of this paper is to analyze that the decay heat removal function of the safe shutdown functions for CANDU type reactors is achieved under the fire induced multiple spurious operation. The scenarios of the fire induced multiple spurious operations (MSO) for the systems used for the decay heat cooling were analyzed. Additionally, Integrated Severe Accident Analysis code for CANDU plants (ISAAC) for determining success criteria of thermal hydraulic analysis was used. Decay heat cooling systems of CANDU reactors are the auxiliary feedwater system, the emergency water supply system, and the shutdown cooling system. A big fire can threat the safety of nuclear power plants, and safe shutdown conditions. The regulatory body in Korea requires the fire hazard analysis including fire induced MSOs. The safe shutdown functions for CANDU reactors are the shutdown function, the decay heat removal function, the containment function, the monitoring and control function, and the supporting service function. The number of spurious operations for the auxiliary feedwater system is more than six and that for the emergency water supply system is one. Additionally, misoperations for the shutdown cooling system are more than two. Accordingly, if total nine components could be spuriously operated, the decay heat removal function would be lost entirely

  12. VLTI auxiliary telescopes: a full object-oriented approach

    Science.gov (United States)

    Chiozzi, Gianluca; Duhoux, Philippe; Karban, Robert

    2000-06-01

    The Very Large Telescope (VLT) Telescope Control Software (TCS) is a portable system. It is now in use or will be used in a whole family of ESO telescopes VLT Unit Telescopes, VLTI Auxiliary Telescopes, NTT, La Silla 3.6, VLT Survey Telescope and Astronomical Site Monitors in Paranal and La Silla). Although it has been developed making extensive usage of Object Oriented (OO) methodologies, the overall development process chosen at the beginning of the project used traditional methods. In order to warranty a longer lifetime to the system (improving documentation and maintainability) and to prepare for future projects, we have introduced a full OO process. We have taken as a basis the United Software Development Process with the Unified Modeling Language (UML) and we have adapted the process to our specific needs. This paper describes how the process has been applied to the VLTI Auxiliary Telescopes Control Software (ATCS). The ATCS is based on the portable VLT TCS, but some subsystems are new or have specific characteristics. The complete process has been applied to the new subsystems, while reused code has been integrated in the UML models. We have used the ATCS on one side to tune the process and train the team members and on the other side to provide a UML and WWW based documentation for the portable VLT TCS.

  13. The Range of Gapping and the Status of Auxiliaries.

    Science.gov (United States)

    Warner, A. R.

    Full verbs and auxiliaries are subject to gapping. In the simplest cases, this construction type involves apparent ellipsis within one or more clausal conjuncts under identity with the finite verb or auxiliary of a preceding conjunct. It has often been suggested that the apparent ellipsis must involve at least a verb. Some researchers see in the…

  14. Auxiliary office chair

    OpenAIRE

    Pascual Osés, Maite

    2007-01-01

    The aim of this project is to develop an auxiliary office chair, which favorably will compete with the existing chairs on the market. Evolutions of ergonomical survey in the work environment and on the configuration of offices require new products which fulfill the requirements properly. In order to achieve it a survey about office chairs has been carried out: types, characteristics, ways of usage and products on the market besides a large antropometrical study and ergonomics related to work ...

  15. Generating Selected Color using RGB, Auxiliary Lights, and Simplex Search

    Directory of Open Access Journals (Sweden)

    Kim HyungTae

    2015-01-01

    Full Text Available A mixed light source generates various colors, with the potential to adjust intensities of multiple LEDs, which makes it possible to generate arbitrary colors. Currently, PCs and OSs provide color selection windows that can obtain the RGB or HSL color coordinates of a user’s selection. Mixed light sources are usually composed of LEDs in the primary colors, with LEDs in auxiliary colors such as white and yellow used in a few cases. When using auxiliary color LEDs, the number of LED inputs, the dimming levels, is larger than the number of elements in the color coordinate, which causes an under-determined problem. This study proposed how to determine the dimming levels of LEDs based on the selected color. Commercial LEDs have di_erent optical power values and impure color coordinates, even if they are RGB. Hence, the characteristics of the LEDs were described using a linear model derived from the tri-stimulus values (an XYZ color coordinate model and dimming levels. Color mixing models were derived for the arbitrary number of auxiliary color LEDs. The under-determined problem was solved using a simplex search method without an inverse matrix operation. The proposed method can be applied to a machine vision system and an RGBW light mixer for semiconductor inspection. The dimming levels, obtained using the proposed method were better than derived using other methods.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  17. Development of the nuclear plant analyzer for Korean standard Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Shin Hwan; Kim, Hyeong Heon; Song, In Ho; Hong, Eon Yeong; Oh, Yeong Taek [Korea Power Engineering Company Inc., Yongin (Korea, Republic of)

    2000-12-15

    The purpose of this study is to develop an NPA for the Ulchin Nuclear Power Plant Unit 3 and 4, the first KSNP type plant. In this study, the process model simulating the overall plant systems, GUI and simulation executive which provide the functions of an engineering simulator were developed, and the NPA was completed by integrating them. The contents and the scope of this study are as follows : main feedwater system, auxiliary feedwater system, Chemical and Volume Control System(CVCS), Safety Injection System(SIS), Shutdown Cooling System(SCS), electric power supply system, Core Protection Calculator(CPC), various plant control system, development of the graphics screens for each system, real-time simulation, simulation control for the enhancement of functional capabilities, user friendly GUI, collection of the design and operating data, establishment of the NPA database, integration of the GUI and simulation control program with process model, collection of the data for the verification and validation of the developed NPA, collection of the plant test data, collection and review of the results of other computer codes, verification of the simulation accuracy by comparing the NPA results with the actual plant data, validation of the simulation capability of the NPA, comparison against available data from other analysis suing different computer codes.

  18. The Quantum N-Body Problem and the Auxiliary Field Method

    International Nuclear Information System (INIS)

    Semay, C.; Buisseret, F.; Silvestre-Brac, B.

    2011-01-01

    Approximate analytical energy formulas for N-body semirelativistic Hamiltonians with one- and two-body interactions are obtained within the framework of the auxiliary field method. We first review the method in the case of nonrelativistic two-body problems. A general procedure is then given for N-body systems and applied to the case of baryons in the large-N c limit. (author)

  19. PEMFC Optimization Strategy with Auxiliary Power Source in Fuel Cell Hybrid Vehicle

    Directory of Open Access Journals (Sweden)

    Tinton Dwi Atmaja

    2012-02-01

    Full Text Available Page HeaderOpen Journal SystemsJournal HelpUser You are logged in as...aulia My Journals My Profile Log Out Log Out as UserNotifications View (27 new ManageJournal Content SearchBrowse By Issue By Author By Title Other JournalsFont SizeMake font size smaller Make font size default Make font size largerInformation For Readers For Authors For LibrariansKeywords CBPNN Displacement FLC LQG/LTR Mixed PMA Ventilation bottom shear stress direct multiple shooting effective fuzzy logic geoelectrical method hourly irregular wave missile trajectory panoramic image predator-prey systems seawater intrusion segmentation structure development pattern terminal bunt manoeuvre Home About User Home Search Current Archives ##Editorial Board##Home > Vol 23, No 1 (2012 > AtmajaPEMFC Optimization Strategy with Auxiliary Power Source in Fuel Cell Hybrid VehicleTinton Dwi Atmaja, Amin AminAbstractone of the present-day implementation of fuel cell is acting as main power source in Fuel Cell Hybrid Vehicle (FCHV. This paper proposes some strategies to optimize the performance of Polymer Electrolyte Membrane Fuel Cell (PEMFC implanted with auxiliary power source to construct a proper FCHV hybridization. The strategies consist of the most updated optimization method determined from three point of view i.e. Energy Storage System (ESS, hybridization topology and control system analysis. The goal of these strategies is to achieve an optimum hybridization with long lifetime, low cost, high efficiency, and hydrogen consumption rate improvement. The energy storage system strategy considers battery, supercapacitor, and high-speed flywheel as the most promising alternative auxiliary power source. The hybridization topology strategy analyzes the using of multiple storage devices injected with electronic components to bear a higher fuel economy and cost saving. The control system strategy employs nonlinear control system to optimize the ripple factor of the voltage and the current

  20. A new approach for AC loss reduction in HTS transformer using auxiliary windings, case study: 25 kA HTS current injection transformer

    Science.gov (United States)

    Heydari, Hossein; Faghihi, Faramarz; Aligholizadeh, Reza

    2008-01-01

    AC loss is one of the important parameters in HTS (high temperature superconducting) AC devices. Among the HTS AC power devices, the transformer is an essential part in the electrical power system. The AC losses in an HTS tape depend on the magnetic field. One of the techniques usually adopted to mitigate the unwanted magnetic field is using a system of coils that produce a magnetic field opposite to the incident one, reducing the total magnetic field. In this paper adding two auxiliary windings to the HTS transformer to produce this opposite magnetic field is proposed. The proper use of these auxiliary windings could reduce the leakage flux and, therefore, the AC loss. A mathematical model is used to describe the behaviour of a transformer operating with auxiliary windings, based on the theory of electromagnetic coupled circuits. The influence of the auxiliary windings on the leakage field is studied by the finite element method (FEM) and the AC loss of an HTS transformer was calculated. Also, the simulation results show that employing auxiliary windings will improve the HTS transformer efficiency.

  1. A new approach for AC loss reduction in HTS transformer using auxiliary windings, case study: 25 kA HTS current injection transformer

    International Nuclear Information System (INIS)

    Heydari, Hossein; Faghihi, Faramarz; Aligholizadeh, Reza

    2008-01-01

    AC loss is one of the important parameters in HTS (high temperature superconducting) AC devices. Among the HTS AC power devices, the transformer is an essential part in the electrical power system. The AC losses in an HTS tape depend on the magnetic field. One of the techniques usually adopted to mitigate the unwanted magnetic field is using a system of coils that produce a magnetic field opposite to the incident one, reducing the total magnetic field. In this paper adding two auxiliary windings to the HTS transformer to produce this opposite magnetic field is proposed. The proper use of these auxiliary windings could reduce the leakage flux and, therefore, the AC loss. A mathematical model is used to describe the behaviour of a transformer operating with auxiliary windings, based on the theory of electromagnetic coupled circuits. The influence of the auxiliary windings on the leakage field is studied by the finite element method (FEM) and the AC loss of an HTS transformer was calculated. Also, the simulation results show that employing auxiliary windings will improve the HTS transformer efficiency

  2. A new approach for AC loss reduction in HTS transformer using auxiliary windings, case study: 25 kA HTS current injection transformer

    Energy Technology Data Exchange (ETDEWEB)

    Heydari, Hossein; Faghihi, Faramarz; Aligholizadeh, Reza [Center of Excellence for Power System Automation and Operation, Electrical Engineering Department, Iran University of Science and Technology, Tehran (Iran, Islamic Republic of)

    2008-01-15

    AC loss is one of the important parameters in HTS (high temperature superconducting) AC devices. Among the HTS AC power devices, the transformer is an essential part in the electrical power system. The AC losses in an HTS tape depend on the magnetic field. One of the techniques usually adopted to mitigate the unwanted magnetic field is using a system of coils that produce a magnetic field opposite to the incident one, reducing the total magnetic field. In this paper adding two auxiliary windings to the HTS transformer to produce this opposite magnetic field is proposed. The proper use of these auxiliary windings could reduce the leakage flux and, therefore, the AC loss. A mathematical model is used to describe the behaviour of a transformer operating with auxiliary windings, based on the theory of electromagnetic coupled circuits. The influence of the auxiliary windings on the leakage field is studied by the finite element method (FEM) and the AC loss of an HTS transformer was calculated. Also, the simulation results show that employing auxiliary windings will improve the HTS transformer efficiency.

  3. The relationship among the solutions of two auxiliary ordinary differential equations

    International Nuclear Information System (INIS)

    Liu Xiaoping; Liu Chunping

    2009-01-01

    In a recent article [Phys. Lett. A 356 (2006) 124], Sirendaoreji extended their auxiliary equation method by introducing a new auxiliary ordinary differential equation (NAODE) and its 14 solutions. Then the author studied some nonlinear evolution equations (NLEEs) and got more exact travelling wave solutions. In this paper, we will show that the 14 solutions of the NAODE are actually the same as the solutions obtained by original auxiliary equation method, and they are only different in the form.

  4. Simplified technique for auxiliary orthotopic liver transplantation using a whole graft

    Science.gov (United States)

    ROCHA-SANTOS, Vinicius; NACIF, Lucas Souto; PINHEIRO, Rafael Soares; DUCATTI, Liliana; ANDRAUS, Wellington; D'ALBURQUERQUE, Luiz Carneiro

    2015-01-01

    Background Acute liver failure is associated with a high mortality rate and the main purposes of treatment are to prevent cerebral edema and infections, which often are responsible for patient death. The orthotopic liver transplantation is the gold standard treatment and improves the 1-year survival. Aim To describe an alternative technique to auxiliary liver transplant on acute liver failure. Method Was performed whole auxiliary liver transplantation as an alternative technique for a partial auxiliary liver transplantation using a whole liver graft from a child removing the native right liver performed a right hepatectomy. The patient met the O´Grady´s criteria and the rational to indicate an auxiliary orthotopic liver transplantation was the acute classification without hemodynamic instability or renal failure in a patient with deterioration in consciousness. Results The procedure improved liver function and decreased intracranial hypertension in the postoperative period. Conclusion This technique can overcome some postoperative complications that are associated with partial grafts. As far as is known, this is the first case of auxiliary orthotopic liver transplantation in Brazil. PMID:26176253

  5. Methodology for testing a system for remote monitoring and control on auxiliary machines in electric vehicles

    Directory of Open Access Journals (Sweden)

    Dimitrov Vasil

    2017-01-01

    Full Text Available A laboratory system for remote monitoring and control of an asynchronous motor controlled by a soft starter and contemporary measuring and control devices has been developed and built. This laboratory system is used for research and in teaching. A study of the principles of operation, setting up and examination of intelligent energy meters, soft starters and PLC has been made as knowledge of the relevant software products is necessary. This is of great importance because systems for remote monitoring and control of energy consumption, efficiency and proper operation of the controlled objects are very often used in different spheres of industry, in building automation, transport, electricity distribution network, etc. Their implementation in electric vehicles for remote monitoring and control on auxiliary machines is also possible and very useful. In this paper, a methodology of tests is developed and some experiments are presented. Thus, an experimental verification of the developed methodology is made.

  6. On the prohibition of automatic redundant power supply of 6 and 0.4 kV auxiliary sections at Leningrad NPP

    International Nuclear Information System (INIS)

    Mokeev, S.F.

    1987-01-01

    At Leningrad NPP the automatic switching on of the auxiliary power supply sources of 6 and 0.4 kV is prohibited to provide personnel safety and preserve destruction of electroequipment. With excess of current maximum value in the 6 kV section immediate detachment of all electric motors occurs. At that moment by the switchers detechment emergency signal it occurs immediate swithing on of auxiliary systems or emergency switching on of auxiliary supply of the main circulating pumps, that insreases abruptly operation reliability of plant technological department

  7. Auxiliary VHF transmitter to aid recovery of solar Argos/GPS PTTs

    Science.gov (United States)

    Christopher P. Hansen; Mark A. Rumble; R. Scott Gamo; Joshua J. Millspaugh

    2014-01-01

    While conducting greater sage-grouse (Centrocercus urophasianus) research, we found that solar-powered global positioning systems platform transmitter terminals (GPS PTTs) can be lost if the solar panel does not receive adequate sunlight. Thus, we developed 5-g (mortality sensor included; Prototype A) and 9.8-g (no mortality sensor; Prototype B) auxiliary very high...

  8. Evaluation of the PRHRS Performance Degradation due to Non-Condensable Gas for the Small and Medium Reactor using MARS-KS code

    International Nuclear Information System (INIS)

    Kim, Sook Kwan; Sim, Suk Ku; Park, Ju Yeop; Seol, Kwang Won; Ryu, Yong Ho

    2011-01-01

    The effect of non-condensable gas on the performance of PRHRS (Passive Residual Heat Removal System) of the Small and Medium Reactor(SMR) was evaluated during a loss of flow event. Since the TMI accident in 1979, the passive systems have been considered in the advanced reactors as a feature of design improvement because the passive system simplifies the system and thus increases the reliability of the system. The Westinghouse received the design certification from the USNRC for the AP600 and AP1000 passive type pressurized water reactors. The APR+ under development by KEPCO considers the use of PAFS (Passive Auxiliary Feedwater System). And the PRHRS is adopted as a passive secondary heat removal system for the SMART (System-integrated Modular Advanced ReacTor)

  9. ISP 22 OECD/NEA/CSNI International standard problem n. 22. Evaluation of post-test analyses

    International Nuclear Information System (INIS)

    1992-07-01

    The present report deals with the open re-evaluation of the originally double-blind CSNI International Standard Problem 22 based on the test SP-FW-02 performed in the SPES facility. The SPES apparatus is an experimental simulator of the Westinghouse PWR-PUN plant. The test SP-FW-02 (ISP22) simulates a complete loss of feedwater with delayed injection of auxiliary feedwater. The main parts of the report are: outline of the test facility and of the SP-FW-02 experiment; overview of pre-test activities; overview of input models used by post-test participants; evaluation of participant predictions; evaluation of qualitative and quantitative code accuracy of pre-test and post-test calculations

  10. Auxiliary Electrodes for Chromium Vapor Sensors

    Energy Technology Data Exchange (ETDEWEB)

    Fergus, Jeffrey; Shahzad, Moaiz; Britt, Tommy

    2018-05-15

    Measurement of chromia-containing vapors in solid oxide fuel cell systems is useful for monitoring and addressing cell degradation caused by oxidation of the chomia scale formed on alloys for interconnects and balance-of-plant components. One approach to measuring chromium is to use a solid electrolyte with an auxiliary electrode that relates the partial pressure of the chromium containing species to the mobile species in the electrolyte. One example is YCrO3 which can equilibrate with the chromium containing vapor and yttrium in yttria stabilized zirconia to establish an oxygen activity. Another is Na2CrO4 which can equilibrate with the chromium-containing vapor to establish a sodium activity.

  11. Auxiliary aggregates: Effects on fuel consumption and potential fuel savings; Einfluss und Potenzial von Nebenaggregaten auf den Kraftstoffverbrauch

    Energy Technology Data Exchange (ETDEWEB)

    Rauchfuss, Lutz; Hindorf, Kay [Robert Bosch GmbH (Germany)

    2008-07-01

    New concepts and new developments in the field of auxiliary aggregates will help to reduce fuel consumption in the future. One aspect is the higher efficiency of components, the other on-demand power supply management. The contribution shows that thermomanagement, intelligent generator control, the start/stop system, auxiliary power control and fuel pumps may achieve considerable savings in the driving cycle and also in real operation. Efficient air conditioning systems and higher generator efficiency will show their potential in real operation. Cost-profit analyses are presented to prove the economic effects for the customer. (orig.)

  12. Exact fluctuations of nonequilibrium steady states from approximate auxiliary dynamics

    OpenAIRE

    Ray, Ushnish; Chan, Garnet Kin-Lic; Limmer, David T.

    2017-01-01

    We describe a framework to significantly reduce the computational effort to evaluate large deviation functions of time integrated observables within nonequilibrium steady states. We do this by incorporating an auxiliary dynamics into trajectory based Monte Carlo calculations, through a transformation of the system's propagator using an approximate guiding function. This procedure importance samples the trajectories that most contribute to the large deviation function, mitigating the exponenti...

  13. Goal programming for cyclical auxiliary police scheduling at UiTM Cawangan Perlis

    Science.gov (United States)

    Mustapar, Wasilatul Effah; Nasir, Diana Sirmayunie Mohd; Nor, Nor Azriani Mohamad; Abas, Sharifah Fhahriyah Syed

    2017-11-01

    Constructing a good and fair schedule for shift workers poses a great challenge as it requires a lot of time and effort. In this study, goal programming has been applied on scheduling to achieve the hard and soft constraints for a cyclical schedule that would ease the head of auxiliary police at building new schedules. To accomplish this goal, shift types were assigned in order to provide a fair schedule that takes into account the auxiliary police's policies and preferences. The model was run using Lingo software. Three out of four goals set for the study were achieved. In addition, the results considered an equal allocation for every auxiliary police, namely 70% for total duty and 30% for the day. Furthermore, the schedule was able to cyclically generate another 10 sets schedule. More importantly, the model has provided unbiased scheduling of auxiliary policies which led to high satisfaction in auxiliary police management.

  14. Minimal set of auxiliary fields and S-matrix for extended supergravity

    Energy Technology Data Exchange (ETDEWEB)

    Fradkin, E S; Vasiliev, M A [Physical Lebedev Institute - Moscow

    1979-05-19

    Minimal set of auxiliary fields for linearized SO(2) supergravity and one-parameter extension of the minimal auxiliary fields in the SO(1) supergravity are constructed. The expression for the S-matrix in SO(2) supergravity are given.

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

    International Nuclear Information System (INIS)

    Hsieh Minchih; Chiu Mingchuan; Hwang Sheueling

    2014-01-01

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

  16. BE, DO, and Modal Auxiliaries of 3-Year-Old African American English Speakers

    Science.gov (United States)

    Newkirk-Turner, Brandi L.; Oetting, Janna B.; Stockman, Ida J.

    2014-01-01

    Purpose: This study examined African American English--speaking children's use of BE, DO, and modal auxiliaries. Method: The data were based on language samples obtained from 48 three-year-olds. Analyses examined rates of marking by auxiliary type, auxiliary surface form, succeeding element, and syntactic construction and by a number of child…

  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. Experimental Results of A1.2 Test for OECD-ATLAS Project

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung-Ho; Bae, Byoung-Uhn; Park, Yu-Sun; Kim, Jong-Rok; Choi, Nam-Hyun; Choi, Ki-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In order to meet the international interests in the multiple high-risk design extension conditions (DECs) raised after the Fukushima accident, KAERI (Korea Atomic Energy Research Institute) is operating an OECD/NEA project (hereafter, OECD-ATLAS project) by utilizing a thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). As for a prolonged SBO transient of the OECD-ATLAS project, two tests, named A1.1 and A1.2, were determined to be performed. In particular, passive safety systems are considered as the most promising alternatives to reinforce the safety and reliability of an ultimate heat removal system without any operator actions in the SBO transients. As one of the new safety improvement concepts to mitigate an SBO accident efficiently, a cooling and operational performance of the passive auxiliary feedwater system (PAFS) is investigated in the framework of the OECD-ATLAS project to produce clearer knowledge of the actual phenomena and to provide the best guidelines for accident management. As the second test of the OECD-ATLAS project, the A1.2 test was conducted to simulate a prolonged SBO with asymmetric secondary cooling through the supply of passive auxiliary feedwater only to SG-2. When the collapsed water level of steam generator reached a wide range of 25%, PAFS was actuated. PAFS played a key role in cooling down the primary system by the heat transfer and the natural circulation. With the actuation of PAFS, the fluid temperatures at the core inlet and outlet started to decrease without any excursion of the maximum heater surface temperature in the core. This integral effect test data of A1.2 test can be used to evaluate the prediction capability of existing safety analysis codes and identify any code deficiency for an SBO simulation with an operation of a passive system such as PAFS.

  19. LOX/LH2 vane pump for auxiliary propulsion systems

    Science.gov (United States)

    Hemminger, J. A.; Ulbricht, T. E.

    1985-01-01

    Positive displacement pumps offer potential efficiency advantages over centrifugal pumps for future low thrust space missions. Low flow rate applications, such as space station auxiliary propulsion or dedicated low thrust orbiter transfer vehicles, are typical of missions where low flow and high head rise challenge centrifugal pumps. The positive displacement vane pump for pumping of LOX and LH2 is investigated. This effort has included: (1) a testing program in which pump performance was investigated for differing pump clearances and for differing pump materials while pumping LN2, LOX, and LH2; and (2) an analysis effort, in which a comprehensive pump performance analysis computer code was developed and exercised. An overview of the theoretical framework of the performance analysis computer code is presented, along with a summary of analysis results. Experimental results are presented for pump operating in liquid nitrogen. Included are data on the effects on pump performance of pump clearance, speed, and pressure rise. Pump suction performance is also presented.

  20. Evaluation of the performance of the systems cooling auxiliary of the Almaraz NPP turbine building; Evaluacion del rendimiento de los sistemas de refrigeracion auxiliar del edificio de turbinas de CNA

    Energy Technology Data Exchange (ETDEWEB)

    Vilar Carmona, G.

    2011-07-01

    After the successive performance improvements, to evacuate the thermal loads, and maintain acceptable temperatures throughout the year, added an auxiliary cooling system of the building of turbine, TCA system, composed of 5 shot forced by unit cooling towers.

  1. The SOL-2/Neto auxiliary protein modulates the function of AMPA-subtype ionotropic glutamate receptors.

    Science.gov (United States)

    Wang, Rui; Mellem, Jerry E; Jensen, Michael; Brockie, Penelope J; Walker, Craig S; Hoerndli, Frédéric J; Hauth, Linda; Madsen, David M; Maricq, Andres V

    2012-09-06

    The neurotransmitter glutamate mediates excitatory synaptic transmission by gating ionotropic glutamate receptors (iGluRs). AMPA receptors (AMPARs), a subtype of iGluR, are strongly implicated in synaptic plasticity, learning, and memory. We previously discovered two classes of AMPAR auxiliary proteins in C. elegans that modify receptor kinetics and thus change synaptic transmission. Here, we have identified another auxiliary protein, SOL-2, a CUB-domain protein that associates with both the related auxiliary subunit SOL-1 and with the GLR-1 AMPAR. In sol-2 mutants, behaviors dependent on glutamatergic transmission are disrupted, GLR-1-mediated currents are diminished, and GLR-1 desensitization and pharmacology are modified. Remarkably, a secreted variant of SOL-1 delivered in trans can rescue sol-1 mutants, and this rescue depends on in cis expression of SOL-2. Finally, we demonstrate that SOL-1 and SOL-2 have an ongoing role in the adult nervous system to control AMPAR-mediated currents. Copyright © 2012 Elsevier Inc. All rights reserved.

  2. Modified Darboux transformations with foreign auxiliary equations

    International Nuclear Information System (INIS)

    Schulze-Halberg, Axel

    2011-01-01

    We construct a new type of first-order Darboux transformations for the stationary Schroedinger equation. In contrast to the conventional case, our Darboux transformations support arbitrary (foreign) auxiliary equations. We show that among other applications, our formalism can be used to systematically construct Darboux transformations for Schroedinger equations with energy-dependent potentials, including a recent result (Lin et al., 2007) as a special case. -- Highlights: → We generalize the Darboux transformation for the Schroedinger equation. → By admitting arbitrary auxiliary functions, we provide a new tool for generating solutions. → As a special case we recover a recent result on energy-dependent potentials. → We extend the latter result to very general energy-dependence.

  3. Hybrid mesons with auxiliary fields

    International Nuclear Information System (INIS)

    Buisseret, F.; Mathieu, V.

    2006-01-01

    Hybrid mesons are exotic mesons in which the color field is not in the ground state. Their understanding deserves interest from a theoretical point of view, because it is intimately related to nonperturbative aspects of QCD. Moreover, it seems that some recently detected particles, such as the π 1 (1600) and the Y(4260), are serious hybrid candidates. In this work, we investigate the description of such exotic hadrons by applying the auxiliary fields technique (also known as the einbein field method) to the widely used spinless Salpeter Hamiltonian with appropriate linear confinement. Instead of the usual numerical resolution, this technique allows to find simplified analytical mass spectra and wave functions of the Hamiltonian, which still lead to reliable qualitative predictions. We analyse and compare two different descriptions of hybrid mesons, namely a two-body q system with an excited flux tube, or a three-body qg system. We also compute the masses of the 1 -+ hybrids. Our results are shown to be in satisfactory agreement with lattice QCD and other effective models. (orig.)

  4. D and D of a plutonium research laboratory and related auxiliary systems

    International Nuclear Information System (INIS)

    Diaz Arocas, P.; Martinez Ortega, A.; Sama Colao, J.; Garcia Diaz, A.; Torre Rodriguez, J.; Diaz Diaz, J.L.; Argiles, E.; Garrido, C.

    2010-01-01

    CIEMAT, former Junta de Energia Nuclear (JEN) started nuclear research at the 60. decade, focussed on the development of pacific uses of Nuclear Energy. At that time, CIEMAT research and pilot plants developed involved the whole nuclear fuel cycle steps. It means from the uranium recovery to the spent fuel reprocessing. With this scope a plutonium research laboratory was constructed and operated from 1961 to the 90's focussed on chemistry of plutonium studies, separation processes and radiochemical analyses, in order to assist the working pilot plants at the Centre. Thereafter, as the result of the changes on the research objectives of CIEMAT, the plutonium laboratory suffered several modifications and finally it was safety stopped due to the obsolescence of its equipments and auxiliary systems. Present paper shows the D and D activities performed and techniques developed to avoid alpha emitter contamination. In every dismantling phase there were established the measures of operational radiological protection adapted to the radiological risk. Dosimetric controls realized during dismantlement showed that incorporation of radionuclides was not detected. Radiological final control was performed applying the derived levels of declassification to request the installation decommissioning. (authors)

  5. Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

    International Nuclear Information System (INIS)

    Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

    2012-01-01

    KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described

  6. Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

    Energy Technology Data Exchange (ETDEWEB)

    Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described.

  7. MARS-KS Code Assessment for Condensation Heat Transfer in Horizontal Tube with the Presence of Non-Condensable Gas using Purdue Experiment

    International Nuclear Information System (INIS)

    Jeon, Seong Su; Lee, Byung Chul; Park, Ju Yeop; Seul, Kwang Won

    2011-01-01

    In South Korea, advanced power reactor plus (APR+), as a Korean specific reactor, is currently under development for the export strategy. In order to raise competitiveness of the APR+ in the world market, it is necessary to develop the original technology for the improved technology, economics, and safety features. For this purpose, a passive auxiliary feedwater system (PAFS) was adopted as an improved safety design concept of APR+: and then there have been many efforts to develop the PAFS. According to PAFS design concept, PAFS can completely replace the auxiliary feedwater system. When the design basis accident, in which feedwater is unavailable, occurs, the PAFS can remove the residual heat in the core and then prevent the core damage. In the PAFS with the horizontal type heat exchanger, two-phase natural circulation, condensation heat transfer in tube, boiling heat transfer in pool, natural convection in pool, etc. are considered as very important thermalhydraulic phenomena (see Fig. 1). Compared with the vertical heat exchanger from these phenomena, the major difference of the horizontal heat exchanger is the condensation heat transfer phenomena in the tube side. There have been many efforts to understand the condensation heat transfer with in the presence of NC gas in tube but most researches focused on the condensation heat transfer in vertical tube. Therefore the details of the condensation heat transfer in the presence of NC gas in horizontal condenser tubes are not well understood. In order to develop the safety evaluation system for APR+ PAFS, it is required to evaluate the capability and applicability of the MARS-KS code for modeling the condensation heat transfer in the horizontal tube with NC gas because many heat transfer correlations in MARS-KS are known to have much uncertainty. In particular, there is no reliable model for the condensation phenomena in horizontal tube with NC gas. In order to assess the MARS-KS code results and identify the

  8. Dummy auxiliaries in child and adult second language acquisition of Dutch

    NARCIS (Netherlands)

    Blom, W.B.T.; de Korte, S.

    2011-01-01

    In previous research it has been observed that second language (L2) learners of Dutch and German use analytic verbs in contexts where the target language requires synthetic verbs. These analytic verbs consist of a semantically empty auxiliary (dummy auxiliary) that selects a lexical infinitive. In

  9. Extensions of the auxiliary field method to solve Schroedinger equations

    International Nuclear Information System (INIS)

    Silvestre-Brac, Bernard; Semay, Claude; Buisseret, Fabien

    2008-01-01

    It has recently been shown that the auxiliary field method is an interesting tool to compute approximate analytical solutions of the Schroedinger equation. This technique can generate the spectrum associated with an arbitrary potential V(r) starting from the analytically known spectrum of a particular potential P(r). In the present work, general important properties of the auxiliary field method are proved, such as scaling laws and independence of the results on the choice of P(r). The method is extended in order to find accurate analytical energy formulae for radial potentials of the form aP(r) + V(r), and several explicit examples are studied. Connections existing between the perturbation theory and the auxiliary field method are also discussed

  10. Extensions of the auxiliary field method to solve Schroedinger equations

    Energy Technology Data Exchange (ETDEWEB)

    Silvestre-Brac, Bernard [LPSC Universite Joseph Fourier, Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, Avenue des Martyrs 53, F-38026 Grenoble-Cedex (France); Semay, Claude; Buisseret, Fabien [Groupe de Physique Nucleaire Theorique, Universite de Mons-Hainaut, Academie universitaire Wallonie-Bruxelles, Place du Parc 20, B-7000 Mons (Belgium)], E-mail: silvestre@lpsc.in2p3.fr, E-mail: claude.semay@umh.ac.be, E-mail: fabien.buisseret@umh.ac.be

    2008-10-24

    It has recently been shown that the auxiliary field method is an interesting tool to compute approximate analytical solutions of the Schroedinger equation. This technique can generate the spectrum associated with an arbitrary potential V(r) starting from the analytically known spectrum of a particular potential P(r). In the present work, general important properties of the auxiliary field method are proved, such as scaling laws and independence of the results on the choice of P(r). The method is extended in order to find accurate analytical energy formulae for radial potentials of the form aP(r) + V(r), and several explicit examples are studied. Connections existing between the perturbation theory and the auxiliary field method are also discussed.

  11. Aspectual auxiliary verbs in Xitsonga

    African Journals Online (AJOL)

    Kate H

    Let him always come' e. á vá hátl-é vá yá étlélà. OPT 3PL quickly-OPT 3PL go sleep. 'Let them quickly go to bed'. 3.4 Negative markers. Negation is marked on AA verbs. The auxiliary verb hatla 'quickly' is negated in three tenses.

  12. Comparative Studies of the Operation Method of Solar Energy Water Heating System with Auxiliary Heat Pump Heater%热泵辅助供热太阳能热水系统运行模式对比分析

    Institute of Scientific and Technical Information of China (English)

    林辩启; 罗会龙; 王浩; 田盼雨

    2015-01-01

    太阳能热水系统与热泵辅助供热合理结合可取长补短,有效降低建筑能耗。简要概述了空气源热泵、水源热泵、地源热泵辅助供热太阳能热水系统的结构形式及其运行模式。在此基础上,对比分析了热泵辅助供热太阳能热水系统各种典型运行模式的特点及其适用的应用环境。%The appropriate combination of solar water heating system and heat pump auxiliary heating is an effective way to reduce the building energy consumption. The structure and operation method of solar water heating system with different auxiliary heating, such as air-source heat pump, water-source heat pump, and soil-source heat pump, were introduced briefly. The characteristics of all kinds of solar water heating system with auxiliary heating were compared and analyzed. The suitable application environment of solar water heating system with auxiliary heating was also presented.

  13. Bistable energy harvesting enhancement with an auxiliary linear oscillator

    Science.gov (United States)

    Harne, R. L.; Thota, M.; Wang, K. W.

    2013-12-01

    Recent work has indicated that linear vibrational energy harvesters with an appended degree-of-freedom (DOF) may be advantageous for introducing new dynamic forms to extend the operational bandwidth. Given the additional interest in bistable harvester designs, which exhibit a propitious snap through effect from one stable state to the other, it is a logical extension to explore the influence of an added DOF to a bistable system. However, bistable snap through is not a resonant phenomenon, which tempers the presumption that the dynamics induced by an additional DOF on bistable designs would inherently be beneficial as for linear systems. This paper presents two analytical formulations to assess the fundamental and superharmonic steady-state dynamics of an excited bistable energy harvester to which is attached an auxiliary linear oscillator. From an energy harvesting perspective, the model predicts that the additional linear DOF uniformly amplifies the bistable harvester response magnitude and generated power for excitation frequencies less than the attachment’s resonance while improved power density spans a bandwidth below this frequency. Analyses predict bandwidths having co-existent responses composed of a unique proportion of fundamental and superharmonic dynamics. Experiments validate key analytical predictions and observe the ability for the coupled system to develop an advantageous multi-harmonic interwell response when the initial conditions are insufficient for continuous high-energy orbit at the excitation frequency. Overall, the addition of an auxiliary linear oscillator to a bistable harvester is found to be an effective means of enhancing the energy harvesting performance and robustness.

  14. An evaluation of the Davis-Besse loss of feedwater event (June 1985) from an accident management perspective

    International Nuclear Information System (INIS)

    Di Salvo, R.; Leonard, M.T.; Wreathall, J.

    1986-01-01

    An accident management perspective is used to analyze events associated with a total loss-of-feedwater at the Davis-Besse nuclear power plant in June 1985. The relationships of accident management to the closely associated concepts of risk management and emergency management are delineated. The analysis shows that the principal contributors to the event's occurrence were shortcomings in risk management. Successful performance by the operators in accident management was principally responsible for terminating the event without consequence to public health

  15. Dynamic analysis of auxiliary buildings in nuclear power plants

    International Nuclear Information System (INIS)

    Subramanian, K.V.; Madhava Rao, A.S.; Warudkar, A.S.

    1989-01-01

    All nuclear power plants have a large number of auxiliary buildings housing various services and control systems required for the operation of the plant. Illustrative examples are turbine building, control building, service building etc. These buildings are seismically qualified as Class I or Class II structures. Usually, these auxiliary buildings are of low rise type with two or three floors and floor heights varying from five to eight meters and of framed construction in steel or concrete or a combination of both the materials. The floors are usually staggered with large cutouts and may not extend over the full area in plan. Some of the bays are often of double story height with the columns continuous over a story in order to accommodate cranes and other equipment. The structural elements supporting the roof may consist of steel roof trusses instead of beams. The seismic analysis of these structures involves the formulation of the analytical model that can simulate the physical behavior of the structure as close as possible taking into consideration the practical aspects. The criteria adopted to formulate the mathematical model has an important bearing on the evaluated dynamic characteristics and seismic response

  16. Research on vibration properties of auxiliary bearing cage used in HTR-10 GT project

    International Nuclear Information System (INIS)

    Qin Qingquan; Yang Guojun; Shi Zhengang; Yu Suyuan

    2009-01-01

    Auxiliary Bearings (ABs) is one of the most important parts in Active Magnetic Bearing (AMB) system, which was used in HTR-10 GT project. This paper uses finite element method to analyze the centrifugal stress and free vibration properties of the cage according to its work condition. And different geometric parameters of the cage that has effects on its vibration performance are discussed. The results show that the highest centrifugal stress is in the middle of the cage side sill. The low odder vibration modes of the cage can be induced when the auxiliary bearings are working. Proper geometric parameters and ball pocket number can enhance the performance of the cage. (authors)

  17. Theoretical and numerical analysis of auxiliary heating for cryogenic target fabrication

    International Nuclear Information System (INIS)

    Yang Xiaohu; Tian Chenglin; Yin Yan; Xu Han; Zhuo Hongbin

    2008-01-01

    In order to compensate for the nonspherical-symmetric heat flux in the hohlraum, auxiliary heating is usually applied to the outside wall of the hohlraum during the cooling process. A one-dimensional heat exchange theoretical model has been proposed in the indirect-drive target, to analyze the required auxiliary heat flux. With a two dimensional axisymmetric model, the auxiliary heating mechanism has been simulated by FLUENT code. The optimum heat flux which is 635 W/m 2 has been obtained as the heaters around the outside of the hohlraum about 1.3 mm above and below the mid-plane. The result is in good agreement with the theoretical model. (authors)

  18. The Future Mission Tasking and Resourcing of the U.S. Coast Guard Auxiliary

    Science.gov (United States)

    2012-09-01

    the years. Therefore, this study cannot present what the demographic makeup of the Auxiliary was at any past point, and cannot identify if there were...Auxiliary, and the Nation’s projected demographic composition. The demographic makeup of the country, the country’s future volunteers, is changing...command in the spring of 2010.104 This one-page document broad- brushes three prioritized overarching Auxiliary missions: 1) to promote and improve

  19. 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

  20. 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.