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Sample records for auxiliary feedwater system

  1. Auxiliary feedwater system aging study

    International Nuclear Information System (INIS)

    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

  2. System Study: Auxiliary Feedwater 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-31

    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 2013 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.

  3. 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.

  4. Operating experiences and degradation detection for auxiliary feedwater systems

    International Nuclear Information System (INIS)

    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

  5. Design Characteristics of the Passive Auxiliary Feedwater System in APR+

    International Nuclear Information System (INIS)

    The passive auxiliary feedwater system (PAFS) is a typical passive safety system implemented for the APR+. The auxiliary feedwater system (AFWS) in the APR1400, which is the reference plant of the APR+, consists of two motor driven pumps, two turbine driven pumps, two water storage tanks, and related pipes and valves. The AFWS feeds emergency water to steam generators to cool down the reactor coolant system when the main feedwater is lost. To enhance the safety, the PAFS replaces the AFWS with a passive condensation heat exchanger (PCHX), a passive condensation cooling tank (PCCT), and a few valves and pipes in the APR+ design. In this paper, we propose the design requirements and conceptual design of the PAFS in order to evaluate the operability of the PAFS, to develop the APR+'s general arrangements for the auxiliary building, and to identify the important parameters to be quantified by experiments

  6. Performance analysis of a Passive Auxiliary Feedwater System in APR+

    International Nuclear Information System (INIS)

    The Advanced Power Reactor Plus (APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system (PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system (AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. The purpose of this paper is to evaluate the performance of the PAFS under design basis events using best-estimated thermalhydraulic codes

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

    International Nuclear Information System (INIS)

    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

  8. The development of a passive auxiliary feedwater system in APR+

    International Nuclear Information System (INIS)

    The Advanced Power Reactor Plus (APR+) is being developed in Korea. APR+ is a GEN III+ reactor on the basis of the APR1400. To meet the requirements of GEN III+ reactors, the economics and the safety of the APR+ are further enhanced. One of the basic principles of APR+ safety systems is the adoption of hybrid safety systems. Passive safety systems replace the current active safety systems from an economic point of view. The passive aux. feedwater system (PAFS) is one of the passive safety features adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system by introducing a natural driving force mechanism while maintaining the system's basic Junction of cooling down the primary side and removing the decay heat. In order to satisfy the single failure criterion, the PAFS is composed of two independent trains. Each train has one steam condensing heat exchanger of 100% capability and one PCCT (Passive Condensation Cooling water Tank) of 100% capability. Basic design is underway and separate effect tests and integral effect tests will be performed to demonstrate the performance of the PAFS. (authors)

  9. Seismic qualification of PWR plant auxiliary feedwater systems

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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. MARS calculation of PAFS (passive auxiliary feedwater system) heat exchanger in APR+

    International Nuclear Information System (INIS)

    APR+ (Advanced Power Reactor Plus), the next generation nuclear power plant in Korea, adopts PAFS (Passive Auxiliary Feedwater System) as one of the advanced safety feature. To design the condensation heat exchanger in PAFS, the two-phase flow phenomena in horizontal U-tube and PCCT (Passive Condensate Cooling Tank) were investigated by MARS calculation. By benchmarking with NOKO experimental result, MARS code showed a reasonable capability to quantitatively predict the condensation in horizontal tube heat exchanger. And the design of PAFS heat exchanger was proved to sufficiently remove the decay heat by the condensation heat transfer without any active auxiliary feedwater system

  13. Evaluation of Effect of N2 Gas on the Cooling Capability of Passive Auxiliary Feedwater System

    International Nuclear Information System (INIS)

    Advanced Power Reactor Plus (APR+), a next generation nuclear power plant in Korea, adopts Passive Auxiliary Feedwater System (PAFS) to replace the conventional active auxiliary feedwater system. Because PAFS removes decay heat from the reactor core, it is required to verify the performance of PAFS in postulated accidents cases. In addition, an effect of non-condensable gas such as N2 gas on the heat removal capability of PAFS should be evaluated since the non-condensable gas may deteriorate a condensation heat transfer through the condensation heat exchanger in PAFS. In this study, MARS code is used to evaluate the effect of N2 gas

  14. MARS calculation of PAFS (passive auxiliary feedwater system) heat exchanger in APR+

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byoung Uhn; Yun, Byong Jo; Bae, Sung Won; Choi, Ki Yong; Song, Chul Hwa [KAERI, Daejeon (Korea, Republic of); Cheon, Jong [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2009-07-01

    APR+ (Advanced Power Reactor Plus), the next generation nuclear power plant in Korea, adopts PAFS (Passive Auxiliary Feedwater System) as one of the advanced safety feature. To design the condensation heat exchanger in PAFS, the two-phase flow phenomena in horizontal U-tube and PCCT (Passive Condensate Cooling Tank) were investigated by MARS calculation. By benchmarking with NOKO experimental result, MARS code showed a reasonable capability to quantitatively predict the condensation in horizontal tube heat exchanger. And the design of PAFS heat exchanger was proved to sufficiently remove the decay heat by the condensation heat transfer without any active auxiliary feedwater system.

  15. Resolution of concerns in auxiliary feedwater piping

    International Nuclear Information System (INIS)

    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

  16. Integral effect test on operational performance of the PAFS (Passive Auxiliary Feedwater System) for a FLB (Feedwater Line Break) accident

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byoung Uhn; Kim, Seok; Park, Yu Sun; Kim, Bok Deuk; Kang, Kyoung Ho [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    PAFS (Passive Auxiliary Feedwater System) is one of the advanced safety features adopted in the APR+, which is intended to completely replace a conventional active auxiliary feedwater system. It cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism; i.e., condensing steam in the PCHX (Passive Condensation Heat Exchanger) submerged inside the PCCT (Passive Condensate Cooling Tank). The test facility, ATLAS PAFS, was constructed to experimentally investigate the thermal hydraulic behavior in the primary and secondary systems of the APR+ during the transient when the PAFS is actuated. Among the anticipated accidents with the PAFS actuation, the FLB (Feedwater Line Break) was considered as the most important accident in evaluating the cooling capability of the PAFS, during the development of PIRT (Phenomena Identification and Ranking Table) of the PAFS. In this study, the PAFS FLB EC 01 test was performed to simulate a break on the pipe connected to the SG 1 economizer, which was analyzed as the most severe case in the APR+ SSAR (Standard Safety Analysis Report). The main objectives of this test were not only to provide physical insight into the system response of the APR+ during the FLB accident but also to produce an integral effect test data to validate a thermal hydraulic safety analysis code.

  17. Integral effect test on operational performance of the PAFS (Passive Auxiliary Feedwater System) for a FLB (Feedwater Line Break) accident

    International Nuclear Information System (INIS)

    PAFS (Passive Auxiliary Feedwater System) is one of the advanced safety features adopted in the APR+, which is intended to completely replace a conventional active auxiliary feedwater system. It cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism; i.e., condensing steam in the PCHX (Passive Condensation Heat Exchanger) submerged inside the PCCT (Passive Condensate Cooling Tank). The test facility, ATLAS PAFS, was constructed to experimentally investigate the thermal hydraulic behavior in the primary and secondary systems of the APR+ during the transient when the PAFS is actuated. Among the anticipated accidents with the PAFS actuation, the FLB (Feedwater Line Break) was considered as the most important accident in evaluating the cooling capability of the PAFS, during the development of PIRT (Phenomena Identification and Ranking Table) of the PAFS. In this study, the PAFS FLB EC 01 test was performed to simulate a break on the pipe connected to the SG 1 economizer, which was analyzed as the most severe case in the APR+ SSAR (Standard Safety Analysis Report). The main objectives of this test were not only to provide physical insight into the system response of the APR+ during the FLB accident but also to produce an integral effect test data to validate a thermal hydraulic safety analysis code

  18. Construction Report of Separate Effect Test Facility for Passive Auxiliary Feedwater System (PASCAL)

    International Nuclear Information System (INIS)

    A separate effect test facility for PAFS(Passive Auxiliary Feedwater System, PAFS), PASCAL, was constructed to evaluate the cooling performance of PAFS and the condensation heat transfer models. This report includes the scope of the separate effect tests, the design of PASCAL facility, and measuring principles. From the design and construction of the separate effect test facility, PASCAL facility was composed of the fluid system, the auxiliary system, the measurement system, the electricity system, the control system and the data acquisition system. This report will be utilized to make the experiment procedure and perform the test

  19. Equipment Reliability Improvement for Koeberg Nuclear Power Plant Auxiliary Feedwater System

    International Nuclear Information System (INIS)

    This paper investigated how the performance of the Koeberg Auxiliary Feedwater System could be improved using the 'maintenance rule'. As a conclusion, this paper figured out AFWS pumps and the TDP control circuit need special attention in improving the reliability of the AFWS, this lead to an improved maintenance strategy for the system. The purpose of this study is to apply maintenance rule to enhance the Auxiliary Feedwater System (AFWS) maintenance strategy at Koeberg Nuclear Power Plant (KNPP). Currently, Koeberg AFWS health status is red, needing an improvement. This study seeks to use maintenance rule to identify components that enable AFWS to fulfill its essential functions so as to focus maintenance resources and have the greatest beneficial impact on improving reliability and availability of the system

  20. Scaling Analysis of Separate Effect Test Facility for PAFS (Passive Auxiliary Feedwater System)

    International Nuclear Information System (INIS)

    PAFS (Passive Auxiliary Feedwater System) is one of the passive cooling systems of APR+. It can replace the conventional active system for auxiliary feedwater injection to the steam generator. A diagram of PAFS in APR+ is shown in Figure 1. It cools down the secondary system by heat transfer at a horizontal U-tube heat exchanger in PCCT (Passive Condensation Cooling Tank). To validate a performance of PAFS, separate effect test loop is being developed, which is named as PASCAL(PAFS Condensing heat removal Assessment Loop). This study aims at analyzing the scaling effect of PASCAL by MARS (Multi-dimensional Analysis for Reactor Safety) code analysis. Transient simulation results for the case of LOCV(Loss of Condenser Vacuum) scenario were compared between PASCAL and prototype

  1. Scaling Analysis of Separate Effect Test Facility for PAFS (Passive Auxiliary Feedwater System)

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byoung Uhn; Yun, Byong Jo; Bae, Sung Won; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    PAFS (Passive Auxiliary Feedwater System) is one of the passive cooling systems of APR+. It can replace the conventional active system for auxiliary feedwater injection to the steam generator. A diagram of PAFS in APR+ is shown in Figure 1. It cools down the secondary system by heat transfer at a horizontal U-tube heat exchanger in PCCT (Passive Condensation Cooling Tank). To validate a performance of PAFS, separate effect test loop is being developed, which is named as PASCAL(PAFS Condensing heat removal Assessment Loop). This study aims at analyzing the scaling effect of PASCAL by MARS (Multi-dimensional Analysis for Reactor Safety) code analysis. Transient simulation results for the case of LOCV(Loss of Condenser Vacuum) scenario were compared between PASCAL and prototype

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

    International Nuclear Information System (INIS)

    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

  3. Separate-effect Test for Cooling Performance of PAFS(Passive Auxiliary Feedwater System)

    International Nuclear Information System (INIS)

    APR+ (Advanced Power Reactor Plus) is a next generation nuclear power plant being developed in Korea. It adopts PAFS (Passive Auxiliary Feedwater System) for the steam generator (SG) instead of an active auxiliary feedwater system for the conventional nuclear power plant (NPP). It can replace the conventional active auxiliary feedwater system for the SG by a passive way. It is composed of a steam-supply line, a condensation heat exchanger, a return-water line, and a PCCT (Passive Condensate Cooling Tank). When the water level in the SG becomes lower than 25% of the wide range of the water level transmitter during an accident situation, the actuation valve at the return-water line is open and then the natural convection flow of the PAFS can be made. To validate a cooling performance of PAFS, separate effect test loop, which is named PASCAL (PAFS Condensing heat removal Assessment Loop) was constructed at KAERI (Korea Atomic Energy Research Institute) for investigating the cooling capability of the condensation heat exchanger and the characteristic of the natural convection. This study focuses on the experimental study of the separate effect test with PASCAL facility. From the experimental results, two-phase flow phenomena in the condensation heat exchanger and PCCT are investigated for the verification of the design of PAFS

  4. Separate-effect Test for Cooling Performance of PAFS(Passive Auxiliary Feedwater System)

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byoung Uhn; Kim, Seok; Kang, Kyung Ho; Yun, Byong Jo; Kim, Bok Duk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    APR+ (Advanced Power Reactor Plus) is a next generation nuclear power plant being developed in Korea. It adopts PAFS (Passive Auxiliary Feedwater System) for the steam generator (SG) instead of an active auxiliary feedwater system for the conventional nuclear power plant (NPP). It can replace the conventional active auxiliary feedwater system for the SG by a passive way. It is composed of a steam-supply line, a condensation heat exchanger, a return-water line, and a PCCT (Passive Condensate Cooling Tank). When the water level in the SG becomes lower than 25% of the wide range of the water level transmitter during an accident situation, the actuation valve at the return-water line is open and then the natural convection flow of the PAFS can be made. To validate a cooling performance of PAFS, separate effect test loop, which is named PASCAL (PAFS Condensing heat removal Assessment Loop) was constructed at KAERI (Korea Atomic Energy Research Institute) for investigating the cooling capability of the condensation heat exchanger and the characteristic of the natural convection. This study focuses on the experimental study of the separate effect test with PASCAL facility. From the experimental results, two-phase flow phenomena in the condensation heat exchanger and PCCT are investigated for the verification of the design of PAFS

  5. Analysis of Condensation Phenomena in PAFS (Passive Auxiliary Feedwater System) Horizontal Heat Exchanger of APR+

    International Nuclear Information System (INIS)

    APR+ (Advanced Power Reactor Plus) is the next generation nuclear power plant in Korea. It adopts PAFS (Passive Auxiliary Feedwater System) on the secondary system. It can replace the conventional active system for auxiliary feedwater injection to the steam generator, and it enable to supply the coolant by a passive system. It cools down the secondary system by heat transfer at a horizontal U-tube in PCCT (Passive Condensate Cooling Tank). High pressure steam flow from the steam generator is condensed in the horizontal heat exchanger. The water in PCCT is maintained at an atmospheric pressure, so that boiling heat transfer at the outside wall of heat exchanger and natural convection occur in PCCT pool. The heat exchanger and PCCT is higher than steam generator, so condensate can be drained and injected to feedwater system without any active system. This study aims at design of the horizontal heat exchanger in PAFS. It should remove the heat generated in the steam generator. To satisfy this requirement, a system code analysis is conducted. The amount of condensation heat transfer is investigated by MARS (Multi-dimensional Analysis for Reactor Safety) code analysis

  6. Analysis of Condensation Phenomena in PAFS (Passive Auxiliary Feedwater System) Horizontal Heat Exchanger of APR+

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byoung Uhn; Yun, Byong Jo; Bae, Sung Won; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cheon, Jong [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2009-10-15

    APR+ (Advanced Power Reactor Plus) is the next generation nuclear power plant in Korea. It adopts PAFS (Passive Auxiliary Feedwater System) on the secondary system. It can replace the conventional active system for auxiliary feedwater injection to the steam generator, and it enable to supply the coolant by a passive system. It cools down the secondary system by heat transfer at a horizontal U-tube in PCCT (Passive Condensate Cooling Tank). High pressure steam flow from the steam generator is condensed in the horizontal heat exchanger. The water in PCCT is maintained at an atmospheric pressure, so that boiling heat transfer at the outside wall of heat exchanger and natural convection occur in PCCT pool. The heat exchanger and PCCT is higher than steam generator, so condensate can be drained and injected to feedwater system without any active system. This study aims at design of the horizontal heat exchanger in PAFS. It should remove the heat generated in the steam generator. To satisfy this requirement, a system code analysis is conducted. The amount of condensation heat transfer is investigated by MARS (Multi-dimensional Analysis for Reactor Safety) code analysis.

  7. Evaluation of the APR+ Passive Auxiliary Feedwater System Performance during Main Feedwater Line Break Accident using MARS-KS

    International Nuclear Information System (INIS)

    Ever since the Nuclear Power Plant (NPP) started commercial operation, advanced NPPs have been developed to enhance performance and safety as well as the economics of the plant. As a part of a regulatory safety research of the advanced nuclear reactors, MARS-KS regulatory safety analysis code has been selected to evaluate the performance of the Passive Auxiliary Feedwater System (PAFS) during Main Feedwater Line Break (MFLB) accident of the APR+ (Advanced Power Reactor+) which is under development by Korea Hydro and Nuclear Power (KHNP). The results of the APR+ MFLB analysis and the performance of the PAFS are presented herein. MATS-KS MFLB analysis shows that the MARS-KS code well simulates dynamic thermal hydraulic behavior of the MFLB and maximum RCS pressure satisfies the acceptance criteria of 120% design RCS pressure for the MFLB accident. APR+ PAFS effectively removes the core decay heat by the natural circulation during the MFLB accidents, however, comprehensive performance of the PAFS should be evaluated against the design basis of 8 hours core heat removal until the conditions for the initiation of the Shutdown Cooling System (350 .deg. F and 400 psia) are met

  8. Evaluation of the APR+ Passive Auxiliary Feedwater System Performance during Main Feedwater Line Break Accident using MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Minjeong; Ralph, Marigomena; Sim, S. K. [Environment and Energy Technology, Inc., Daejeon (Korea, Republic of); Bang, Youngseok [KINS, Daejeon (Korea, Republic of)

    2013-05-15

    Ever since the Nuclear Power Plant (NPP) started commercial operation, advanced NPPs have been developed to enhance performance and safety as well as the economics of the plant. As a part of a regulatory safety research of the advanced nuclear reactors, MARS-KS regulatory safety analysis code has been selected to evaluate the performance of the Passive Auxiliary Feedwater System (PAFS) during Main Feedwater Line Break (MFLB) accident of the APR+ (Advanced Power Reactor+) which is under development by Korea Hydro and Nuclear Power (KHNP). The results of the APR+ MFLB analysis and the performance of the PAFS are presented herein. MATS-KS MFLB analysis shows that the MARS-KS code well simulates dynamic thermal hydraulic behavior of the MFLB and maximum RCS pressure satisfies the acceptance criteria of 120% design RCS pressure for the MFLB accident. APR+ PAFS effectively removes the core decay heat by the natural circulation during the MFLB accidents, however, comprehensive performance of the PAFS should be evaluated against the design basis of 8 hours core heat removal until the conditions for the initiation of the Shutdown Cooling System (350 .deg. F and 400 psia) are met.

  9. Analysis of Heat Removal Capability of PAFS (Passive Auxiliary Feedwater System) in APR (Advanced Power Reactor Plus)

    International Nuclear Information System (INIS)

    As passive safety features for nuclear power plants receive increasing attention, South Korea has designed PAFS (Passive Auxiliary Feedwater System) for APR+ (Advanced Power Reactor Plus). Because the PAFS replaces a conventional active auxiliary feedwater system and plays a role in the ultimate heat sink for decay heat, it is necessary to evaluate the heat removal capability of PAFS under postulated accidents conditions. Therefore, the performance analysis is carried out for two accident cases: Loss of Condenser Vacuum (LOCV) and Feedwater Line Break (FLB) accidents. For the analysis, MARS-KS code is used and MARS-KS model is developed by adding PAFS model to the existing APR1400 model

  10. Analysis of Heat Removal Capability of PAFS (Passive Auxiliary Feedwater System) in APR (Advanced Power Reactor Plus)

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Y. J.; Kang, K. H.; Yun, B. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-03-15

    As passive safety features for nuclear power plants receive increasing attention, South Korea has designed PAFS (Passive Auxiliary Feedwater System) for APR+ (Advanced Power Reactor Plus). Because the PAFS replaces a conventional active auxiliary feedwater system and plays a role in the ultimate heat sink for decay heat, it is necessary to evaluate the heat removal capability of PAFS under postulated accidents conditions. Therefore, the performance analysis is carried out for two accident cases: Loss of Condenser Vacuum (LOCV) and Feedwater Line Break (FLB) accidents. For the analysis, MARS-KS code is used and MARS-KS model is developed by adding PAFS model to the existing APR1400 model.

  11. Assessment of Heat Removal Capability of Passive Auxiliary Feedwater System using MARS Code

    International Nuclear Information System (INIS)

    Passive Auxiliary Feedwater System (PAFS) is one of advanced safety features under development for Advanced Power Reactor Plus (APR+). Because PAFS removes decay heats from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of PAFS under the postulated accidents conditions. The target accidents cases analyzed in this study are the Loss of Condenser Vacuum (LOCV) and the Main Feedwater Line Break (MFLB). In the case of LOCV accident, PAFS in both loops are available but a single loop is operational in MFLB accident condition. Thus, these two accidents scenario are the proper selection to evaluate the capability of PAFS. For the analysis, MARS code is utilized and MARS model for PAFS is developed

  12. Maintenance centered on reliability applied to a NPP auxiliary feedwater system

    International Nuclear Information System (INIS)

    The main objective of maintenance in a NPP is to assure that structures, systems and components will perform their design functions with reliability and/or availability in order to allow a safe and economic electric power generation. Reliability-Centered Maintenance (RCM) is a method of systematic review to either develop or optimize Preventive Maintenance Programs. This paper presents the objectives, concepts, organization and methods used in the application of RCM to NPP. Some application examples are include in this paper, considering some components of the Auxiliary Feedwater System of a generic Westinghouse designed two-loop PWR NPP. (author). 4 refs., 3 figs

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

  15. Experimental program for validation of cooling and operational performance of the APR+ Passive auxiliary feedwater system

    International Nuclear Information System (INIS)

    PAFS (Passive Auxiliary Feedwater System) is one of the advanced passive safety systems adopted in the APR+ (Advanced Power Reactor plus), which is intended to completely replace the conventional active auxiliary feedwater system. PAFS cools down the steam generator's secondary side, and eventually removes the decay heat from the reactor core by introducing a natural driving force mechanism; i.e., condensing steam in nearly horizontal U-tubes submerged inside the passive condensation cooling tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, an experimental program is in progress at KAERI (Korea Atomic Energy Research Institute), which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL (PAFS Condensing Heat Removal Assessment Loop), is in progress to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. The integral effect test is being performed to confirm the operational performance of the PAFS coupled with the other reactor coolant systems (RCS) using the thermal hydraulic integral effect test facility, ATLAS (Advanced Thermal hydraulic test Loop for Accident Simulation). This paper summarizes the up to date experimental results of the separate effect test and the integral effect test for PAFS from a cooling and operational performance point of view

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

    International Nuclear Information System (INIS)

    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

  17. 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

  18. Experimental study on the operational and the cooling performance of the APR+ passive auxiliary feedwater system

    International Nuclear Information System (INIS)

    The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+ which is intended to completely replace the conventional active auxiliary feedwater system. The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by introducing a natural driving force mechanism; i.e., condensing steam in nearly-horizontal U-tubes submerged inside the passive condensation cooling tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, the separate effect test, PASCAL (PAFS Condensing Heat Removal Assessment Loop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. A single nearly-horizontal U-tube whose dimension is same as the prototypic U-tube of the APR+ PAFS is simulated in the PASCAL test. By performing the PASCAL test, the major thermal-hydraulic parameters such as local/overall heat transfer coefficients, fluid temperature inside the tube, wall temperature of the tube, and pool temperature distribution in the PCCT were produced not only to evaluate the current condensation heat transfer model but also to present database for the safety analysis related with the PAFS. (authors)

  19. Feasibility study of helically coiled tube condensation heat exchanger for a passive auxiliary feedwater system

    International Nuclear Information System (INIS)

    The Passive Auxiliary Feedwater System (PAFS) with nearly-horizontal heat exchangers is one of passive safety features of APR+ (Advanced Power Reactor Plus) which provides the auxiliary feedwater by means of natural circulation with condensation. It is feasible to increase the heat transfer capacity of the PAFS by employing a helically coiled heat exchanger due to additional secondary flow effect by centrifugal force. In addition, a compact and flexible design can be achieved in a fixed volume by using the helically coiled heat exchanger, which is one of the most important merits of implementing this heat exchanger. In this paper, the helically coiled heat exchanger has been employed for the PAFS instead of nearly-horizontal heat exchanger. In order to evaluate the heat transfer performance of the helically coiled heat exchanger, an in-tube condensation heat transfer correlation by Wongwises has been introduced into the system analysis code, MARS-KS. A comparative numerical study was conducted for both heat exchangers. The result shows that helically coiled heat exchanger has 20% higher heat transfer efficiency than existing nearly-horizontal heat exchanger. (author)

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

    International Nuclear Information System (INIS)

    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

  1. Assessing influence of the auxiliary emergency feedwater system on Rivne-1,2 WWER-440/213 core damage frequency

    International Nuclear Information System (INIS)

    The influence of the auxiliary emergency feedwater system on Rivne-1 ana 2 core damage frequency is assessed in this paper. The influence is assessed by means of probabilistic safety analysis methods. Results of preliminary and already implemented designs of the system are compared. Besides, the paper presented specifics of modeling the system elements

  2. Analytical studies of the heat removal capability of a passive auxiliary feedwater system (PAFS)

    International Nuclear Information System (INIS)

    Highlights: ► The MARS model was developed by adding the PAFS model to the APR1400 model. ► Analysis results show that the capacity of PAFS is sufficient to remove the decay heat in the LOCV and FLB accident cases. ► The PAFS control logic for MSIV has the advantages of maintaining the feedwater inventory in the intact side. - Abstract: As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for third-generation (GEN-III) nuclear power plants that are driven by passive systems, such as natural circulation, gravity, and resistance to high temperatures. Thus, South Korea has designed the Advanced Power Reactor Plus (APR+) with a two-loop PWR and 1500 MWe by adding passive safety features to the Advanced Power Reactor 1400 MWe (APR1400). The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the APR+, and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. Therefore, in this paper, after introducing the characteristics of the PAFS and its design requirements, a performance analysis of the PAFS is performed for two accident cases: Loss of Condenser Vacuum (LOCV) and Feedwater Line Break (FLB). For the analysis, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used, and a MARS model is developed by adding the PAFS model to the existing APR1400 model. The analysis results show that the PAFS has enough capacity to remove decay heat under the postulated accident conditions. In addition, the adequacy of modified control logic for main steam isolation valve (MSIV) is validated by comparing the traditional control logic.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byoung-Uhn, E-mail: bubae@kaeri.re.kr; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

    2014-08-15

    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.

  4. 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)

    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

  5. Investigation on Ledinegg Instability in Condensate Tube of Passive Auxiliary Feedwater System

    International Nuclear Information System (INIS)

    Passive Auxiliary Feedwater System (PAFS) is one of advanced safety features under development for Advanced Power Reactor Plus (APR+). Because the condensate flow is driven by natural circulation, it is important to ensure not to induce instabilities inside the condensate tube in PAFS for the effective cooling capability. Among the flow instabilities, the Ledineggtype instability may cause the severe deterioration of heat removal capability of PAFS since it can reduce the condensate flow even with slight change of pressure loss. Because the Ledinegg instability occurs when the pressure drop decreases with increasing flow, to evaluate the behavior of the pressure drop according to the change of mass flow rate is essential. For this reason, one-dimensional, integrated flow model is formulated and two-phase flow characteristics in the condensate tube are mathematically solved

  6. Feasibility Study on Passive Auxiliary Feedwater System in Loss of Condenser Vacuum Accident

    International Nuclear Information System (INIS)

    Nuclear leading countries are developing and constructing technology intensive pressurized water reactors (PWRs) such as AP1000 (United State), EPR (Europe), and US-APWR (Japan), and these advanced reactors adopt several passive safety features in order to enhance the safety and reliability. Domestic advanced reactor APR1400 already completed the earlier development in 2002, and technology gap from the nuclear leading countries become large. In particular, China requires technology transfer in the order of new power plant construction. Thus it is expected difficult to export the power plant to the newly developing countries without our own technology. Therefore, the improvement of competitive power and establishment of infra structure of advanced nuclear industry through innovative technology enhancement are urgent and essential to international competitive marketing. Passive safety features have been always adopted as an improved design concept in the development of innovative reactor design. Domestic nuclear industry has stated the development of APR+ as a Korean specific reactor for the export strategy. In the development of APR+ a passive auxiliary feedwater system (PAFS) has been considered as a noticeable candidate of improved design. Reference 2 reported that the adoption of PAFS, which can replace the auxiliary feedwater system, can prevent core damage in the accident of station black out (SBO), since Class 1E DC power operates the related valves, and 8 hours hot standby operation of plant without operation action is achievable. This PAFS contributes to the safety and economics, in that it decreases the core damage frequency 26% from 2.45E- 06/r-y to 1.80E-06/r-y, and it saves the construction cost 20 million Kr-Won. This paper discusses on the performance of PAFS for the accident of loss of condenser vacuum as a precursor of detailed design specification

  7. Feasibility Study on Passive Auxiliary Feedwater System in Loss of Condenser Vacuum Accident

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon; Lee, Byung Chul [FNC Tech., Daejeon (Korea, Republic of); Cheon, Jong; Kim, Han Gon [NETEC, Daejeon (Korea, Republic of)

    2009-05-15

    Nuclear leading countries are developing and constructing technology intensive pressurized water reactors (PWRs) such as AP1000 (United State), EPR (Europe), and US-APWR (Japan), and these advanced reactors adopt several passive safety features in order to enhance the safety and reliability. Domestic advanced reactor APR1400 already completed the earlier development in 2002, and technology gap from the nuclear leading countries become large. In particular, China requires technology transfer in the order of new power plant construction. Thus it is expected difficult to export the power plant to the newly developing countries without our own technology. Therefore, the improvement of competitive power and establishment of infra structure of advanced nuclear industry through innovative technology enhancement are urgent and essential to international competitive marketing. Passive safety features have been always adopted as an improved design concept in the development of innovative reactor design. Domestic nuclear industry has stated the development of APR+ as a Korean specific reactor for the export strategy. In the development of APR+ a passive auxiliary feedwater system (PAFS) has been considered as a noticeable candidate of improved design. Reference 2 reported that the adoption of PAFS, which can replace the auxiliary feedwater system, can prevent core damage in the accident of station black out (SBO), since Class 1E DC power operates the related valves, and 8 hours hot standby operation of plant without operation action is achievable. This PAFS contributes to the safety and economics, in that it decreases the core damage frequency 26% from 2.45E- 06/r-y to 1.80E-06/r-y, and it saves the construction cost 20 million Kr-Won. This paper discusses on the performance of PAFS for the accident of loss of condenser vacuum as a precursor of detailed design specification.

  8. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  12. Evaluation of Effect of N2 Gas on the Cooling Capability of Passive Auxiliary Feedwater System (PAFS) in APR+

    International Nuclear Information System (INIS)

    In Korea, Advanced Power Reactor Plus (APR+) has being developed by adding passive safety features to Advanced Power Reactor 1400MWe (APR1400). Passive Auxiliary Feedwater System (PAFS) is one of passive system adopted in the APR+ to replace the conventional active auxiliary feedwater system. Because PAFS removes decay heat from the reactor core, it is required to verify the performance of PAFS in postulated accidents cases. In addition, an effect of noncondensable gas on the heat removal capability of PAFS should be evaluated since the non-condensable gas may deteriorate a condensation heat transfer through the condensation heat exchanger in PAFS. In this study, the effect of N2 gas was evaluated using MARS

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

    International Nuclear Information System (INIS)

    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

  14. Auxiliary feedwater system risk-based inspection guide for the Virgil C. Summer Nuclear Power Plant

    International Nuclear Information System (INIS)

    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 emergency/auxiliary feedwater (EFW/AFW) system at press water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses costing 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 genetic 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. Virgil C. Summer plant was selected as one m a series of plants for study. The product of this effort is a priority 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 Virgil C. Summer plant

  15. Study on the importance and sensibility of the parameters used in the Angra-1 auxiliary feedwater system reliability analysis

    International Nuclear Information System (INIS)

    In this paper some procedures are presented in order to develop an importance and sensitivity analysis on the parameters considered in the reliability study of the Auxiliary Feedwater System of Angra-1. The importance analysis is performed to determine the events which have contribution on the top event. The results obtained from the sensitivity analysis can show the effects of variations in probability values of the dominant component failures on the probability of the top event. (author). 7 refs., 9 figs., 5 tabs

  16. Analysis of a potential two phase flow instability in a PWR passive auxiliary feedwater system

    International Nuclear Information System (INIS)

    The APR+ incorporates a passive auxiliary feedwater system (PAFS). The PAFS is comprised of two separate mechanical divisions. Each division is a closed loop which is aligned to feed condensed water to its corresponding steam generator (SG), and is equipped with one passive condensation heat exchanger (PCHX), some associated isolation/drain/vent valves, check valves, instrumentation and control, and pipes. The PAFS is designed to start its operation after reactor trip and maintain its function of residual heat removal for 8 hours or longer without AC power or operator action, and to ensure a subsequent cooldown of RCS to the shutdown cooling entry conditions. During the PAFS operation mode, steam in the SG secondary side moves up due to buoyancy force and passes through the main steam line, and then flows into the PCHX where steam is condensed inside the tubes of which the outer wall surfaces are cooled by the water stored in a condensation cooling tank. The condensate is passively fed into the SG economizer by gravity. Because a natural circulation loop is susceptible to two phase flow instability, it is requisite to confirm the PAFS is designed adequately to avoid the potential challenges to its operational safety due to the instability. This paper addresses an analytical model for assessing if the loop has possible thermal and fluid mechanical characteristics which could lead to an undesirable unstable or oscillating water level in the APR+ PAFS

  17. Separate and integral effect tests for validation of cooling and operational performance of the APR+ passive auxiliary feedwater system

    International Nuclear Information System (INIS)

    The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL (PAFS Condensing Heat Removal Assessment Loop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

  18. Design of condensation heat exchanger for the PAFS (Passive Auxiliary Feedwater System) of APR+ (Advanced Power Reactor Plus)

    International Nuclear Information System (INIS)

    Highlights: ► Condensation heat exchanger for the PAFS (Passive Auxiliary Feedwater System) was designed. ► The requirement of the heat removal rate and the prevention of water hammer phenomena were considered. ► The proposed design of the heat exchanger satisfied the requirement of the passive heat removal system. - Abstract: The APR+ (Advanced Power Reactor Plus), a next generation nuclear power plant in Korea, has adopted the PAFS (Passive Auxiliary Feedwater System) on the secondary system of the steam generator (SG) as an advanced safety feature. It is intended to replace the conventional auxiliary feedwater system, which consists of active components for the SG in a passive way. It removes decay heat from the reactor core by cooling down the secondary system of the SG using a condensation heat exchanger installed in the PCCT (Passive Condensation Cooling Tank). The objective of this study is to design a condensation heat exchanger for the PAFS and to evaluate the cooling performance for the proposed design using the thermal hydraulic system analysis code, MARS (Multi-dimensional Analysis for Reactor Safety). Requirements such as the heat removal capacity and the prevention of water hammer were preferentially considered to determine the design parameters of the heat exchanger tube. The MARS code analysis result showed that the proposed design of the PAFS heat exchanger is able to cool down the required amount of decay heat. The distribution of a liquid volume fraction and flow regime predicted by the MARS code shows that the proposed design of the heat exchanger excludes the water hammer inside the tube. Estimation of a two-phase flow pressure drop indicates that the pressure drop inside the tube is negligible compared to the total pressure drop in the PAFS. From the MARS code analysis, it is concluded that the proposed design of the condensation heat exchanger in the PAFS satisfies the overall criteria for the performance of the passive heat removal

  19. 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)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kyu Cho, Hyoung, E-mail: chohk@snu.ac.kr [Seoul National University, Department of Nuclear Engineering, Seoul 151-742 (Korea, Republic of); Cho, Yun Je; Yoon, Han Young [Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2014-07-01

    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.

  1. An experimental study on the validation of cooling capability for the Passive Auxiliary Feedwater System (PAFS) condensation heat exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok; Bae, Byoung-Uhn; Cho, Yun-Je; Park, Yu-Sun; Kang, Kyoung-Ho [Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Yun, Byong-Jo, E-mail: bjyun@pusan.ac.kr [School of Mechanical Engineering, Pusan National University, 30 Jangjeon-dong, Geumjeong-gu, Busan, 609-735 (Korea, Republic of)

    2013-07-15

    Highlights: • PAFS is designed to replace a conventional active Auxiliary Feedwater System. • A SET facility is constructed for investigating the thermal-hydraulic behavior of the PAFS system. • Experimental results proved that the PCHX design satisfied the heat removal requirements. • Results of the MARS-KS code provided a conservative prediction of the heat transfer phenomena. -- Abstract: The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+). PAFS is designed to replace a conventional active Auxiliary Feedwater System (AFWS). The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by a natural circulation mechanism, i.e., condensing steam in nearly horizontal U-tubes submerged inside a pool. A separate effect test facility was constructed with the aim of validating the cooling and operational performance of the PAFS. The PAFS Condensing Heat Removal Assessment Loop (PASCAL) was constructed by simulating a single Passive Condensation Heat Exchanger (PCHX) tube submerged in the Passive Condensation Cooling Tank (PCCT) according to the volumetric scaling methodology. Quasi-steady state (SS) test cases and PCCT level decrease (PL) were sequentially performed with the steam generator heater power set at 540 kW to investigate the thermal-hydraulic behavior of the PAFS system and the characteristics of the natural circulation in the loop. The experimental results proved that the current PCHX design satisfied the heat removal requirement for cooling down the reactor core during an accident condition. Therefore, the PAFS can replace a conventional active AFWS in the APR+ by utilizing the two-phase natural circulation flow. The Multi-dimensional Analysis of Reactor Safety, KINS Standard Version (MARS-KS), a thermal hydraulic system analysis code, was utilized to validate the present experimental data. The results of the MARS

  2. An experimental study on the validation of cooling capability for the Passive Auxiliary Feedwater System (PAFS) condensation heat exchanger

    International Nuclear Information System (INIS)

    Highlights: • PAFS is designed to replace a conventional active Auxiliary Feedwater System. • A SET facility is constructed for investigating the thermal-hydraulic behavior of the PAFS system. • Experimental results proved that the PCHX design satisfied the heat removal requirements. • Results of the MARS-KS code provided a conservative prediction of the heat transfer phenomena. -- Abstract: The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+). PAFS is designed to replace a conventional active Auxiliary Feedwater System (AFWS). The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by a natural circulation mechanism, i.e., condensing steam in nearly horizontal U-tubes submerged inside a pool. A separate effect test facility was constructed with the aim of validating the cooling and operational performance of the PAFS. The PAFS Condensing Heat Removal Assessment Loop (PASCAL) was constructed by simulating a single Passive Condensation Heat Exchanger (PCHX) tube submerged in the Passive Condensation Cooling Tank (PCCT) according to the volumetric scaling methodology. Quasi-steady state (SS) test cases and PCCT level decrease (PL) were sequentially performed with the steam generator heater power set at 540 kW to investigate the thermal-hydraulic behavior of the PAFS system and the characteristics of the natural circulation in the loop. The experimental results proved that the current PCHX design satisfied the heat removal requirement for cooling down the reactor core during an accident condition. Therefore, the PAFS can replace a conventional active AFWS in the APR+ by utilizing the two-phase natural circulation flow. The Multi-dimensional Analysis of Reactor Safety, KINS Standard Version (MARS-KS), a thermal hydraulic system analysis code, was utilized to validate the present experimental data. The results of the MARS

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

    International Nuclear Information System (INIS)

    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

  4. A study on the mitigating capability of an auxiliary feedwater system during SBO for APR1400

    International Nuclear Information System (INIS)

    The objective of this paper is to establish an auxiliary feedwater (AFW) operational technical bases for the Korean Next Generation Reactor (APR1400) by modeling the plant, and by analyzing station blackout (SBO) using the MELCOR code. For the integrity of the reactor vessel and containment safety against severe accidents, it is essential to understand the severe accident sequences and to assess accident progression accurately using computer codes. Furthermore, it is important to attain the capability to analyze the advanced nuclear reactor design for the severe accident prevention and mitigation. Accident analyses are also undertaken to find out how effective AFW is mitigating in severe accident progresses. A nominal base case for SBO without AFW, time interval between feedwater stop and reactor vessel failure is 12,740 seconds. When AFW operates to mitigate the SBO accident progression 2, 4 and 8 hours after SBO starts, the reactor vessel failure is delayed for 20,415 seconds, 22,633 seconds and 26,508 seconds, respectively thus the operator has more time available for AC recovery and accident mitigation to prevent reactor vessel failure. (author)

  5. Safety analysis and justification for modification of auxiliary feed-water system in Daya Bay Nuclear Power Plant

    International Nuclear Information System (INIS)

    The major feed-water line break accident is re-analyzed, which is based on Guangdong Daya Bay nuclear power station final safety analysis report, to justify the impacts of the decreasing of auxiliary feed-water flow rate on the safety margin in Daya Bay. The results showed that the accident analysis can meet the demands of acceptance criteria with the auxiliary feed-water flowrate decreasing from 45 m3/h to 41.8 m3/h, and enough safety margin is still retained

  6. Experimental study of the start-up transient effect on cooling performance of the PAFS (Passive Auxiliary Feedwater System)

    International Nuclear Information System (INIS)

    PAFS (Passive Auxiliary Feedwater System) is a passive cooling system on the secondary system of APR+ (Advanced Power Reactor Plus). It can replace the conventional active cooling system for auxiliary feedwater injection to the steam generator by a passive way, and it cools down the secondary system of the steam generator by heat transfer at the condensation heat exchanger installed in the PCCT (Passive Condensation Cooling Tank). To validate a cooling performance of PAFS, a separate effect test loop has been constructed at KAERI (Korea Atomic Energy Research Institute), which is named PASCAL (PAFS Condensing heat removal Assessment Loop). It simulates a single tube of the horizontal heat exchanger, which is equivalent to 1/240 of the prototype according to a volumetric scaling methodology. In this study, two-phase flow phenomena in a horizontal heat exchanger and PCCT (Passive Condensate Cooling Tank) for the facility were experimentally investigated and the cooling capability of the condensation heat exchanger was validated in the initial start-up transient state. (author)

  7. Modifications and addition of an auxiliary feedwater pump to SONGS [San Onofre nuclear generating station] 1

    International Nuclear Information System (INIS)

    This paper discusses the recent modifications and equipment additions to the auxiliary feedwater (AFW) system at San Onofre nuclear generating station (SONGS) Unit 1. As a result of the Three Mile Island accident, several modifications to the AFW system were required. The addition of a third auxiliary feedwater pump was necessary to allow auxiliary feedwater operation following a main steam-line break concurrent with a single active failure

  8. Experimental Study on the PAFS (Passive Auxiliary Feedwater System) during the Quasisteady State and the MSSV Open

    International Nuclear Information System (INIS)

    APR+ (Advanced Power Reactor Plus) is a GEN-III+ nuclear power plant being developed in Korea. The PAFS (Passive Auxiliary Feedwater System) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of the PAFS, the experimental program of the separate effect test is in progress at KAERI (Korea Atomic Energy Research Institute). The test facility, PASCAL (PAFS Condensing heat removal Assessment Loop) was constructed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. In this study, six tests were performed for validating cooling performance of the PAFS during a quasi-steady state. With a given thermal power of electrical heaters in the steam generator from 200 kW to 750 kW (SS-200-P1, SS-300-P1, SS-400-P1, SS-650, and SS-750-P1), a heat removal rate in the PCHX was measured and the characteristics of the natural convection in the loop were investigated. In the test of MSSV open, the thermal hydraulic behavior in the system was investigated after an abrupt open and close of the MSSV

  9. Experimental Study on the PAFS (Passive Auxiliary Feedwater System) during the Quasisteady State and the MSSV Open

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byoung Uhn; Kim, Seok; Park, Yu Sun; Kim, Bok Deuk; Kang, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    APR+ (Advanced Power Reactor Plus) is a GEN-III+ nuclear power plant being developed in Korea. The PAFS (Passive Auxiliary Feedwater System) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of the PAFS, the experimental program of the separate effect test is in progress at KAERI (Korea Atomic Energy Research Institute). The test facility, PASCAL (PAFS Condensing heat removal Assessment Loop) was constructed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. In this study, six tests were performed for validating cooling performance of the PAFS during a quasi-steady state. With a given thermal power of electrical heaters in the steam generator from 200 kW to 750 kW (SS-200-P1, SS-300-P1, SS-400-P1, SS-650, and SS-750-P1), a heat removal rate in the PCHX was measured and the characteristics of the natural convection in the loop were investigated. In the test of MSSV open, the thermal hydraulic behavior in the system was investigated after an abrupt open and close of the MSSV

  10. Quantitative common cause failure modeling for auxiliary feedwater system involving the seismic-induced degradation of flood barriers

    International Nuclear Information System (INIS)

    Flood barriers are important defenses which will reduce the internal flood-induced failure risk of safety-related equipment in the turbine building. Contrarily, the degradation of flood barriers will increase the risk of internal flood-induced common cause failure (CCF). Two layouts of auxiliary feedwater pumps system are compared to demonstrate the quantitative risk assessment of the possible degradation of flood barriers. The alpha decomposition method has been developed by the authors in order to quantitatively evaluate the CCF parameters based on the causal inference. Occurrence frequency and CCF triggering ability are two important elements which will decide the CCF risk significance of potential common causes. The seismic-induced internal flood combining with the degradation of flood barriers is analyzed. The degradation of flood barriers is treated as a stochastic process and a Markov model is applied to consider the time-dependent states. The failure time of three auxiliary feedwater pumps is calculated based on the water flow rate through flood barriers. CCF triggering abilities of internal floods are calculated which are represented as decomposed alpha factors. This article shows the updating process of CCF parameters according to Bayesian inference and hypothetical databases. It is concluded that the issue of CCF modeling is not only decided by the number of redundant components but also decided by causes and plant-specific design. (author)

  11. Application of reliability-centered maintenance to the auxiliary feedwater system at San Onofre Nuclear Generating Station

    International Nuclear Information System (INIS)

    Reliability-centered maintenance (RCM) is a systematic methodology for defining applicable and effective preventive maintenance (PM) tasks. In 1984, the Electric Power Research Institute (EPRI) studied the air transport industry's RCM program as a candidate for technology transfer to the nuclear power industry. EPRI initiated two RCM pilot projects that directly utilized the RCM methodology developed by the aviation industry. The first RCM application was to the component cooling water systems of Florida Power and Light's Turkey Point Units 3 and 4. The second application was to the main feedwater system at Duke Power's McGuire Station. The results of these studies clearly indicate the benefits of the system-oriented RCM approach, and many areas for cost-effective improvements to PM programs were identified. After the completion of these two pilot studies, Southern California Edison and EPRI initiated an application of RCM to the auxiliary feedwater (AFW) system at the San Onofre Nuclear Generating Station, Unit 2. In contrast to the previous EPRI-sponsored applications to normally operating systems, the AFW system is a standby safety system. The study results demonstrate the usefulness of extending the RCM methodology to standby safety systems. The specific results show promise of reducing the PM costs for the AFW system at San Onofre while maintaining highly reliable system performance. The recommendations from this study are currently being considered for implementation by the plant maintenance staff

  12. Passive Condensation Cooling Tank (PCCT) Water Level Effect for Cooling Performance of Passive Auxiliary Feedwater System (PAFS)

    International Nuclear Information System (INIS)

    APR+ (Advanced Power Reactor Plus) is a next generation nuclear power plant being developed in Korea. It adopts PAFS (Passive Auxiliary Feedwater System) for the steam generator (SG) instead of an active auxiliary feedwater system for the conventional nuclear power plant (NPP). The passive safety system is advantageous in that it can enhance the reliability and reduce the effect of operator mistakes, which have been fundamental weak points as indicated in the safety analysis including the PSA (Probability Safety Assessment). The PAFS can replace the conventional active auxiliary feedwater system for the SG by a passive way. A schematic diagram of the PAFS for the APR+ is shown in Figure 1. It is composed of a steam-supply line, a condensation heat exchanger, a return-water line, and a PCCT (Passive Condensate Cooling Tank). When the water level in the SG becomes lower than 25% of the wide range of the water level transmitter during an accident situation, the actuation valve at the return water line is open and then the natural convection flow of the PAFS can be made. It cools down the secondary system of the SG by heat transfer at the condensation heat exchanger installed in the PCCT. The steam generated from the SG in the high pressure condition is condensed in the condensation heat exchanger tube. The absolute pressure at the top of PCCT is maintained at an atmospheric pressure, so that natural convection accompanying boiling heat transfer at the outside wall of the heat exchanger tubes occurs in the PCCT pool side. Since the heat exchanger and the PCCT are located at a higher elevation than the SG, condensate water can be returned back to the SG with a natural driving force. From the experiment, two-phase flow phenomena in the horizontal heat exchanger and PCCT were investigated and the cooling capability of the condensation heat exchanger was validated. Test results showed that the design of the condensation heat exchanger in PAFS could satisfy the requirement

  13. Passive Condensation Cooling Tank (PCCT) Water Level Effect for Cooling Performance of Passive Auxiliary Feedwater System (PAFS)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok; Bae, Byoung Uhn; Cho, Yun Je; Kim, Bok Deuk; Kang, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Yun, Byong Jo [Pusan National University, Busan (Korea, Republic of)

    2011-10-15

    APR+ (Advanced Power Reactor Plus) is a next generation nuclear power plant being developed in Korea. It adopts PAFS (Passive Auxiliary Feedwater System) for the steam generator (SG) instead of an active auxiliary feedwater system for the conventional nuclear power plant (NPP). The passive safety system is advantageous in that it can enhance the reliability and reduce the effect of operator mistakes, which have been fundamental weak points as indicated in the safety analysis including the PSA (Probability Safety Assessment). The PAFS can replace the conventional active auxiliary feedwater system for the SG by a passive way. A schematic diagram of the PAFS for the APR+ is shown in Figure 1. It is composed of a steam-supply line, a condensation heat exchanger, a return-water line, and a PCCT (Passive Condensate Cooling Tank). When the water level in the SG becomes lower than 25% of the wide range of the water level transmitter during an accident situation, the actuation valve at the return water line is open and then the natural convection flow of the PAFS can be made. It cools down the secondary system of the SG by heat transfer at the condensation heat exchanger installed in the PCCT. The steam generated from the SG in the high pressure condition is condensed in the condensation heat exchanger tube. The absolute pressure at the top of PCCT is maintained at an atmospheric pressure, so that natural convection accompanying boiling heat transfer at the outside wall of the heat exchanger tubes occurs in the PCCT pool side. Since the heat exchanger and the PCCT are located at a higher elevation than the SG, condensate water can be returned back to the SG with a natural driving force. From the experiment, two-phase flow phenomena in the horizontal heat exchanger and PCCT were investigated and the cooling capability of the condensation heat exchanger was validated. Test results showed that the design of the condensation heat exchanger in PAFS could satisfy the requirement

  14. Development of the Phenomena Identification Ranking Table (PIRT) for the Passive Auxiliary Feedwater System (PAFS) of the APR+

    International Nuclear Information System (INIS)

    The APR+ (Advanced Power Reactor plus) is a Gen- III+ pressurized water reactor (PWR) of which the standard design is currently being developed in Korea. This reactor adopts new design features which are believed to contribute not only to enhancement in nuclear safety but also to improvement in economic competitiveness. While the conventional nuclear power plants have utilized the active cooling systems, the APR+ adopts two types of passive safety features; an advanced fluidic device (FD+) and a passive auxiliary feedwater system (PAFS). The PAFS is one of the passive cooling systems of the APR+ which can replace an active system for auxiliary feedwater injection to a steam generator. A schematic diagram of the PAFS is shown in Fig. 1. It cools down the secondary system by heat transfer at horizontal heat exchangers in a PCCT (Passive Condensation Cooling Tank). High pressure steam flow from the steam generator is condensed in the horizontal heat exchanger, and the water in the PCCT pool is evaporated by a boiling heat transfer at the outside wall of the heat exchanger. With an aim of validating the cooling and operational performance of the PAFS, a separate effect test, PASCAL (PAFS Condensing heat removal Assessment Loop) is being performed at KAERI (Korea Atomic Energy Research Institute). In this study, Phenomena Identification and Ranking Table (PIRT) has been developed for identifying the major parameters affecting the thermal-hydraulic phenomena which originate from the adoption of the PAFS in the APR+. The PIRT process can be widely used to improve a safety analysis code for a new application and to establish experimental programs and to support the resolution of the licensing issues. The PIRT process used in this study follows the methodology previously applied in the APR1400 (Advanced Power Reactor 1400 MWe) PIRTs for large break loss of coolant accident (LBLOCA) and direct vessel injection (DVI) line break events

  15. Development of the Phenomena Identification Ranking Table (PIRT) for the Passive Auxiliary Feedwater System (PAFS) of the APR+

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Chung, Bub Dong; Kang, Kyoung Ho; Kang, Han Ok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Yun, Byong Jo [Pusan National University, Busan (Korea, Republic of); Bang, Young Suk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Seong, Ho Je [KEPCO E and C, Yongin (Korea, Republic of); Hong, Soon Joon [FNC Technology Co. Ltd., Seoul (Korea, Republic of); Sim, Suk Ku [EN2t Inc., Daejeon (Korea, Republic of)

    2012-05-15

    The APR+ (Advanced Power Reactor plus) is a Gen- III+ pressurized water reactor (PWR) of which the standard design is currently being developed in Korea. This reactor adopts new design features which are believed to contribute not only to enhancement in nuclear safety but also to improvement in economic competitiveness. While the conventional nuclear power plants have utilized the active cooling systems, the APR+ adopts two types of passive safety features; an advanced fluidic device (FD+) and a passive auxiliary feedwater system (PAFS). The PAFS is one of the passive cooling systems of the APR+ which can replace an active system for auxiliary feedwater injection to a steam generator. A schematic diagram of the PAFS is shown in Fig. 1. It cools down the secondary system by heat transfer at horizontal heat exchangers in a PCCT (Passive Condensation Cooling Tank). High pressure steam flow from the steam generator is condensed in the horizontal heat exchanger, and the water in the PCCT pool is evaporated by a boiling heat transfer at the outside wall of the heat exchanger. With an aim of validating the cooling and operational performance of the PAFS, a separate effect test, PASCAL (PAFS Condensing heat removal Assessment Loop) is being performed at KAERI (Korea Atomic Energy Research Institute). In this study, Phenomena Identification and Ranking Table (PIRT) has been developed for identifying the major parameters affecting the thermal-hydraulic phenomena which originate from the adoption of the PAFS in the APR+. The PIRT process can be widely used to improve a safety analysis code for a new application and to establish experimental programs and to support the resolution of the licensing issues. The PIRT process used in this study follows the methodology previously applied in the APR1400 (Advanced Power Reactor 1400 MWe) PIRTs for large break loss of coolant accident (LBLOCA) and direct vessel injection (DVI) line break events

  16. Evaluation of Effect of N{sub 2} Gas on the Cooling Capability of Passive Auxiliary Feedwater System (PAFS) in APR+

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yun Je; Kang, Kyong Ho; Yun, Byong Jo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    In Korea, Advanced Power Reactor Plus (APR+) has being developed by adding passive safety features to Advanced Power Reactor 1400MWe (APR1400). Passive Auxiliary Feedwater System (PAFS) is one of passive system adopted in the APR+ to replace the conventional active auxiliary feedwater system. Because PAFS removes decay heat from the reactor core, it is required to verify the performance of PAFS in postulated accidents cases. In addition, an effect of noncondensable gas on the heat removal capability of PAFS should be evaluated since the non-condensable gas may deteriorate a condensation heat transfer through the condensation heat exchanger in PAFS. In this study, the effect of N{sub 2} gas was evaluated using MARS

  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)

    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 condensation heat transfer model for nearly horizontal tubes of the Passive Auxiliary Feedwater System in APR+

    International Nuclear Information System (INIS)

    A new condensation heat transfer model based on the flow regime in the nearly horizontal tube has been developed for the Passive Auxiliary Feed-water System (PAFS) of Korean Advanced Power Reactor Plus (APR+). This study focused on the stratified flow in horizontal tubes in which two different heat transfer mechanisms are involved. The void fraction was determined from the 1-D separated flow model (SFM) which incorporates closure relations for shear stress defined by single-phase based expressions and geometric relations for a concave interface using the eccentric circles. The wetted angle proposed by Hart's correlation (1989) was used to classify flow regimes into annular, stratified-wavy and stratified-smooth flow. The new film condensation heat transfer correlation based on Nusselt's integral analysis (1916) was proposed to predict the heat transfer coefficient affected by the vapor flow on the upper portion of tube in the stratified flow. Furthermore, the convective heat transfer correlation for single-phase heat transfer was used to predict the heat transfer coefficient for condensate flowing on the entire perimeter of annular flow and the bottom of the stratified flow. Both heat transfer correlations use Reynolds number based on the phasic actual velocities and geometric variables obtained from SFM. Finally, the new condensation heat transfer model package was evaluated against available experimental data for water and it showed good results. (author)

  19. Aging assessment of auxiliary feedwater pumps

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  1. Preliminary simulation of the PAF (Passive Auxiliary Feedwater System) Using The Cupid Code

    International Nuclear Information System (INIS)

    After the Fukushima accident, an emphasis has been given to the implementation of an inherent passive safety system of a nuclear reactor. The PAFS is one of the advanced safety features applied to the APR+ (Advanced Power Reactor +) of Korea aiming to change the conventional active safety system into passive one. The PAFS is consisted of two cooling systems, PCHX (Passive Condensation Heat Exchanger) and PCCT (Passive Condensation Cooling Tank). In this research, the PCCT is independently simulated using the CUPID code, in which a natural circulation happens. The PCCT is modelled using a two-dimensional area and the sub-structures inside the tank are modelled using a porous medium. For the validation of simulations, the collapsed water level, the natural circulation velocities, and the liquid temperature are investigated quantitatively. The results show the simulated natural circulation using the CUPID code coinciding well with the experimental results

  2. Simulation of Transient Scenarios for Passive Auxiliary Feedwater System in APR+

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byounguhn; Kim, Seok; Park, Yusun; Kang, Kyungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    In this study, postulated transient scenarios occurring in the PAFS were simulated to evaluate the operational performance of system and investigate the thermal hydraulic phenomena of the two-phase natural convection flow. The transient tests simulated in this study are PAFS start-up actuation test (SU) and non-condensable gas effect test (NC). In this study, postulated transient cenarios occurring in the PAFS were simulated to evaluate the performance of the condensation heat transfer and investigate the thermal hydraulic phenomena of the two-phase natural convection flow. Start-up actuation test simulated the initial transient when the PAFS actuation signal was generated and the natural convection flow was initiated in the loop, and any significant two-phase flow instability was not observed in the test. The purpose of the non-condensable gas effect test is to study the characteristics of the condensation heat transfer in the heat exchanger when the nitrogen gas was injected. The test results proved that the existence of the non-condensable gas up did not produce a meaningful decrease of the cooling capability in the PAFS. From the experimental results described above, the cooling and operating performance of the PAFS was validated with respect to occurrence of the various transient scenarios and it was proved that the function of the PAFS can be effectively performed during the transient situation. The result will be also utilized in validation of the thermal hydraulic system code in the future.

  3. Simulation of Transient Scenarios for Passive Auxiliary Feedwater System in APR+

    International Nuclear Information System (INIS)

    In this study, postulated transient scenarios occurring in the PAFS were simulated to evaluate the operational performance of system and investigate the thermal hydraulic phenomena of the two-phase natural convection flow. The transient tests simulated in this study are PAFS start-up actuation test (SU) and non-condensable gas effect test (NC). In this study, postulated transient cenarios occurring in the PAFS were simulated to evaluate the performance of the condensation heat transfer and investigate the thermal hydraulic phenomena of the two-phase natural convection flow. Start-up actuation test simulated the initial transient when the PAFS actuation signal was generated and the natural convection flow was initiated in the loop, and any significant two-phase flow instability was not observed in the test. The purpose of the non-condensable gas effect test is to study the characteristics of the condensation heat transfer in the heat exchanger when the nitrogen gas was injected. The test results proved that the existence of the non-condensable gas up did not produce a meaningful decrease of the cooling capability in the PAFS. From the experimental results described above, the cooling and operating performance of the PAFS was validated with respect to occurrence of the various transient scenarios and it was proved that the function of the PAFS can be effectively performed during the transient situation. The result will be also utilized in validation of the thermal hydraulic system code in the future

  4. Heat Structure Coupling of CUPID and MARS for the Passive Auxiliary Feedwater System Analysis

    International Nuclear Information System (INIS)

    The two-phase phenomena in the steam supply system including the condensation in the Passive Condensate Heat Exchanger (PCHX) were calculated by MARS and those in the Passive Condensate Cooling Tank (PCCT) including the natural circulation and the boil-off were modeled by CUPID. This paper presents the coupling method and the simulation results using the coupled codes. In the present study, the multi-scale thermal-hydraulic analysis method using the coupled MARS-CUPID code was applied for the simulation of the passive condensation cooling phenomena. The primary side of the PASCAL test facility including the PCHX was simulated by MARS and the secondary side, the PCCT, was modeled by the CUPID. It was found that the overall two-phase behaviors inside the water pool and the condensation heat transfer inside the heat exchanger were qualitatively well reproduced with the coupled code. Comparison of various parameters between the test and the simulation will be performed in the future for a quantitative analysis

  5. Heat Structure Coupling of CUPID and MARS for the Passive Auxiliary Feedwater System Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Hyoungkyu [Seoul National Univ., Seoul (Korea, Republic of); Cho, Yunje; Lee, Seungjun; Yoon, Hanyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The two-phase phenomena in the steam supply system including the condensation in the Passive Condensate Heat Exchanger (PCHX) were calculated by MARS and those in the Passive Condensate Cooling Tank (PCCT) including the natural circulation and the boil-off were modeled by CUPID. This paper presents the coupling method and the simulation results using the coupled codes. In the present study, the multi-scale thermal-hydraulic analysis method using the coupled MARS-CUPID code was applied for the simulation of the passive condensation cooling phenomena. The primary side of the PASCAL test facility including the PCHX was simulated by MARS and the secondary side, the PCCT, was modeled by the CUPID. It was found that the overall two-phase behaviors inside the water pool and the condensation heat transfer inside the heat exchanger were qualitatively well reproduced with the coupled code. Comparison of various parameters between the test and the simulation will be performed in the future for a quantitative analysis.

  6. Analysis of PAFS (Passive Auxiliary Feedwater System) horizontal heat exchanger in APR+ and the scale-up capability of experimental loop

    International Nuclear Information System (INIS)

    APR+ (Advanced Power Reactor Plus) is the next generation nuclear power plant in Korea. It adopts PAFS (Passive Auxiliary Feedwater System) on the secondary system. It can replace the conventional active system for auxiliary feedwater injection to the steam generator, and it enables the coolant to be supplied by a passive system. It cools down the secondary system by heat transfer at a horizontal U-tube in PCCT (Passive Condensation Cooling Tank). High pressure steam flow from the steam generator is condensed in the horizontal heat exchanger. The water in PCCT is maintained at an atmospheric pressure, so that boiling heat transfer at the outside wall of heat exchanger and natural convection occur in PCCT pool. The heat exchanger and PCCT is higher than the steam generator, so condensate can be drained and injected to feedwater system without any active system. This study aims at analyzing the heat removal capacity for the design of the horizontal heat exchanger in PAFS. To design the condensation heat exchanger in PAFS, and the two-phase flow phenomena in horizontal U-tube and were investigated by MARS (Multi-dimensional Analysis for Reactor Safety, a thermal hydraulic system analysis code) calculation. By benchmarking with NOKO experimental result, MARS code showed a reasonable capability to quantitatively predict the condensation in horizontal tube heat exchanger. For the design of PAFS heat exchanger in APR+, the calculation results proved to sufficiently remove the decay heat of 138 MW in total by the condensation heat transfer without any active auxiliary feedwater system during TLOFW (Total Loss of Feed Water) accident. In the analysis, the distribution of thermal equilibrium quality and local liquid fraction in the horizontal U-tube was also investigated. In order to experimentally investigate the condensation phenomena and natural convection in PAFS, a test loop with a single horizontal U-tube and PCCT is under construction at KAERI (Korea Atomic Energy

  7. Development of a Prototypical Condensation Model for the Nearly Horizontal Heat Exchanger Tube of the APR+ PAFS(Passive Auxiliary Feed-water System)

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Tae Hwan; Yun, Byong Jo; Jeong, Jae Jun [Pusan National Univ., Busan (Korea, Republic of)

    2013-10-15

    In this study, a new condensation heat transfer model was developed for the separated flow regime in the nearly horizontal tube. A new heat transfer model which takes into account of different heat transfer mechanisms occurred in the upper and lower regions of perimeters was developed in the nearly horizontal tube. The present model shows good prediction capability against experimental data. Passive Auxiliary Feed-water System(PAFS) which is adopted in the Korean advanced nuclear power plant, APR+, removes decay heat by condensing steam from the secondary side in the nearly horizontal tubes under the accident condition. The comparison of prediction against PASCAL(2011) experimental data obtained by KAERI indicated that the best estimated safety analysis code such as MARS tends to underestimate the condensation heat transfer coefficient.

  8. Development of a Prototypical Condensation Model for the Nearly Horizontal Heat Exchanger Tube of the APR+ PAFS(Passive Auxiliary Feed-water System)

    International Nuclear Information System (INIS)

    In this study, a new condensation heat transfer model was developed for the separated flow regime in the nearly horizontal tube. A new heat transfer model which takes into account of different heat transfer mechanisms occurred in the upper and lower regions of perimeters was developed in the nearly horizontal tube. The present model shows good prediction capability against experimental data. Passive Auxiliary Feed-water System(PAFS) which is adopted in the Korean advanced nuclear power plant, APR+, removes decay heat by condensing steam from the secondary side in the nearly horizontal tubes under the accident condition. The comparison of prediction against PASCAL(2011) experimental data obtained by KAERI indicated that the best estimated safety analysis code such as MARS tends to underestimate the condensation heat transfer coefficient

  9. Discussion on RELAP5 and RETRAN3D Modeling for Passive Condensate Cooling Tank of Passive Auxiliary Feedwater System in APR+

    International Nuclear Information System (INIS)

    Domestic nuclear industry has started the development of APR+ as a Korean specific reactor for the export strategy. In the development of APR+ a passive auxiliary feedwater system (PAFS) has been considered as a noticeable candidate of improved design. The outline of PAFS and passive condensate cooling tank (PCCT) containing horizontal heat exchanger is shown in Fig. 1. For the successful design of PAFS, performance analyses or safety analyses are prerequisite using best estimate thermal hydraulic codes such as RELAP5 or RETRAN3D. Because of the inherent features of RELAP5 or RETRAN3D, pool model and condensation in horizontal tube have not been well-setup nor widely studied. This paper discusses about the PCCT phenomena including steam condensation in horizontal tube and pool heat transfer, and RELAP5 and RETRAN3D modeling

  10. Auxiliary systems

    International Nuclear Information System (INIS)

    For a undisturbed reactor operation, the various Auxiliary and Ancillary Systems must function perfectly with the Reactor Coolant System together. While the Auxiliary Systems are directly connected to the Reactor Coolant System and therefore have contact with the Reactor Coolant, the Ancillary Systems perform tasks which do not directly influence reactor operation and in part are necessary exclusively for environment protection. The design criteria of the individual systems are a result of these tasks, especially in relation to availability, operational readiness and probability of failure. (orig.)

  11. 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)

  12. Feedwater system of nuclear power plant

    International Nuclear Information System (INIS)

    The present invention concerns a feedwater system of a BWR power plant and, especially, it relates to a heater drain system from a high pressure feedwater heater. That is, the feedwater system comprises (1) a feedwater pump for supplying condensates discharged from a turbine to the reactor as feedwater, (2) a feedwater heater for heating the feedwater by steams extracted from the turbine, (3) a drain reservoir for recovering the drain from the feedwater heater, (4) a drain pump for sending the drain in the drain reservoir to the upstream of the feedwater pump and (5) a drain pipeline. In this feedwater system, a drain cooling device is disposed between the drain reservoir and a drain pump, to cool the drain by the condensates in the upstream of the feedwater heater. With such a constitution, it is possible to lower the temperature of the drain on the suction side of the drain pump, which is at a temperature near the almost at a saturated temperature and tends to boil and generate bubbles. Accordingly, damages of pumps due to boiling of the drain (flush) can be prevented. (I.S.)

  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)

    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.1130F) for TMI-1 and approx.44 K (approx.800F) for Zion-1

  14. Auxiliary systems

    International Nuclear Information System (INIS)

    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.)

  15. Auxiliary feedwater spreading and countercurrent flow flooding in a model once-through steam generator: Final report

    International Nuclear Information System (INIS)

    Experiments were performed in a separate effects test facility to investigate auxiliary feedwater spreading and counter-current flow flooding in a once-through steam generator model. The test objectives were to investigate: (1) flooding at the tube support plate, (2) spreading of the auxiliary feedwater across the tubes, (3) tube wetting characteristics, and (4) droplet entrainment in the exit steam flow. Eighty tests were performed in three phases involving air-water and steam-water experiments. The tests covered a wide range of auxiliary feedwater flow, air/steam flow, primary side water temperature, and flow rate. A scaling analysis was performed, based on a falling film mathematical model, which guided the selection of test conditions and subsequent data analysis. Experimental measurements and visual observations showed that: (1) entrainment was negligible, (2) spreading of the auxiliary feedwater and tube wetting depend mainly on the auxiliary feedwater injection rate, and (3) evaporation at the tube support plate controls flooding. The analysis showed, and the experiments confirmed, that there are two scaling parameters which control counter-current flow flooding in the passages of the tube support plate. A flooding regime map was constructed which showed the boundaries of the no-flooding and flooding regions as a function of these two scaling parameters

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

    International Nuclear Information System (INIS)

    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

  17. Feedwater temperature control methods and systems

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Two-phase flow and heat transfer in a once-through steam generator during auxiliary feedwater injection

    International Nuclear Information System (INIS)

    In this paper, a phenomenological model is developed for the thermal-hydraulic processes on the secondary side of a once-through steam generator during auxiliary feedwater injection. Based on experimental observations, the flow of auxiliary feedwater in the secondary side is modeled as a turbulent falling film on the tubes, in direct contact with a countercurrent flow of steam, that receives heat from the primary side. Conservation equations for the falling film and steam on the secondary side, and for the primary-side coolant, are derived. Boiling in the falling film, evaporation and/or condensation at the falling film-gas interphase, and countercurrent flow limitation in the tube support plate passages are modeled. Numerical solution of the conservation equations provide the axial variation of flow rates and temperatures in the primary and secondary side

  19. CAREM-25. Auxiliary systems

    International Nuclear Information System (INIS)

    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)

  20. CAREM-25. Auxiliary systems

    International Nuclear Information System (INIS)

    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

  1. Operation auxiliary system (SAO)

    International Nuclear Information System (INIS)

    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)

  2. Modeling of transient thermal-hydraulic phenomena in a once-through steam generator during auxiliary feedwater injection

    International Nuclear Information System (INIS)

    A phenomenological model is developed for the transient thermal-hydraulic processes on the secondary side of a once-through steam generator during auxiliary feedwater (AFW) injection. Detailed modeling of the thermal-hydraulic processes above the top tube support plate (TSP) is particularly emphasized. The nonuniform distribution of the AFW on the secondary-side tubes is represented by dividing the secondary side into a number of tube groups. For each tube group, the quasi-steady-state conservation equations representing the flow of a falling liquid film and steam on the secondary side and the primary coolant on the primary side are numerically solved for each time step, thereby providing the axial variation of flow rates and temperatures in the primary and secondary sides. Modeled processes include cooling due to the impingement of the AFW jet on the tubes, the forced convection/boiling heat transfer at the liquid film-tube interface, evaporation and condensation at the film-gas interphase, countercurrent flow limitation in the TSP passages, and the formation of a swollen two-phase pool above the top TSP. The aforementioned model for the thermal-hydraulic phenomena above the top TSP is incorporated into a transient model for the entire steam generator where the secondary side is divided into four regions. Global conservation equations representing the transient behavior of each region are numerically solved. Model predictions are compared with a typical test from the Multiloop Integral System Test experiments. Parametric and sensitivity calculations are also reported

  3. Cernavoda NPP feedwater system transient analysis using MMS package

    International Nuclear Information System (INIS)

    In this analysis we studied thoroughly the operation of the Feedwater System with purpose of preparing of a simulator as complete as possible for Candu 600 nuclear power plants, particularly for Cernavoda NPP. We made a complete simulation for the feedwater system because we intended to study all functional regimes of the system. We considered as boundaries of analyses the following: - Main Condenser System - deaerator; - Feedwater nozzle from Steam Generator; - Outlet nozzles from steam turbine to regeneration preheaters; - Secondary Drain thermal drain cycle. The computational codes used in this analysis are PIPENET and MMS. The following types of analyses were carried out: - hydraulic analysis for stationary regimes - 100% MCR (Main Continuous Rate), 80% MCR, 60%MCR with 2 Main Feedwater Pumps working, 60%MCR with 1 Main Feedwater Pump, 40%MCR. Those analyses were done using hydraulic package called PIPENET; - thermal hydraulic analyses; - stationary regime of 100%MCR using as starting point data resulted form PIPENET; - transient regimes starting from nominal values regime and various boundaries conditions. Transient regimes analyzed in this work are: all normal functional regimes of Feedwater System according with Design Manual; all abnormal regimes of Feedwater System without accident regimes. For all thermal hydraulic analyses we used MMS package. The results obtained with this MMS package were compared with data from heat-balance calculated of stationary regimes (100%, 80%, 60%, 40%) provided by General Electric - turbogenerator supplier

  4. Development and implementation of a BWR digital feedwater control system

    International Nuclear Information System (INIS)

    EPRI and Northern States Power Company (NSP) realized that fault-tolerant digital technology could improve Feedwater Control System reliability and operations. The Digital Feedwater Control System (DFCS) is the first major fault-tolerant digital control application in a nuclear power plant in the US. The microprocessor-based controller replaced the analog controller in the feedwater control loop to improve performance and reliability of control including ease of maintenance and spare parts supply. The DFCS at Monticello plant has been in operation since July 1986 without any failure of the control system. This system replaced and upgraded the main and start-up analog controllers at Monticello BWR. It features automatic control, on-line signal validation, controller self-diagnosis, and fault tolerance. The dual-redundant hardware configuration minimizes spare parts availability problems. At a control room panel, operators select each feedwater valve's operating mode (one or three element control, manual, and so on) or set bias inputs for individual feed-water-valve demand. These and other features permit more exact feedwater control system tuning, improving feedwater control in all modes of plant operation. Signal validation using parity-space techniques isolated failed sensors and permit system switching to accurate sensors, thus avoiding outages. To ensure successful operation of the system, extensive verification and validation effect were conducted. These included design reviews, factory acceptance testing using simulation code, site acceptance testing using full-scale plant simulator, and pre-operational and operational testing at Monticello plant

  5. 核电厂辅助给水系统控制方案设计研究%Study on Control System of Auxiliary Feedwater System of Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    鲁超

    2015-01-01

    Auxiliary water supply system is an important system in the design of safety facilities of nuclear power station. The common cause fault of software and the power plant station blackout are two key factors for control system of ASG. So in the design of the control system, in order to deal with the two failure factors, the control strategy of diversity and emergency power supply are respectively adopted. Through analysis diversity control strategy is an effect method to prevent the failure of the safety functions resulting from software common cause failure which meets single failure criterion. In the case of station blackout, it is necessary to provide emergency power supply for control system of the starting ASG, and ensure the system safety functions workable.%辅助给水系统(ASG)是核电厂专设安全设施中重要的系统。对于实现对ASG功能控制的安全级系统,软件共因故障和全厂失电是导致控制失效的两个关键因素。因此,在进行控制系统设计时,为应对这两大失效因素,分别采用了多样性和增加应急电源的控制策略。通过分析,采用多样性控制策略可以有效地防止软件共因故障导致安全功能丧失的风险,保证系统满足单一故障的要求。同时在全厂断电的情况下,增加应急供电电源,对启动ASG功能的控制系统进行紧急供电,保证系统安全功能可执行。

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

    International Nuclear Information System (INIS)

    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

  7. Microprocessor system for low power feedwater control

    International Nuclear Information System (INIS)

    The results of an ongoing Combustion Engineering development program to improve steam generator level control during low power operation are presented. A discussion is presented on the analytical tools, the verification process and the simulation capabilities that are currently available for the design and evaluation of advanced steam generator control systems. A review is then presented on the dynamic processes internal to the steam generator that must be controlled, the conceptual approach used in arriving at a design, followed by the verification of the design using a detailed digital simulation of the process and the control system. Finally, a discussion is presented on the reasons for the selection of a microprocessor based systems over a conventional analog system and the potential improvements in reliability and flexibility

  8. Transient analysis passive emergency feedwater system of CPR1000

    International Nuclear Information System (INIS)

    The transient thermal hydraulic characteristics of CPR1000 were analyzed by using RELAP5/MOD 3.4 code to verify the capability of the passive emergency feedwater system (PEFWS) for accident mitigation under the condition of station blackout accident (SBO). The calculation results show that the PEFWS of CPR1000 can supply the water to steam generator immediately and remove the core residual heat after the SBO successfully, and it also shows that the design of the PEFWS of CPR1000 is successful. (authors)

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

    International Nuclear Information System (INIS)

    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

  10. Switching mode simulation of parallel pumps feedwater system

    International Nuclear Information System (INIS)

    In this paper, a model of Parallel Pumps Feedwater System (PPFS) is established, which is verified the validity via experiments. Then simulating study of three different switching modes is presented for PPFS during the alternate startup process. The first switching mode is starting the pump in standby before shutting down the pump in operation, the second one is starting the pump in standby after shutting down the pump in operation, and the third one is operating two pumps at the same time. The simulation results show that the loss of water in the first switching mode is the smallest and the system flow oscillation in the third mode is the smallest. Moreover, in the same switching mode, the shorter the time between the pumps switching is, the smaller system flow fluctuation is and the shorter the time is needed to recover stability. (authors)

  11. Loss of Normal Feedwater analyses for Krsko Full Scope Simulator verification

    International Nuclear Information System (INIS)

    The purpose of these analyses was to perform calculations of a Loss of Normal (Main) Feedwater transient for Krsko NPP. The results of calculations were used for the verification of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. To perform the thermal-hydraulic analyses, the RELAP5/MOD2 computer code and the NPP Krsko input card deck were used. In the presented paper two scenarios have been analyzed. Both of them started with a loss of normal feedwater event. Thus, a reduction or an interruption of the heat removal by the secondary system occurred. The first scenario assumed that auxiliary feedwater was available during the transient, while in the second scenario both normal and auxiliary feedwater were unavailable. The results showed that with auxiliary feedwater pumps unavailable additional operator actions would be needed to prevent overheating of the core. (author)

  12. Engine Auxiliary System Guideline: Cooling Systems

    OpenAIRE

    Kela, Suvi

    2015-01-01

    The thesis was done for Wärtsilä Technical Services organization. The assignment was to consolidate a guideline for cooling systems as an engine auxiliary system covering the Wärtsilä 4-stroke engines currently in production. The guideline was to include information considering both marine and power plants installations. The sources of information were internal documentation from Wärtsilä, literature review and discussions with Wärtsilä cooling system experts. The guideline includes informati...

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  15. Study on the PWR Steam Generator Behavior with improved steam-driven aux feedwater system under prolonged SBO Accident

    International Nuclear Information System (INIS)

    The only available passive decay heat removal system of current PWRs is a turbine-driven auxiliary steam generator (SG) feedwater (TD-AFW) system. If a SG water level becomes too high, however, turbine blades could be damaged due to a large amount of becomes too high, however, turbine blades could be damaged due to a large amount of moisture in steam and the SG cooling capability would not be maintained any longer. Therefore, the SG water level should be controlled to prevent the turbine from being damaged during a Station Black-Out (SBO) accident. In this paper, an improved design feature is proposed to provide electric power for controlling SG water level when both off-site power and the emergency diesel generators are not available. There are additional SG level gauges and valve controllers to control the steam flow into the auxiliary turbine in an improved TD-AFW system. Electric power for this control system is provided by a small additional generator which is connected to the existing auxiliary turbine shaft. Using this new feature, decay heat cooling is available for 29 hours with only 1 condensate storage tank which is the water source of the AFW. Eventually, it is concluded that the improved TD-AFW system with an additional SG level controller and generator can avoid an early SG full level and continue long term cooling during a prolonged SBO accident

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

    International Nuclear Information System (INIS)

    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)

  17. Accurate feedwater iron control for dose rate reduction by advanced resin cleaning system in Tokai-2

    International Nuclear Information System (INIS)

    Dose rate reduction of out-of-core piping is one of main issues in Boiling Water Nuclear Power Plant (BWR). Main source of the out-of-core piping dose rate is 60Co which adhered to the piping and it is influenced by feedwater iron concentration. A relationship between feedwater iron concentration and amount of iron and cobalt, 60Co which deposited on fuel surface had been evaluated at Tokai-2 (1,100 MWe BWR, operated by The Japan Atomic Power Company, commercial operation started on 1978). As the results, it was demonstrated that to keep the amount of deposited iron on fuel surface around 2000μg/cm2 to reduce Co radioactivation. And, when feedwater iron concentration is around 0.5 ppb, that was achieved. But, when feedwater iron becomes less than 0.5 ppb, soluble 60Co concentration in reactor coolant increases and that makes out-of-core piping dose rate increase. So, necessity to control feedwater iron is shown from these behaviors. At Tokai-2, condensate water iron is removed by only condensate demineralizer resin, because Tokai-2 has no condensate filter. That is, iron removal performance of condensate demineralizer resin affects feedwater iron concentration directly. And, iron removal performance of condensate demineralizer resin is caused by resin cleanness. The resin has been cleaned by a resin cleaning method named 'backwash'. But iron on the surface of the resin could not be removed efficiently by the backwash. As the result, feedwater iron could not be reduced to 0.5 ppb. So, Advanced Resin Cleaning System (ARCS) which can remove almost the iron on the resin was retrofitted to Tokai-2, in October 2005 (21nd outage), to reduce feedwater iron. After applying ARCS, resin cleanness was improved, and feedwater iron decreased to around 0.5 ppb same as that of BWR plants with condensate filter. Also, feedwater iron concentration was maintained in around 0.5 ppb by changing frequency of resin cleaning. By using these results, an optimum control method of

  18. Analysis of the Leibstadt power plant condensate and feedwater systems during selected operational transients

    International Nuclear Information System (INIS)

    This paper presents a selection of plant analyses that were carried out by PSI in support of the Leibstadt Nuclear Power Plant (Swiss boiling water reactor). The analyses were performed as part of a collaboration between Leibstadt and PSI, to help resolve some operational problems that were experienced during the power uprate beginning in 1998. The issues under investigation were related to the behavior of the condensate and feedwater systems during transients initiated by a turbine trip, load rejection and a single feedwater pump trip, all of which increased the risk of an inadvertent reactor shutdown by reaching reactor pressure vessel water level limits. The possibility of a reactor shutdown was related to perturbations in the feedwater flow caused by transitory pump cavitation of the feedwater pumps, due to a rapid depressurization in the feedwater tank. In addition to a direct analysis of plant measurement provided by Leibstadt, steady-state and transient simulations of the events were performed at PSI using the system codes TRAC-BF1 and TRACE. Through a combination of the analysis of the plant measurements, the code simulations and an analysis of the whole plant behavior using the Leibstadt plant-simulator appropriate modifications of the plant hardware, control system and operational set points were proposed. The implementation and success of these changes were verified by a number of plant tests. Finally, the original designed plant capability not to shutdown during the aforementioned transients was demonstrated

  19. Reliability improvement of a three-element feedwater control system for a pressurized water reactor

    International Nuclear Information System (INIS)

    A study of the three-element feedwater flow control system design from the perspective of qualitative Failure Modes and Effects Analysis and quantitative Fault Tree Analyses has been made. The study quantified the impact of reliability on valve control instrumentation system components and logic including electrical/electronic and pneumatic control components and valve operators. The study resulted in recommendations for modifications in the feedwater control system, which should improve its reliability and, therefore, increase unit availability through a reduction in hardware related control system malfunctions. These recommendations were ranked by their cost and their cost-benefit in order to facilitate utility management selection of those recommendations to be implemented

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

    International Nuclear Information System (INIS)

    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)

    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.

  2. Improvement of Equipment reliability for Auxiliary Feed Water System

    International Nuclear Information System (INIS)

    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 CHR are not

  3. An Intelligent Auxiliary Vacuum Brake System

    OpenAIRE

    Tong, Chia-Chang; Lin, Jhih-Yu; Li, Shih-Fan; Li, Jiun-Yi

    2009-01-01

    The purpose of this paper focuses on designing an intelligent, compact, reliable, and robust auxiliary vacuum brake system (VBS) with Kalman filter and self-diagnosis scheme. All of the circuit elements in the designed system are integrated into one programmable system-on-chip (PSoC) with entire computational algorithms implemented by software. In this system, three main goals are achieved: (a) Kalman filter and hysteresis controller algorithms are employed within PSoC chip by software to sur...

  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)

    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. Network analysis of turbine and feedwater systems of the 'Fugen' nuclear power plant

    International Nuclear Information System (INIS)

    The present study describes the thermal-hydraulic network analysis of the turbine and feedwater systems of the 'Fugen' reactor. Turbines, feedwater heaters, and corresponding piping systems are modeled using the network calculation code NETFLOW++ and thermal-hydraulic conditions are calculated using the coupled numerical model. As a result of the calculation, distributions of important characteristics of the single-phase flow and two-phase flow in the piping such as pressure and void fraction are clarified. Flow patterns in the piping were investigated using the calculated result. It was found that the state of the coolant in the drainpipe changes from saturated liquid at the inlet to a two-phase flow with a large void fraction at the connection to the feedwater heaters. This is attributed to the pressure difference between the inlet and outlet of the drainpipes. Even the drainpipe from the moisture separator to the shell of feedwater heater no.4 shows a similar behavior, and the flow pattern changes from single phase to slug flow. The steam quality in the extraction line is very high, although a large number of droplets are contained in the flow. Contrary to expectation, these droplets do not completely evaporate in spite of the low-pressure conditions. (author)

  6. Water quality control device for condensation and feedwater system in reactor

    International Nuclear Information System (INIS)

    The object of the invention is to decrease the accumulation amount of ionic radioactivities in nuclear reactors, etc. thereby reducing the operators' exposure dose upon inspecting operation. The water quality control device for water condensating and feeding systems in a nuclear reactor according to this invention comprises a reservoir for storing cruds contained in back-washing water of condensate filters, a water quality detector disposed at the downstream of a high pressure feedwater pump for detecting Fe/Ni ratio in feedwater and a controller. If the Fe/Ni ratio determined by the water quality detector lowers to less than a predetermined value, cruds in the crud storing reservoir are mixed into feedwater. It is thus possible to maintain the Fe/Ni ratio in feedwater to higher than a predetermined level thereby promoting the effect of adsorbing and removing Fe-induced ionic activity thereby reducing the ionic activity concentration in reactor water. As a result, it is possible to decrease the accumulation amount of ionic radioactivity in the nuclear reactor and reduce the operators' exposure dose during inspecting operation. (K.M.)

  7. The feedwater control system at a PWR. Examples from Ringhals 3

    International Nuclear Information System (INIS)

    GSE Power Systems AB sponsored by SKI has performed a comprehensive investigation of the feedwater control system at Ringhals 3. The assignment is based on signal analysis of measurement data recorded in the plant during operation. Well-known problems in connection to automatic control of feedwater flow in a PWR are also presented in the report outside the area of signal analysis. Some simple feedback systems are analysed and evaluated with the aid of simulation in the introduction of the report. These simulated systems are a good background to understand the actual automatic control systems at Ringhals 3. Typical problems with feedwater control in foreign plants are: Non minimum phase for the water level dynamics in the steam exchanger, windup of the controller with integrating function, hysteresis, backlash or other types of non-linearities in the feedwater valve that can result in limit cycle oscillations and the method with feed forward to reduce the influence of disturbances via the steam flow. These types of problems are demonstrated with simulation and possible ways to reduce the influence are discussed in the report. The automatic control of feedwater flow is performed with three different control systems with different aims at Ringhals 3. The first system controls the feedwater valve with the aim to maintain agreement between the water reference value and water level in each steam generator. This system is called the water level controller. The second system controls the speed of the feedwater pumps with the aim to maintain 8 bar differential pressure over the feedwater valve for the valve out of three with lowest pressure drop. This is called the DP-control system. The third system compensates the speed of the feedwater pumps with the aim to eliminate difference in condenser levels between the two turbines. This is called condenser control system. The results prove that the DP-control acts fastest with a time constant equal 13 seconds. The step tests

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

    International Nuclear Information System (INIS)

    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. Impact of Feedwater Salinity on Energy Requirements of a Small-Scale Membrane Filtration System

    OpenAIRE

    Richards, Bryce S.; Masson, L; Schaefer, Andrea

    2009-01-01

    Many remote communities in both developed and developing countries lack electricity and clean drinking water. One solution, for such communities that rely on brackish groundwater, is a photovoltaic (PV) powered hybrid ultrafiltration (UF) / nanofiltration (NF) or reverse osmosis (RO) membrane filtration system. The system prototype described here can produce between 150 – 280 litres of clean water for each peak sunshine hour, depending on the salinity of the feedwater (1 – 5 g/...

  10. Modeling a high output marine steam generator feedwater control system which uses parallel turbine-driven feed pumps

    Institute of Scientific and Technical Information of China (English)

    QIU Zhi-qiang; ZOU Hai; SUN Jian-hua

    2008-01-01

    Parallel turbine-driven feedwater pumps are needed when ships travel at high speed. In order to study marine steam generator feedwater control systems which use parallel turbine-driven feed pumps,a mathematical model of marine steam generator feedwater control system was developed which includes mathematical models of two steam generators and parallel turbine-driven feed pumps as well as mathematical models of feedwater pipes and feed regulating valves. The operating condition points of the parallel turbine-driven feed pumps were calculated by the Chebyshev curve fit method. A water level controller for the steam generator and a rotary speed controller for the turbine-driven feed pumps were also included in the model. The accuracy of the mathematical models and their controllers was verified by comparing their results with those from a simulator.

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

    International Nuclear Information System (INIS)

    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)

  12. Auxiliary Heating Systems for the Ignitor Project

    Science.gov (United States)

    Sassi, M.; Mantovani, S.; Coppi, B.

    2013-10-01

    Auxiliary plasma heating systems directed at extending the range of plasma regimes that can be accessed by Ohmic heating only are important components of the Ignitor machine. In order to affect the entire plasma column an appropriate ICRH systemhas been designed and components of it have been tested. The adoption of a 280 GHz system affecting, by ECRH, the outer edge of the plasma column has been proposed in order to influence temperature and density profiles in this important region. The ICRH system will operate over the range 80-120 MHz, consistent with magnetic fields in the range 9-13 T. The maximum delivered power goes from 8 MW (at 80 MHz) to 6 MW (at 120 MHz) distributed over 4 ports. A full size prototype of the VTL between the port flange and the antenna straps, with the external support and precise guiding system has been constructed. The innovative quick latching system located at the end of the coaxial cable has been successfully tested, providing perfect interference with the spring Be-Cu electrical contacts. Vacuum levels of 10-6, compatible with the limit of material degassing, and electrical tests up to 12 kV without discharges have been obtained. Special attention was given to the finishing of the inox surfaces, and to the TIG welds. U.S. DOE sponsored.

  13. 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.)

  14. The German and English Auxiliary Systems and Complex Predicates

    Science.gov (United States)

    McCormick, Terrence C.

    1976-01-01

    This paper explores the auxiliary systems of English and German and the use of the auxiliary verbs in various complex predicate structures in the two languages. It aims at alleviating two types of problems in learning German involving governing patterns and ordering problems in clauses. (CHK)

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

    International Nuclear Information System (INIS)

    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.

  16. Auxiliary DCP data acquisition system. [airborne system

    Science.gov (United States)

    Snyder, R. V.

    1975-01-01

    An airborne DCP Data Aquisition System has been designed to augment the ERTS satellite data recovery system. The DCP's are data collection platforms located at pertinent sites. With the appropriate sensors they are able to collect, digitally encode and transmit environmental parameters to the ERTS satellite. The satellite in turn relays these transmissions to a ground station for processing. The satellite is available for such relay duty a minimum of two times in a 24-hour period. The equipment is to obtain continuous DCP data during periods of unusual environmental activity--storms, floods, etc. Two circumstances contributed to the decision to design such a system; (1) Wallops Station utilizes surveillance aircraft in support of rocket launches and also in support of earth resources activities; (2) the area in which Wallops is located, the Delaware and Chesapeake Bay areas, are fertile areas for DCP usage. Therefore, by developing an airborne DCP receiving station and installing it on aircraft more continuous DCP data can be provided from sites in the surrounding areas at relatively low cost.

  17. Vibration monitoring in Angra I nuclear power plant steam generator feedwater system

    International Nuclear Information System (INIS)

    The safety and reliability are the primary criteria in the design and operation of a nuclear power plant. However, due to the aging of the plant and its components, it is difficult to assure that what was originally built and qualified under strict standards is still guaranteed. In order to assure safety, reliability, availability and capacity, aging management through predictive maintenance techniques are being introduced in most plants around the world. In this present work, the monitoring of the vibrations signatures at the Angra I nuclear power plant steam generators feedwater systems main components such as the main feedwater pumps, pressure breaker blocks and the by pass valves, is presented. The vibration data was acquired, afterwards some major repairs were performed and during the startup commissioning procedures. Some of the major repairs performed are: changing of one pump shaft with balancing and alignment, replacement of the original bypass control valves by new disk stack type pneumatic control valves. The results show that no major vibrations anomaly is present after the maintenance indicating adequacy of the repairs made. The monitoring of the vibration in nuclear power plant components is being increasingly used as a tool for predictive maintenance. (author)

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

    International Nuclear Information System (INIS)

    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

  19. Radio frequency auxiliary heating systems design in ITER

    International Nuclear Information System (INIS)

    A combination of radio frequency (RF) auxiliary heating systems will provide at least one half of the required 100 MW of auxiliary power in ITER. Five of the 20 equatorial ports are assigned to RF heating systems. Recent work has focused on developing an integrated equatorial port-plug design concept for all of the RF auxiliary heating systems as well as other equatorial port systems such as diagnostics. Common features of the design approach include the use of identical interfaces to services such as cooling water, vacuum, mechanical connection to the vessel, and maintenance. Based on the integrated port concept, a high level of design integration has been achieved for the RF heating systems. Implementation of the integrated design concept has been accomplished without significantly affecting the individual system performance and with limited impact on the torus layout. (author)

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

    International Nuclear Information System (INIS)

    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)

  1. A Markov Model for Assessing the Reliability of a Digital Feedwater Control System

    Energy Technology Data Exchange (ETDEWEB)

    Chu,T.L.; Yue, M.; Martinez-Guridi, G.; Lehner, J.

    2009-02-11

    A Markov approach has been selected to represent and quantify the reliability model of a digital feedwater control system (DFWCS). The system state, i.e., whether a system fails or not, is determined by the status of the components that can be characterized by component failure modes. Starting from the system state that has no component failure, possible transitions out of it are all failure modes of all components in the system. Each additional component failure mode will formulate a different system state that may or may not be a system failure state. The Markov transition diagram is developed by strictly following the sequences of component failures (i.e., failure sequences) because the different orders of the same set of failures may affect the system in completely different ways. The formulation and quantification of the Markov model, together with the proposed FMEA (Failure Modes and Effects Analysis) approach, and the development of the supporting automated FMEA tool are considered the three major elements of a generic conceptual framework under which the reliability of digital systems can be assessed.

  2. A Markov Model for Assessing the Reliability of a Digital Feedwater Control System

    International Nuclear Information System (INIS)

    A Markov approach has been selected to represent and quantify the reliability model of a digital feedwater control system (DFWCS). The system state, i.e., whether a system fails or not, is determined by the status of the components that can be characterized by component failure modes. Starting from the system state that has no component failure, possible transitions out of it are all failure modes of all components in the system. Each additional component failure mode will formulate a different system state that may or may not be a system failure state. The Markov transition diagram is developed by strictly following the sequences of component failures (i.e., failure sequences) because the different orders of the same set of failures may affect the system in completely different ways. The formulation and quantification of the Markov model, together with the proposed FMEA (Failure Modes and Effects Analysis) approach, and the development of the supporting automated FMEA tool are considered the three major elements of a generic conceptual framework under which the reliability of digital systems can be assessed

  3. Auxiliary-Arc Electrodes for MHD Systems

    International Nuclear Information System (INIS)

    The important role of electrode phenomena in the operation of magneto aerodynamic machines is well known. In particular, the voltage drops which occur in the boundary layer in the immediate neighbourhood of the electrode may reduce the output of the apparatus. These voltage drops are caused partly by the increased resistance presented by the boundary layer in the neighbourhood of the electrode when the latter is appreciably colder than the gas, and partly by the fact that the electrode is not at a temperature sufficient to be emissive. Auxiliary-arc electrodes that have been constructed and tested seem to provide a solution both of the cold boundary layer problem and of the cathode emissivity problem. For this purpose an arc is established between a refractory metal cathode placed behind and clear of the generator wall and an anode forming part of the wall. The arc column can be activated by a rotational movement under the effect of a magnetic field, which may be that of the machine itself. The mechanical arrangement of the electrodes is such that, with a weak flow of gas (argon for example), it is possible to maintain a protective atmosphere around the arc cathode, while the arc anode is strongly cooled by the wall. The gas flow also has the effect of forcing the arc column towards the stream, thus increasing the conductivity of the boundary layer. Furthermore, the arc column behaves as a virtual cathode, from which a sizeable electron current can be extracted. Electrodes constructed on this principle have been tested on gas streams composed of fuel-oil combustion products. By using them as cathodes it has been possible to extract a current of 5 A without the voltage drop between the electrode and the gas exceeding 10 V. Comparative tests have been carried out with cooled metal electrodes, in which case the voltage drop is of the order of 120 V. The arc electrodes tested have operated for several hours without any apparent damage. In spite of the energy which has

  4. The auxiliary system design retrofits of the different coolant pump

    International Nuclear Information System (INIS)

    The coolant pump auxiliary systems retrofits are introduced in detail according to the different type of coolant pumps. The retrofit reasons of the chemical and volume control system, component cooling water system, Nuclear Nitrogen Storage and Distribution System, Vent and drain system, etc. are investigated. The most extraordinary change takes place in the chemical and volume control system and cooling water system. The charging flow temperature of re- generative heat exchanger and discharge flow of charging pump will be changed according to the difference coolant pump seal flow distribution. The commercial CFD software Flow master is employed to validate the charging capability. The other auxiliary systems' retrofits are also introduced in the end of this paper. (authors)

  5. Specifying the auxiliary heating system on TFCX

    International Nuclear Information System (INIS)

    This paper reviews the status of heating systems for the TFCX-S (all superconducting coil) and TFCX-H (hybrid coil) options. Three systems were defined; preheating (electron), current drive, and bulk (ion) heating. Application of systems engineering techniques facilitated fruitful discussions of requirements and their impact on equipment between physicists and engineers. A low-cost, flexible combination of systems allows plasma experiments using all rf startup and current drive

  6. 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)

    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)

  7. 20--500 watt AMTEC auxiliary electric power system

    Energy Technology Data Exchange (ETDEWEB)

    Ivanenok, J.F. III; Sievers, R.K. [Advanced Modular Power Systems, Inc., Ann Arbor, MI (United States)

    1996-12-31

    Numerous design studies have been completed on Alkali Metal Thermal to Electric Converter (AMTEC) power systems for space applications demonstrating their substantial increase in performance. Recently design studies have been initiated to couple AMTEC power conversion with fossil fueled combustion systems. This paper describes the results of a Phase 1 SBIR effort to design an innovative, efficient, reliable, long life AMTEC Auxiliary Electric Power System (AEPS) for remote site applications (20--500 watts). The concept uses high voltage AMTEC cells, each containing 7 to 9 small electrolyte tubes, integrated with a combustor and recuperator. These multi-tube AMTEC cells are low cost, reliable, long life static converters. AMTEC technology is ideal for auxiliary electric power supplies that must operate reliably over a broad range of temperatures, fuel sources, power levels, and operational specifications. The simplicity, efficiency (20% systems) and modularity of this technology allow it to fill applications as varied as light-weight backpacks, remote site power supplies, and military base power. Phase 1 demonstrated the feasibility of a 20% system design, and showed that the development needs to focus on identifying long life AMTEC cell components, determining the AMTEC cell and system reliability, and demonstrating that a 20 watt AMTEC system is 3--5 times more efficient than existing systems for the same application.

  8. Assembly auxiliary system for narrow cabins of spacecraft

    Science.gov (United States)

    Liu, Yi; Li, Shiqi; Wang, Junfeng

    2015-09-01

    Due to the narrow space and complex structure of spacecraft cabin, the existing asssembly systems can not well suit for the assembly process of cabin products. This paper aims to introduce an assembly auxiliary system for cabin products. A hierarchical-classification method is proposed to re-adjust the initial assembly relationship of cabin into a new hierarchical structure for efficient assembly planning. An improved ant colony algorithm based on three assembly principles is established for searching a optimizational assembly sequence of cabin parts. A mixed reality assembly environment is constructed with enhanced information to promote interaction efficiency of assembly training and guidance. Based on the machine vision technology, the inspection of left redundant objects and measurement of parts distance in inner cabin are efficiently performed. The proposed system has been applied to the assembly work of a spacecraft cabin with 107 parts, which includes cabin assembly planning, assembly training and assembly quality inspection. The application result indicates that the proposed system can be an effective assistant tool to cabin assembly works and provide an intuitive and real assembly experience for workers. This paper presents an assembly auxiliary system for spacecraft cabin products, which can provide technical support to the spacecraft cabin assembly industry.

  9. Probabilistic safety analysis of the Kozloduy NPP units 1-4 (WWER-440/230) using independent emergency feedwater system

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  11. Application of the RCM (Reliability Centered Maintenance) approach to the optimization of emergency feedwater pump system maintenance

    International Nuclear Information System (INIS)

    The major steps of the RCM analysis for the emergency feedwater pump system at the Dukovany NPP were as follows: Familiarization with the Dukovany maintenance process and strategy; Collection of information regarding data and information availability; system selection for the RCM study; implementation and use of suitable software; consulting the staff; FFA (Functional Failure Analysis) and FMEA (Failure Mode and Effect Analysis); Assessment of current periodical maintenance; and recommendations from the RCM analysis. The following Annexes are appended: A - Technological layouts; B - List of components; C - Information on functions; D - Critical components; E - FMEA; F - Consulting the staff (notes); G - Failure data (from RCM Workstation); H - Current periodical maintenance tasks; I - Selection of periodical maintenance tasks. (P.A.)

  12. Interaction region design and auxiliary detector systems for an EIC

    Directory of Open Access Journals (Sweden)

    Petti R.

    2016-01-01

    Full Text Available There are a number of exciting physics opportunities at a future electron-ion collider facility. One possible design for such a facility is eRHIC, where the current RHIC facility located at Brookhaven National Lab would be transformed into an electron-ion collider. It is imperative for a seamless integration of auxiliary detector systems into the interaction region design to have a machine that meets the needs for the planned physics analyses, as well as take into account the space constraints due to the tunnel geometry and the necessary beam line elements. In this talk, we describe the current ideas for integrating a luminosity detector, electron polarimeter, roman pots, and a low Q2-tagger into the interaction region for eRHIC.

  13. Auxiliary signal processing system for a multiparameter radar

    Science.gov (United States)

    Chandrasekar, V.; Gray, G. R.; Caylor, I. J.

    1993-01-01

    The design of an auxiliary signal processor for a multiparameter radar is described with emphasis on low cost, quick development, and minimum disruption of radar operations. The processor is based around a low-cost digital signal processor card and personal computer controller. With the use of such a concept, an auxiliary processor was implemented for the NCAR CP-2 radar during a 1991 summer field campaign and allowed measurement of additional polarimetric parameters, namely, the differential phase and the copolar cross correlation. Sample data are presented from both the auxiliary and existing radar signal processors.

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

    International Nuclear Information System (INIS)

    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

  15. Total loss of feedwater events analysis for Ulchin 1 and 2 emergency operating procedure strategy

    International Nuclear Information System (INIS)

    A series of quantitative evaluation was performed for Ulchin 1 and 2 total loss of feedwater events to support technically the validity of strategies presented in the Emergency Operating Procedure (EOP) and to estimate the margin for operator action time during the events. According to the analyses results, the Reactor Coolant Pump (RCP) and pressurizer heater trip strategy was proven to be adequate to get the possibility of the extension of operator action time as well as auxiliary feedwater system recovery time by delaying Steam Generator (SG) dryout time and limiting core exit temperature reaching time. In addition, the pressuirzer safety valves manual opening time could be delayed to almost 20 minutes after core exit temperature reached 330 .deg. C based on the core integrity evaluation results

  16. Study on potential threats against the leak tightness of the reactor containments due to pipe whips from hypothetical pipe ruptures in the steam- and feedwater systems

    International Nuclear Information System (INIS)

    Possible threats against the leak tightness of the reactor containments, due to pipe whips from hypothetical pipe ruptures in the steam- and feedwater systems, have been investigated for Forsmark 3/Oskarshamn 3, Ringhals 1, Oskarshamn 1 and Barsebaeck 2/Oskarshamn 2. Based on available drawings, such as installation drawings and isometric views of pipes, the pipe systems have been put together in new drawings with their bracing supports and containment walls. This inventory shows that pipe whips can occur on a number of places on the containments walls after hypothetical pipe ruptures in the steam- and main feedwater systems. In order to find out whether these pipe whips are real threats against the leak tightness, further analysis needs to be made but are out of the scope of this investigation

  17. Performance of a small solar-powered hybrid membrane system for remote communities under varying feedwater salinities

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, A.I.; Remy, C. [University of Wollongong, NSW (Australia). Environmental Engineering; Richards, B.S. [University of New South Wales, Sydney (Australia). Centre of Excellence for Advanced Silicon Photovoltaics and Photonics

    2005-07-01

    An estimated 1 billion people are living both without access to clean drinking water or electricity. The small photovoltaic (PV)-powered hybrid membrane system described here is designed to address the plight of some of these people. PV and membrane technologies are chosen due to suitability for operation in remote and often harsh conditions. An ultrafiltration (UF) pre-treatment is included to remove bacteria and most pathogens, while a reverse osmosis (RO) or nanofiltration (NF) membrane desalinates the brackish feedwater. Several parameters were examined in order to optimise the system performance, including (i) feed salt concentration, (ii) operating pressure, (iii) system recovery, (iv) specific energy consumption (SEC, kWh/m{sup 3}), and (v) salt retention. In addition, experiments were performed over a whole day to determine system performance under varying levels of solar radiation. The minimum SEC (relatively high due to the current single-pass mode of operation) varies from 5.5 kWh/m{sup 3} at a feed concentration of 1 g/L salt to 26 kWh/m{sup 3} at a feed concentration of 7.5 g/L salt, which is the upper limit of the system in terms of salt concentration. (author)

  18. Indigenous manufacturing realization of twin source and its auxiliary system

    International Nuclear Information System (INIS)

    operation) as well as Vacuum mode (DNB type vacuum immersed operation). The Twin Source shall be manufactured as per ASME guidelines for pressure vessel. Experiments on the Twin Source are foreseen in the near future, as all the auxiliary systems like 180 kW, RF generator system, vacuum vessel with Pumping station, Cooling water system, Data acquisition and control system (DACS) and other power supply systems are already installed in the lab premises. The paper discusses the FEA based engineering design, simplified manufacturing design, manufacturing experience with highlighting quality control and the system integration activities undertaken for the TWIN source test facility. (author)

  19. Low-cost auxiliary system for broadband NMR on strongly magnetic systems

    DEFF Research Database (Denmark)

    Nevald, Rolf; Hansen, Poul Erik

    1978-01-01

    A low cost auxiliary system consisting of He cryostat, superconducting magnet, and sample holder assembly with field probe has been constructed. The system meets the requirements of NMR on strongly paramagnetic or ordered magnetic materials, which are accurate temperature settings over a wide range...

  20. Conceptual design of the integral test loop (II) : Safety system and auxiliary system

    Energy Technology Data Exchange (ETDEWEB)

    Park, Choon Kyung; Song, Chul Hwa; Choi, Byeong Hae; Chung Moon Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    This report describes the results of the conceptual design work on the safety system and auxiliary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the safety and auxiliary systems is the same as that applied to the primary and secondary systems of the ITL as follows ; Reference plant : Korean Standard Nuclear Plant (KSNP), Height ratio :1/1, Volume ration: 1/200, Temperature, Pressure : Real plant conditions, The safety system contains a safety depressurization system (SDS) and a safety injection system (SIS). And the auxiliary system comprises a containment system, a shutdown cooling system (SCS), a volume control system (VCS), a makeup water system and a component cooling water system (CCWS). This conceptual design report describes the configurations and operation of the systems of the reference plant, and also describes the design philosophy of the corresponding components and systems of the ITL. In addition, this report specifies the design criteria and technical specifications of each component and system of the ITL. 6 refs., 11 figs., 21 tabs. (Author)

  1. Feedwater heater workshop proceedings

    International Nuclear Information System (INIS)

    The March 13-14, 1979 EPRI sponsored workshop on design, operation and maintenance problems with nuclear feedwater heaters identified a large number of significant technical areas of concern. The reported problems relate to nearly all facets of feedwater heater technology. In particular, hydrometallurgical problems such as stress corrosion of stainless tubing, erosion/corrosion wear of tube-tubesheet joints, and flashing related tube erosion at the entrance to the drain cooler region were reviewed in depth. Shortcomings in current design standards and design specifications were also discussed. Solutions to specific technical problems such as mechanics of flow induced vibration failures, improved synthesis of feedwater heater orientation with plant layout, and drain outlet piping design were presented. A list of important technical problems requiring additional research and development for improved design guidelines was developed. The proceedings contain all edited material of archival quality developed in this workshop

  2. The dynamic responses of the soil-auxiliary buildings structure interaction system

    International Nuclear Information System (INIS)

    The dynamic responses of the soil-auxiliary buildings structure interaction system in the nuclear power plant are concerned. The main distinguished feature of this study is that the extreme un-symmetry of the auxiliary buildings and reactor containment are considered. A Synthetical mechanical model for study is established. Finally, the analysis of the dynamic response of the Qinshan Nuclear Power Plant structure is taken as a simple example of applying this method and the numerical results are given

  3. Disturbance in the power system caused by auxiliary DC installation failure of switchyard

    Energy Technology Data Exchange (ETDEWEB)

    Mesic, M. [HEP Transmission System Operator, Zagreb (Croatia); Tesnjak, S.; Skok, S. [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2008-07-01

    Auxiliary direct current (DC) installation failures can lead to outages in power plants and compromise the security of power systems. In this study, a simplified stationary model was used to simulate an auxiliary DC installation in a switchyard. The aim of the study was to evaluate new International Electrotechnical Commission (IEC) standards for auxiliary DC installation dimensioning and analysis. Criteria included the dimensioning and selection of batteries; the calculation of conductor heating; voltage drop calculations; conductor squares in relation to permanent currents; and the evaluation of protection elements. The new standards were compared with the previous auxiliary system installation methodology. Results of the study suggested that the new standard has introduced significant improvements in short circuit current calculation. Laboratory tests for the measurement of short circuits showed that the active network has less of an impact on the auxiliary system than previous measuring methods. Alterations to the IEC standard will be required as a result of limitations to the short circuit current and new rectifier technology. Results of the study will be used to develop a new model and scheme for dimensioning and analyzing auxiliary DC installations. 9 refs., 4 tabs., 5 figs.

  4. Containment vessel, its auxiliary system and plant air conditioning system of advanced thermal reactor Fugen

    International Nuclear Information System (INIS)

    The functional requirement for, the design and the construction of, and the functional test on the containment vessel, its auxiliary system, the plant air conditioning and ventilation system of the advanced thermal reactor, Fugen, are described in detail. The main specifications of the containment vessel are as follows: The type enclosed cylinder, the maximum operating pressure 1.35 kg/cm2g, the maximum operating temperature 100 deg C, the leak rate 0.4%/day, the inner diameter 36 m. The height 64 m, the volume 40,900 m3, and the material JIS G3118, SGV-49. The containment vessel is provided with an hatch of 5 m diameter for carrying equipments in two air locks, many high and low voltage cable penetrations, pipe penetrations, a transfer shoot and isolation values. The functions and the specifications of the containment vessel and its auxiliary equipments are explained. The relating auxiliary systems are composed of the containment vessel spray system, the pool facility for steam blow-down, the recirculation system for the air in the vessel, the annulus evacuation system and its pressure control devices, the pressure measuring instruments and pressure relief valves and the temperature measuring devices for the containment vessel, and the object, function, layout and installation of these systems are explained. Concerning the air conditioning system, each main building has the special subsystem, and they are introduced. The progress stage of construction works and the procedure and results of the functional test at the site are described. (Nakai, Y.)

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

    International Nuclear Information System (INIS)

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

  6. Pb-17Li auxiliary and purification systems: design of the auxiliary Pb-Li loop for helium cooled lithium lead test blanket module

    International Nuclear Information System (INIS)

    This technical report describes the Pb-17Li auxiliary system proposed for Helium Cooled Lithium Lead (HCLL) Test Blanket Module (TBM) that will be installed and tested in ITER. The Pb-17Li auxiliary should ensure feeding and circulation of Pb-17Li liquid metal in this breeding blanket and removal of tritium produced by a nuclear reaction in TBM. The container with the Pb-17Li auxiliary system (dimensions HxLxW: 2.315 m x 2.19 m x 1.6 m) will be placed as close as possible to the TBM to prevent tritium permeation from the connection piping. The report describes developed design of the Pb-17Li auxiliary system that is from the functional point of view divided into the following parts: main circuit, detritization unit and cold trap, dosing and sampling systems, heating and cooling systems, and shielding and insulation. The Pb-17Li circuit is a closed loop with forced circulation of Pb-17Li. From the tank that, at the same time, is a Pb-17Li storage tank, liquid metal is pumped into the TBM where tritium is produced. The flow velocity in the Pb-17Li system will be controlled in the range of 0.1 to 1 kg/s. Pb-17Li outlet temperature from the TBM is 550 deg C. Tritium is removed from Pb-17Li in a detritiation unit. Corrosion products and impurities are removed in a cold trap. Design of the key system components as well as their structure material are described. The technical report determines and describes the Pb-17Li auxiliary system operating modes such as filling, start-up, operation at nominal parameters, shut-down, emergency operation and sampling. Also, the limits and terms of the Pb-17Li auxiliary system safe operation are defined. Requirements for the Pb-17Li auxiliary system installation, testing and maintenance are discussed. In conclusion, recommendations for further developments of the Pb-17Li auxiliary system are proposed. (author)

  7. A multivariate correlation method as a tool for fault detection and its application to feedwater system in a nuclear power plant

    International Nuclear Information System (INIS)

    A new method based on multivariate correlation has been developed for surveillance and anomaly detection in nuclear power plants. Natural signal noises observed under normal plant operations are applied in the method, thus the method is suitable for online surveillance. In the method, correlation of signals and noise properties in the process are estimated. Eigenvalues of normalized covariance matrix and correlation function among signals are used for surveillance of process dynamics and operational condition. The effectiveness of the method has been demonstrated with an application to feedwater system of a BWR plant. (author)

  8. Dynamic modelling and response characteristics of a magnetic bearing rotor system including auxiliary bearings

    Science.gov (United States)

    Free, April M.; Flowers, George T.; Trent, Victor S.

    1993-01-01

    Auxiliary bearings are a critical feature of any magnetic bearing system. They protect the soft iron core of the magnetic bearing during an overload or failure. An auxiliary bearing typically consists of a rolling element bearing or bushing with a clearance gap between the rotor and the inner race of the support. The dynamics of such systems can be quite complex. It is desired to develop a rotor-dynamic model and assess the dynamic behavior of a magnetic bearing rotor system which includes the effects of auxiliary bearings. Of particular interest is the effects of introducing sideloading into such a system during failure of the magnetic bearing. A model is developed from an experimental test facility and a number of simulation studies are performed. These results are presented and discussed.

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

    International Nuclear Information System (INIS)

    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)

  10. A study on the decay heat removal capability of a reactor vessel auxiliary cooling system

    International Nuclear Information System (INIS)

    A reactor vessel auxiliary cooling system (RVACS) is a potential candidate as a fully passive decay heat removal system for small FBRs. In this study the heat transfer performance of a collector with fins is discussed through experiment and the evaluation method is proposed for the heat removal capability of the system. (author)

  11. EASY 5 BWR simulation model for digital feedwater control design

    International Nuclear Information System (INIS)

    The development of a BWR simulation model in support of a program to design and evaluate the digital feedwater control system for the Monticello Boiling Water Reactor (BWR) is described. This model was developed in the EASY5 simulation language in conjunction with EPRI's Modular Modeling System (MMS) two-phase Library. The model consists of three main elements: the BWR reactor vessel module, the feedwater system model, and the steamline model. Transient results for the BWR vessel module and the feedwater system model are presented

  12. Process Modeling and Simulation of Feedwater Heaters Drains and Vents System of PFBR

    OpenAIRE

    T. Lakshmi Priyanka; K.R.S. Narayanan; T. Jayanthi; K.K.KuriaKose; S.A.V. Satya Murty

    2013-01-01

    Nuclear Power Plants are a complex system and need to be controlled very meticulously to avoid any catastrophe from occurring. The safety and availability of the power plant relies on the human operators both through their ability and reliability to ensure smooth and trouble-free plant operations. Training the operators on normal plant operation, maintenance, fault diagnosis and unforeseen emergencies in the plant helps reduce the latency period of the plant and thus increase the efficiency. ...

  13. Dynamic modelling and response characteristics of a magnetic bearing rotor system with auxiliary bearings

    Science.gov (United States)

    Free, April M.; Flowers, George T.; Trent, Victor S.

    1995-01-01

    Auxiliary bearings are a critical feature of any magnetic bearing system. They protect the soft iron core of the magnetic bearing during an overload or failure. An auxiliary bearing typically consists of a rolling element bearing or bushing with a clearance gap between the rotor and the inner race of the support. The dynamics of such systems can be quite complex. It is desired to develop a rotordynamic model which describes the dynamic behavior of a flexible rotor system with magnetic bearings including auxiliary bearings. The model is based upon an experimental test facility. Some simulation studies are presented to illustrate the behavior of the model. In particular, the effects of introducing sideloading from the magnetic bearing when one coil fails is studied.

  14. Process Modeling and Simulation of Feedwater Heaters Drains and Vents System of PFBR

    Directory of Open Access Journals (Sweden)

    T.Lakshmi Priyanka

    2013-10-01

    Full Text Available Nuclear Power Plants are a complex system and need to be controlled very meticulously to avoid any catastrophe from occurring. The safety and availability of the power plant relies on the human operators both through their ability and reliability to ensure smooth and trouble-free plant operations. Training the operators on normal plant operation, maintenance, fault diagnosis and unforeseen emergencies in the plant helps reduce the latency period of the plant and thus increase the efficiency. Operator Training Simulator has become an indispensable entity in imparting hands on training to these operators. Development of process simulators calls for the process to be designed, modeled and implemented to replicate the real plant in steady state and transient conditions.

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

    International Nuclear Information System (INIS)

    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)

  16. Performance of turbine auxiliaries and service systems at Rajasthan Atomic Power Station

    International Nuclear Information System (INIS)

    Performance of the turbine auxiliaries and service systems at the Rajasthan Atomic Power Station, India are described. Some of the specific problems encountered in connection with the feed water, turbine governing and common services like compressed air, chilled water, water treatment and chlorination systems are outlined. (K.B.)

  17. Evaluation of thermal striping for the plugging system in the secondary auxiliary cooling system in JOYO

    Energy Technology Data Exchange (ETDEWEB)

    Isozaki, Kazunori; Ogawa, Tooru; Kubo, Atsuhiko; Aoki, Hiroshi; Ozawa, Kenji [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Sugaya, Kazushi

    1998-05-01

    Scrutiny based on the convenient evaluation to verify whether we have the place where thermal striping in the pipe confluence part was thought to be a primary factor for the heavy accident or not has been done in JOYO. As the result, the big temperature difference ({Delta}Tin) existed at the inner pipe confluence part of the plugging system in the secondary main and auxiliary cooling system. Therefore, detailed evaluation of thermal striping was needed. With the thermocouples of high response installed, the temperature fluctuation in outer surface of the pipe was measured on the secondary auxiliary plugging system for the reason why the temperature difference ({Delta}Tin) was the biggest. And, the temperature fluctuation in inner surface of the pipe and stress occurring in the pipe plate thickness direction was evaluated by means of non-linear structure analysis system FINAS`. The above-mentioned evaluation results were as follows. (1) The maximum temperature fluctuation occurring in the pipe was always located from the center of inner pipe confluence to 10 mm position of the down-stream side. (2) The maximum temperature fluctuation range was about 33degC in outer surface of the pipe. And, controlling frequency of the temperature fluctuation was 0.04 Hz and 0.09 Hz. (3) Time delay was almost never contained in the temperature fluctuation elements between inner and outer surface of the pipe. And, the big temperature distribution did not occur in the pipe plate thickness direction was confirmed that the big temperature distribution did not occur in the pipe plate thickness direction. The temperature fluctuation range in the pipe inner surface was almost the same as that of the pipe outer surface. It was confirmed that the stress occurring there was enough lowered than the design fatigue limit of SUS304 which was the materials in the confluence part of the plugging system inner part in the secondary main and auxiliary cooling system. (J.P.N.)

  18. 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.

  19. Auxiliary control system for irradiation specimen automatic transmission based on configuration software

    International Nuclear Information System (INIS)

    Auxiliary control system realizes sequential control and trace display and automatic transmission for irradiated specimen, which bases on configuration software (MCGS) and industrial control computer as the control platform. The system uses digital I/O cards to establish system state detection and output control arrays. It is showed that the structure posses stable, reliable and security characteristics and well meets the needs of specimen transmission and controlling in industrial automation. (authors)

  20. A Model of Ship Auxiliary System for Reliable Ship Propulsion

    OpenAIRE

    Martinović, Dragan; Tudor, Mato; Bernečić, Dean

    2012-01-01

    The main purpose of a vessel is to transport goods and passengers at minimum cost. Out of the analysis of relevant global databases on ship machinery failures, it is obvious that the most frequent failures occur precisely on the generator-running diesel engines. Any failure in the electrical system can leave the ship without propulsion, even if the main engine is working properly. In that case, the consequences could be devastating: higher running expenses, damage to the ship, oil spill or su...

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

    International Nuclear Information System (INIS)

    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

  2. Feedwater control device for a reactor

    International Nuclear Information System (INIS)

    Purpose: To eliminate the water level deviation due to the recycling flowrate, as well as enable a stable control to a reference value even upon changes in the recycling flowrate caused by the variation in the opening degree of a minimum flow valve. Constitution: Reactor recycling system comprises a feedwater pump, a flowrate control valve, a reactor water level detector, and a minimum flow line and a minimum flow valve for preventing the overheating of the feedwater pump at a low flowrate. A flowrate compensator is further disposed, in which a recycling flowrate signal is subtracted from a pump flow rate signal and the result is fedback as a compensated pump flowrate signal. This enables the control system to operate at a rapid response for suppressing the effect of the recycling flowrate as external disturbance, whereby the water level in the reactor can be controlled stably to the reference level and the possibility deviating from an allowable range is decreased significantly. (Kamimura, M.)

  3. 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)

  4. 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)

  5. Lead corrosion and transport in simulated secondary feedwater

    International Nuclear Information System (INIS)

    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)

  6. Lead corrosion and transport in simulated secondary feedwater

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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.

  8. Energy simulation of solar assisted absorption system and examination of clearness index effects on auxiliary heating

    International Nuclear Information System (INIS)

    The smog and pollutants in the atmospheric air of heavily populated urban areas are anticipated to have substantial adverse effects on the collection of solar energy and the performance of solar energy systems. The objectives of this study are (a) to develop a simulation model for analyzing the performance of a water-LiBr solar assisted absorption system with an auxiliary heating source and (b) to examine the effects of clearness index on the auxiliary heating requirements. To achieve the objectives, a numerical model for a water-LiBr solar assisted absorption system is developed, and the influence of a reduction in the clearness index, based on actual recorded data, is investigated for constant and time varying cooling loads. Under the condition of peak solar gain on July 21, when a 1000 m2 solar collector is designed to provide 70% of the heating energy required for a constant cooling load of 1265 MJ/h (=100 refrigeration tons), as the system coefficient of performance decreases due to higher ambient temperatures, it is found that a reduction in the clearness index from 0.63 to 0.52 results in a 67% increase in auxiliary heating required of the boiler. It is concluded that accounting for clearness index data is necessary for accurate prediction of solar energy collection

  9. Creating and manipulating non-Abelian anyons in cold atom systems using auxiliary bosons

    Science.gov (United States)

    Zhang, Yuhe; Sreejith, G. J.; Jain, J. K.

    2015-08-01

    The possibility of realizing bosonic fractional quantum Hall effect in ultracold atomic systems suggests a new route to producing and manipulating anyons, by introducing auxiliary bosons of a different species that capture quasiholes and thus inherit their nontrivial braiding properties. States with localized quasiholes at any desired locations can be obtained by annihilating the auxiliary bosons at those locations. We explore how this method can be used to generate non-Abelian quasiholes of the Moore-Read Pfaffian state for bosons at filling factor ν =1 . We show that a Hamiltonian with an appropriate three-body interaction can produce two-quasihole states in two distinct fusion channels of the topological "qubit." Characteristics of these states that are related to the non-Abelian nature can be probed and verified by a measurement of the effective relative angular momentum of the auxiliary bosons, which is directly related to their pair distribution function. Moore-Read states of more than two quasiholes can also be produced in a similar fashion. We investigate some issues related to the experimental feasibility of this approach, in particular, how large the systems should be for a realization of this physics and to what extent this physics carries over to systems with the more standard two-body contact interaction.

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

    International Nuclear Information System (INIS)

    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. Feasibility of AN Ecrh System for Jet:. Plant Layout, Auxiliaries and Services

    Science.gov (United States)

    Lennholm, M.; Bouquey, F.; Braune, H.; Farthing, J.; Garavaglia, S.; Giruzzi, G.; Granucci, G.; Jennison, M.; Parkin, A.

    2011-02-01

    A study conducted over the last year to asses the desirability and feasibility of installing an ECRH system on the JET tokamak has concluded that such a system is indeed both desirable and feasible. Details of physics studies, launcher and transmission line design, and power supplies are presented elsewhere in these proceedings. This paper concentrates on the logistical implications of installing this system at JET. The paper addresses issues such as port allocation and plant location. The study has concluded that a new building will be needed to house the ECRH plant. Building layout proposals are presented together with considerations regarding the required auxiliary equipment.

  12. Fort St. Vrain helium circulator auxiliary systems: dynamic performance evaluation and acceptance tests

    International Nuclear Information System (INIS)

    The purpose of the tests described is to show that the dynamic performance of the Fort St. Vrain helium circulator auxiliary systems satisfies all the guidelines and criteria established and agreed to by Public Service Company of Colorado (PSC), Proto-Power, and General Atomic Company (GA). Specifically, it is shown that transfers to and from backup bearing water and helium purification system transients do not cause any circulator trips. Furthermore, at PSC's request, in an effort to resolve any NFSC questions concerning these systems, the satisfactory repeatability of their dynamic performance is shown beyond any doubt.

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

    International Nuclear Information System (INIS)

    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

  14. Design, Fabrication, and Testing of an Auxiliary Cooling System for Jet Engines

    Science.gov (United States)

    Leamy, Kevin; Griffiths, Jim; Andersen, Paul; Joco, Fidel; Laski, Mark; Balser, Jeffrey (Technical Monitor)

    2001-01-01

    This report summarizes the technical effort of the Active Cooling for Enhanced Performance (ACEP) program sponsored by NASA. It covers the design, fabrication, and integrated systems testing of a jet engine auxiliary cooling system, or turbocooler, that significantly extends the use of conventional jet fuel as a heat sink. The turbocooler is designed to provide subcooled cooling air to the engine exhaust nozzle system or engine hot section. The turbocooler consists of three primary components: (1) a high-temperature air cycle machine driven by engine compressor discharge air, (2) a fuel/ air heat exchanger that transfers energy from the hot air to the fuel and uses a coating to mitigate fuel deposits, and (3) a high-temperature fuel injection system. The details of the turbocooler component designs and results of the integrated systems testing are documented. Industry Version-Data and information deemed subject to Limited Rights restrictions are omitted from this document.

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

    International Nuclear Information System (INIS)

    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 H2/O2 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-EPRTM 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

  16. 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)

    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)

  17. Steady-state dynamic behavior of an auxiliary bearing supported rotor system

    Science.gov (United States)

    Xie, Huajun; Flowers, George T.; Lawrence, Charles

    1995-01-01

    This paper investigates the steady-state responses of a rotor system supported by auxiliary bearings in which there is a clearance between the rotor and the inner race of the bearing. A simulation model based upon the rotor of a production jet engine is developed and its steady-state behavior is explored over a wide range of operating conditions for various parametric configurations. Specifically, the influence of rotor imbalance, support stiffness, and damping is studied. It is found that imbalance may change the rotor responses dramatically in terms of frequency contents at certain operating speeds. Subharmonic responses of 2nd order through 10th order are all observed except the 9th order. Chaotic phenomenon is also observed. Jump phenomena (or double-valued responses) of both hard-spring type and soft-spring type are shown to occur at low operating speeds for systems with low auxiliary bearing damping or large clearance even with relatively small imbalance. The effect of friction between the shaft and the inner race of the bearing is also discussed.

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

    International Nuclear Information System (INIS)

    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

  19. Simulation Research on Passive Auxiliary Feedwater System for NPP%核电站非能动辅助给水系统仿真研究

    Institute of Scientific and Technical Information of China (English)

    师二兵; 方成跃; 王畅; 赵观辉

    2016-01-01

    以先进核电站AP1000为研究对象,在其蒸汽发生器二次侧设计了1套耗汽驱动汽动辅助给水泵的非能动辅助给水系统.使用RELAP5程序计算分析全厂断电事故下设计系统的运行特性,研究其应对事故工况的能力.计算结果表明:全厂断电事故下,设计的非能动辅助给水系统可有效地排出堆芯余热,保证反应堆的安全;由于冷却剂体积收缩,170 min时稳压器排空;该系统可连续运行200 min,排出事故后的大部分堆芯余热.非能动辅助给水系统可作为全厂断电事故后的应急缓解方案.

  20. Development of a New Condensation Heat Transfer Model for the Nearly Horizontal Tube of the APR+ PAFS (Passive Auxiliary Feed-water System)

    International Nuclear Information System (INIS)

    Korea Atomic Energy Research Institute (KAERI) has recently conducted PASCAL experiment to confirm the performance of PAFS. From the work, it is founded that the condensation heat transfer coefficient predicted by best estimated safety analysis code MARS (Multi-dimensional Analysis of Reactor Safety) was underestimated compared to the experimental data. Shah correlation embedded in MARS code is the typical model using the empirical two-phase multiplier to determine condensation heat transfer coefficient for annular flow in the condensing tube. On the other hand, the PASCAL experiment indicated that a stratified-wavy flow generally tends to occur in the downward inclined tube of the condensation heat exchanger. Therefore, in order to improve the prediction capability of safety analysis codes for PAFS, the present study has proposed a new condensation heat transfer model package for the nearly horizontal tube, which determines mechanistically the local heat transfer coefficient based on the flow regimes. To enhance prediction capability of one-dimensional best estimated code MARS for the PAFS of the Korean 3.5 generation nuclear power plant APR+, a new condensation heat transfer model package has been developed. The model package consists of the one-dimensional separated model for the void fraction, a flow regime model based on wetted angle, and the condensation heat transfer correlations. The model package considers the inclination angle of the condensing tube and various flow regimes that are expected in the tube during condensation process

  1. Development of a New Condensation Heat Transfer Model for the Nearly Horizontal Tube of the APR+ PAFS (Passive Auxiliary Feed-water System)

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Taehwan; Jeong, Jaejun; Yun, Byongjo [Pusan National Univ., Busan (Korea, Republic of); Kang, Kyongho; Park, Yusun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cheon, Jong [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Korea Atomic Energy Research Institute (KAERI) has recently conducted PASCAL experiment to confirm the performance of PAFS. From the work, it is founded that the condensation heat transfer coefficient predicted by best estimated safety analysis code MARS (Multi-dimensional Analysis of Reactor Safety) was underestimated compared to the experimental data. Shah correlation embedded in MARS code is the typical model using the empirical two-phase multiplier to determine condensation heat transfer coefficient for annular flow in the condensing tube. On the other hand, the PASCAL experiment indicated that a stratified-wavy flow generally tends to occur in the downward inclined tube of the condensation heat exchanger. Therefore, in order to improve the prediction capability of safety analysis codes for PAFS, the present study has proposed a new condensation heat transfer model package for the nearly horizontal tube, which determines mechanistically the local heat transfer coefficient based on the flow regimes. To enhance prediction capability of one-dimensional best estimated code MARS for the PAFS of the Korean 3.5 generation nuclear power plant APR+, a new condensation heat transfer model package has been developed. The model package consists of the one-dimensional separated model for the void fraction, a flow regime model based on wetted angle, and the condensation heat transfer correlations. The model package considers the inclination angle of the condensing tube and various flow regimes that are expected in the tube during condensation process.

  2. Linear CCD attitude measurement system based on the identification of the auxiliary array CCD

    Science.gov (United States)

    Hu, Yinghui; Yuan, Feng; Li, Kai; Wang, Yan

    2015-10-01

    Object to the high precision flying target attitude measurement issues of a large space and large field of view, comparing existing measurement methods, the idea is proposed of using two array CCD to assist in identifying the three linear CCD with multi-cooperative target attitude measurement system, and to address the existing nonlinear system errors and calibration parameters and more problems with nine linear CCD spectroscopic test system of too complicated constraints among camera position caused by excessive. The mathematical model of binocular vision and three linear CCD test system are established, co-spot composition triangle utilize three red LED position light, three points' coordinates are given in advance by Cooperate Measuring Machine, the red LED in the composition of the three sides of a triangle adds three blue LED light points as an auxiliary, so that array CCD is easier to identify three red LED light points, and linear CCD camera is installed of a red filter to filter out the blue LED light points while reducing stray light. Using array CCD to measure the spot, identifying and calculating the spatial coordinates solutions of red LED light points, while utilizing linear CCD to measure three red LED spot for solving linear CCD test system, which can be drawn from 27 solution. Measured with array CCD coordinates auxiliary linear CCD has achieved spot identification, and has solved the difficult problems of multi-objective linear CCD identification. Unique combination of linear CCD imaging features, linear CCD special cylindrical lens system is developed using telecentric optical design, the energy center of the spot position in the depth range of convergence in the direction is perpendicular to the optical axis of the small changes ensuring highprecision image quality, and the entire test system improves spatial object attitude measurement speed and precision.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sweeney, F.J. (Oak Ridge National Lab., TN (United States)); Carroll, D.G. (General Electric Co., San Jose, CA (United States)); Chen, C. (Tennessee Univ., Knoxville, TN (United States)); Crane, C.; Dalton, R. (Florida Univ., Gainesville, FL (United States)); Taylor, J.R. (Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States)); Tosunoglu, S. (Texas Univ., Austin, TX (United States))

    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.

  5. Cutting BWR feedwater crud levels further

    International Nuclear Information System (INIS)

    Reducing iron input to the BWR primary system is an important first step in decreasing radiation fields and occupational exposure. For feedwater iron, the specified optimum concentration is 0.1-0.5 ppb, as demonstrated by the newest ''low crud'' plants operating in Japan. Recent advances in condensate filtration will achieve the levels needed for optimised water chemistry and promise great benefits. Of particular interest is a newly developed filterdemineraliser septum that separates the filtration and ion exchange functions to allow each to be specifically optimised for BWR conditions. (author)

  6. Salient aspects on the choice of classs 3 system - the diesel generating sets - engine, auxiliaries (Paper No. 2.5)

    International Nuclear Information System (INIS)

    The class 3 system is a basic requirement in all nuclear installations. Selection of this major equipment and associated auxiliaries plays an important role in the overall performance and reliability of the whole system. This paper deals with various sub-systems of class 3 power like engine, alternator, auxiliaries. It deals with method of arriving at capacity of sets based on connected loads, starting requirement and choice of engine based on site conditions, fuel used, duty, speed, mean effective pressures, typical layout of set and auxiliaries with its various sub-systems. Various engine starting methods, advantages and disadvantages, area requirements, choice of foundation, general guidlines for installation, testing and commissioning of medium size plant will be discussed in reference to applicable codes and practices. (author). 3 refs., 1 fig

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

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

    International Nuclear Information System (INIS)

    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

  10. 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.

  11. 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.

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

    International Nuclear Information System (INIS)

    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.

  13. Modelling a feed-water control system of a steam generator in the framework of the dynamic reliability

    OpenAIRE

    Babykina, Génia; Brinzei, Nicolae; Aubry, Jean-François; Deleuze, Gilles

    2013-01-01

    The paper deals with the exploration of an industrial complex system behaviour and its probabilistic safety assessment (PSA). The main purposes are to build a model which realistically represents the system structure, to carry out the Monte Carlo study of the system behaviour and to perform the analysis of system reliability. The complexity of the system consists in its structure, size, dynamic operational behaviour, complex interactions between the system and its environment, etc. The theore...

  14. 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.

  15. 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)

  16. Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system

    International Nuclear Information System (INIS)

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The RVACS can be characterized as a surface-based decay heat removal system. It is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to the core volume) and decay heat removal capability (proportional to the vessel surface area). Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environmental side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps or annular regions formed between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions among the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very

  17. Scaling analysis for the direct reactor auxiliary cooling system for FHRs

    International Nuclear Information System (INIS)

    Highlights: • A scaling analysis for the direct reactor auxiliary cooling system is performed. • Key dimensionless numbers are developed and similarity laws are proposed. • A scaling methodology that consists of core scaling and loop scaling is developed. • Scientific design of a scaled-down high-temperature DRACS facility is obtained. - Abstract: The Direct Reactor Auxiliary Cooling System (DRACS) is a passive residual heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three natural circulation/convection loops that rely on buoyancy as the driving force and are coupled via two heat exchangers, namely, the DRACS heat exchanger and the natural draft heat exchanger. A fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during reactor normal operation, and to activate the DRACS in accidents when the reactor is shut down. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional flow formulation, the governing equations are non-dimensionalized by introducing appropriate dimensionless parameters. The key dimensionless numbers that characterize the DRACS system are obtained from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has been developed, which consists of a core scaling and a loop scaling. The consistency between the core and loop scaling is examined via the reference volume ratio, which can be obtained from

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

    International Nuclear Information System (INIS)

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

  19. Secondary coolant purification system with demineralizer bypass

    International Nuclear Information System (INIS)

    Apparatus and method are provided for a nuclear stream supply system for adequately controlling the chemistry of the secondary coolant. The invention includes means for the addition of volatile chemicals, a full flow condensate demineralizer, continuous blowdown capability, radiation detection means, a condensate demineralizer bypass line, and an auxiliary demineralizer bypass line, and an auxiliary demineralizer sized to handle full blowdown flow. The auxiliary demineralizer is cut into the system and the steam generator feedwater flow is bypassed around the full flow condensate demineralizer whenever radioactivity is detected in the secondary coolant

  20. Use of Reliability-Centered Maintenance for the McGuire nuclear station feedwater system. Final report

    International Nuclear Information System (INIS)

    An earlier prototype application of Reliability-Centered Maintenance (RCM) to the Component Cooling Water System at Florida Power and Light's Turkey Point Plant demonstrated that the RCM methodology is applicable for developing or revising Preventive Maintenance (PM) programs in operating nuclear power plants. This report discusses an additional RCM application made to the Main Feed Water System at Duke Power Company's McGuire Nuclear Station. The study was conducted in order to demonstrate more fully all aspects of the RCM approach. The report reviews the RCM methodology and details its application to the Main Feed Water System. The resulting RCM-based PM program is briefly compared to the current PM program for the system. The RCM process suggested four current PM tasks as candidates for Condition Directed (CD) consideration or revision and two current CD tasks as candidates for expanded scope. The RCM-based program also added nine new CD tasks as candidates to improve the effectiveness of the PM program and two new Failure Finding tasks as candidates to avoid startup delays. The report also discusses the lessons learned through the two studies. The study shows the RCM methodology to be an effective tool both for defining a PM program and for providing traceable documentation to enhance the credibility of the PM program

  1. Rotor elements of combined turbo-pump units for automated emergency feedwater system of NPP heat and mass exchange equipment

    International Nuclear Information System (INIS)

    The paper offers promising alternative advanced options for drives of feeding devices based on combined turbine units, considering the use of wet steam as a working medium. It explains the dependence of the efficient start of turbo-pump units on the degree of information support of NPP process control system

  2. Performance of a small solar-powered hybrid membrane system for remote communities under varying feedwater salinites

    OpenAIRE

    Schaefer, Andrea; Remy, C.; Richards, B.S.

    2004-01-01

    An estimated 1 billion people are living both without access to clean drinking water or electricity. The small photovoltaic (PV)-powered hybrid membrane system described here is designed to address the plight of some of these people. PV and membrane technologies are chosen due to suitability for operation in remote and often harsh conditions. An ultrafiltration (UF) pre-treatment is included to remove bacteria and most pathogens, while a reverse osmosis (RO) or nanofiltration (NF)...

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Shoda, Y. [Mitsubishi Heavy Industries, Ltd., Kobe City, Hyogo (Japan); Ishihara, N.; Miyata, H. [Mitsubishi Heavy Industries, Ltd., Takasago City, Hyogo (Japan); Ohira, T. [The Japan Atomic Power Company, Tokyo (Japan); Watanabe, Y. [Hokkaido Electric Power Co., Inc., Sapporo City, Hokkaido (Japan); Nonaka, Y. [Kyushu Electric Power Co., Inc., Fukuoka City, Fukuoka (Japan)

    2010-07-01

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

  5. Mining a coal seam below a heating goaf with a force auxiliary ventilation system at Longhua underground coal mine, China

    Institute of Scientific and Technical Information of China (English)

    Wang Gang; Xie Jun; Xue Sheng; Wang Haiyang

    2015-01-01

    Extraction of a coal seam which lies not far below a heating goaf can be a major safety challenge. A force auxiliary ventilation system was adopted as a control method in successful extraction and recovery of the panel 30110 of the #3?1 coal seam, which is about 30–40 m below the heating goaf of the #2?2 seam at Longhua underground coal mine, Shanxi Province, China. Booster fans and ventilation control devices such as doors and regulators were used in the system. The results show that, provided that a force auxiliary ventilation system is properly designed to achieve a pressure balance between a panel and its overlying goaf, the system can be used to extract a coal seam overlain by a heating goaf. This paper describes the design, installation and performance of the ventilation system during the extraction and recovery phases of the panel 30110.

  6. Active mass damper system for high-rise buildings using neural oscillator and position controller considering stroke limitation of the auxiliary mass

    Science.gov (United States)

    Hongu, J.; Iba, D.; Nakamura, M.; Moriwaki, I.

    2016-04-01

    This paper proposes a problem-solving method for the stroke limitation problem, which is related to auxiliary masses of active mass damper systems for high-rise buildings. The proposed method is used in a new simple control system for the active mass dampers mimicking the motion of bipedal mammals, which has a neural oscillator synchronizing with the acceleration response of structures and a position controller. In the system, the travel distance and direction of the auxiliary mass of the active mass damper is determined by reference to the output of the neural oscillator, and then, the auxiliary mass is transferred to the decided location by using a PID controller. The one of the purpose of the previouslyproposed system is stroke restriction problem avoidance of the auxiliary mass during large earthquakes by the determination of the desired value within the stroke limitation of the auxiliary mass. However, only applying the limited desired value could not rigorously restrict the auxiliary mass within the limitation, because the excessive inertia force except for the control force produced by the position controller affected on the motion of the auxiliary mass. In order to eliminate the effect on the auxiliary mass by the structural absolute acceleration, a cancellation method is introduced by adding a term to the control force of the position controller. We first develop the previously-proposed system for the active mass damper and the additional term for cancellation, and verity through numerical experiments that the new system is able to operate the auxiliary mass within the restriction during large earthquakes. Based on the comparison of the proposed system with the LQ system, a conclusion was drawn regarding which the proposed neuronal system with the additional term appears to be able to limit the stroke of the auxiliary mass of the AMD.

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

    Science.gov (United States)

    Xu, Weidong; Xuan, Weimin; Yao, Lieying; Wang, Yingqiao

    2012-03-01

    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.

  8. Analysis of problems in application and start up of control system of steam turbine for feedwater pump%给水泵汽轮机控制系统应用及启动中的问题分析

    Institute of Scientific and Technical Information of China (English)

    毕华南; 李红艳

    2009-01-01

    论述了西门子锅炉给水泵汽轮机控制系统在发电厂的使用情况,介绍了该系统的控制原理,集中描述了转速脉冲信号(转速测量值)、转速设定值的形成,最终实现锅炉给水泵汽轮机转速控制指令的全过程.结合该系统在发电厂给水泵汽轮机中的具体应用情况,对运行过程中存在的问题及处理方法进行了简要说明.%The service condition of control system of the steam turbine for boiler feedwater pump produced by Siemens was discoursed. The control principle of the system was introduced. The all process, in which includes the measurement of speed impulse signal, the formation of speed settings, and the creation of control instruction for the steam turbine speed, were described in detail. Combining with the concrete service condition of the system in steam turbine for feedwater pump, some problems occurred in op-eration and handling methods were explained tersely.

  9. Scaling analysis for the direct reactor auxiliary cooling system for FHRs

    International Nuclear Information System (INIS)

    The Direct Reactor Auxiliary Cooling System (DRACS) is a passive residual heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines the coated particle fuel and graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three natural circulation/convection loops that rely on buoyancy as the driving force and are coupled via two heat exchangers, namely, the DRACS heat exchanger and the natural draft heat exchanger. A fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during reactor normal operation, and to activate the DRACS in accidents when the reactor is shut down. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. In this paper, a detailed scaling analysis for the DRACS is performed, which will provide guidance for the design of scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional flow formulation, the governing equations are non-dimensionalized by introducing appropriate dimensionless parameters. The key dimensionless numbers that characterize the DRACS system are obtained from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has been developed, which consists of a core scaling and a loop scaling. The consistency between the core and loop scaling is examined via the reference volume ratio, which can be obtained from both the core and loop scaling processes. The scaling methodology and similarity laws have been applied to obtain a scientific design of a scaled-down high-temperature DRACS test facility

  10. Holistic Modeling, Design & Analysis of Integrated Stirling and Auxiliary Clean Energy Systems for Combined Heat and Power Applications

    Science.gov (United States)

    Nayak, Amrit Om

    The research revolves around the development of a model to design and analyze Stirling systems. Lack of a standard approach to study Stirling systems and difficulty in generalizing existing approaches pose stiff challenges. A stable mathematical model (integrated second order adiabatic and dynamic model) is devised and validated for general use. The research attempts to design compact combined heat and power (CHP) system to run on multiple biomass fuels and solar energy. Analysis is also carried out regarding the design of suitable auxiliary systems like thermal energy storage system, biomass moisture removal system and Fresnel solar collector for the CHP Stirling system.

  11. IE Information Notice No. 86-04: Transient due to loss of power to integrated control system at a pressurized water reactor designed by Babcock and Wilcox

    International Nuclear Information System (INIS)

    On December 26, 1985, Rancho Seco was operating on automatic control at a constant power level of 710 MWe (76% of licensed power). At 4:14 a.m., power to the integrated control system (ICS) was lost. The annunciator alarm for ''Loss of ICS or Fan Power'' sounded. As designed, ICS demand signals went to midscale. The main feedwater valves closed to 50%, and the atmospheric dump valves, turbine bypass valves, and one set of auxiliary feedwater valves opened to 50%. The main feedwater pump speed was reduced to minimum. Low discharge pressure at the main feedwater pump caused the motor-driven auxiliary feedwater pump to start automatically. The net decrease in feedwater flow caused the reactor to trip on high reactor coolant system (RCS) pressure. After the reactor trip, the above ICS valves remained at 50% (i.e., could not be operated from the control room) causing excessive cooling of the RCS which was exacerbated by autostarting of the dual-drive auxiliary feedwater pump. During the 26 minutes required to restore ICS power, operators acted to minimize the resulting transient. However, difficulties were experienced with manipulation of valves, operation of pumps, and control of various liquid levels, pressures, and temperatures. RCS pressure decreased to a minimum of 1,064 psig at 4:21 a.m. At 4:40 a.m., the lowest RCS temperature (386 F) during the cooling transient was reached. RCS pressure at that time was 1,413 psig. Eventually, a senior reactor operator discovered that switches which supplied power to the ICS dc power supplies were in the off position and set them to the on position

  12. PS auxiliary magnet

    CERN Multimedia

    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. Synchronous dynamics of a coupled shaft/bearing/housing system with auxiliary support from a clearance bearing: Analysis and experiment

    Science.gov (United States)

    Lawen, James L., Jr.; Flowers, George T.

    1995-01-01

    This study examines the response of a flexible rotor supported by load sharing between linear bearings and an auxiliary clearance bearing. The objective is to develop a better understanding of the dynamical behavior of a magnetic bearing supported rotor system interacting with auxiliary bearings during a critical operating condition. Of particular interest is the effect of coupling between the bearing/housing and shaft vibration on the rotordynamical responses. A simulation model is developed and a number of studies are performed for various parametric configurations. An experimental investigation is also conducted to compare and verify the rotordynamic behavior predicted by the simulation studies. A strategy for reducing synchronous shaft vibration through appropriate design of coupled shaft/bearing/housing vibration modes is identified.

  14. Electrochemical potential monitoring in the feedwater at the St. Lucie 2 PWR

    International Nuclear Information System (INIS)

    Reducing conditions are necessary in nuclear steam generators to minimize local corrosion phenomena, e.g., stress corrosion cracking. To achieve this goal, the electrochemical potential (ECP) should be maintained below the threshold at which these corrosion mechanisms occur. Ideally, in-situ ECP measurements in the steam generator would allow for monitoring and controlling oxidant ingress to maintain the desired reducing conditions. Unfortunately such measurements are difficult to implement. However, oxidant ingress can be monitored and controlled via ECP measurements in the high pressure feedwater. A feedwater ECP monitoring system was installed at the St. Lucie 2 PWR of Florida Power and Light Company to assess this approach. Variations in feedwater ECP and corrosion product transport were monitored. As chemistry was varied, Alloy 600 ECP was shown to depend on the feedwater hydrazine/condensate dissolved oxygen ratio. Operation at hydrazine/oxygen ratio above 6 was necessary to maintain a low ECP

  15. Header feedwater supply and power distribution stability in channel boiling water cooled reactors

    International Nuclear Information System (INIS)

    Boundaries of radial-azimuthal instability of the reactor neutron field during the supply of all feedwater and a part of it (25%) to downtake pipes of the separating drum (75% of feedwater come to distributive group headers) are found out for NPP with a RBMK type reactor. Results of computer calculation of the transient process at NPP caused by 2% step increase of nominal pressure in a head collector of a feedwater electric pump are also presented for comparison of the above methods of feed-water supply. Calculation is carried out according to the OKA program with provision for the control system of the reactor total power. It is shown that the boundary of ''mean period'' instability does not change but the reserve in respect to the ''fast'' space instability slightly increases when header feedwater supply at NPP from RBMK is used. It is noted that requirements to the pressure regulator system quick action in a separating drum are increased when the header feedwater supply is used. This fact is explained by the fact that considerable pressure drop in a separating drum occurs during some accidents (for example, at false operation of the emergensy protective system)

  16. Boiler feedwater treatment using reverse osmosis at Suncor OSG

    International Nuclear Information System (INIS)

    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

  17. Feedwater heaters functional analysis at Embalse NGS

    International Nuclear Information System (INIS)

    This study is concerned with the analysis or feedwater heaters, to detect actual failure or a bad trend beyond acceptable operating limits. When these situations are identified, preventive or corrective maintenance must be done. 2 tabs., 14 figs

  18. 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

  19. Reconstructive inverse dynamics in feedwater control

    International Nuclear Information System (INIS)

    The history of nuclear reactor operations during the last two decades has indicated the need to use advanced control, monitoring, and diagnostics techniques to achieve major improvements in power plant performance. In this paper, the authors present an application of reconstructive inverse dynamics (RID) to the feedwater-train control system for a typical pressurized water reactor (PWR). Existing controllers used in power plants, such as proportional-integral-derivative (PID) controllers, are designed using linear techniques. Although PIDs have proven simple and reliable in reactor operations, their limitation is related to the degree of linearity of the system dynamics. When the system undergoes a state-transition caused by a significant nonlinearity, the linear controller gains may not provide adequate compensation on the system dynamics, regardless of how sophisticated the control designs are. This fact is the main motivation behind searching for an appropriate nonlinear control law. The RID is a version of the inverse-dynamics method currently used in robotics. The power of RID lies in its unique control law for any system with plant nonlinearities. The control law includes inverse dynamics and state reconstruction to follow a demanded trajectory. The RID structure is best understood with a coupled system dynamics

  20. Decontamination of the reactor pressure vessel and further internals and auxiliary systems in the German boiling water reactor Isar-1

    International Nuclear Information System (INIS)

    The German nuclear power plant ISAR 1 (KKI 1), a 878 MWe boiling water reactor of KWU design, was shut down on March 17th, 2011. With the objective to minimize the plants activity inventory accompanied by the reduction of contact dose rates of systems and components the project 'decontamination of the RPV incl. steam dryer and water separator and the connected auxiliary systems' was implemented in the first quarter of 2015. One major focus within the project was the specific in-situ decontamination of the steam dryer.

  1. RELAP5 Analysis of OECD/NEA ROSA Project Experiment Simulating a PWR Loss-of-Feedwater Transient with High-Power Natural Circulation

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2012-01-01

    Full Text Available A ROSA/LSTF experiment was conducted for OECD/NEA ROSA Project simulating a PWR loss-of-feedwater (LOFW transient with specific assumptions of failure of scram that may cause natural circulation with high core power and total failure of high pressure injection system. Auxiliary feedwater (AFW was provided to well observe the long-term high-power natural circulation. The core power curve was obtained from a RELAP5 code analysis of PWR LOFW transient without scram. The primary and steam generator (SG secondary-side pressures were maintained, respectively, at around 16 and 8 MPa by cycle opening of pressurizer (PZR power-operated relief valve and SG relief valves for a long time. Large-amplitude level oscillation occurred in SG U-tubes for a long time in a form of slow fill and dump while the two-phase natural circulation flow rate gradually decreased with some oscillation. RELAP5 post-test analyses were performed to well understand the observed phenomena by employing a fine-mesh multiple parallel flow channel representation of SG U-tubes with a Wallis counter-current flow limiting correlation at the inlet of U-tubes. The code, however, has remaining problems in proper predictions of the oscillative primary loop flow rate and SG U-tube liquid level as well as PZR liquid level.

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

    International Nuclear Information System (INIS)

    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)

  3. Determination of the thermal loadings affecting the auxiliary lines of the reactor coolant system in French PWR plants

    International Nuclear Information System (INIS)

    The various incidents, imputed to thermal fatigue, which occurred throughout the world on the auxiliary lines of Reactor Coolant System (SIS, RHR, CVC), led EDF to urge a research program in order to determine the origins and the consequences of these problems for the French nuclear power plants. In 1992, following the crossing crack discovered at Dampierre 2 on the un-isolable part of a Safety Injection System pipe, a program of instrumentation was defined and is described in this paper. Among the objectives, two of the principal goals were to determine the thermal loadings really supported by the various lines and to highlight the thermal hydraulic phenomena affecting them. Indeed, in order to explain the discovered damages, it was essential to know the real thermal loadings to compare them with those of design and to carry out mechanical calculations of resistance to thermal fatigue. The instrumentations installed on the 900 MW units enabled to check the resistance with the fatigue of all the auxiliary lines in spite of significant differences between the real loadings and those envisaged at the design. They contributed to the knowledge improvement on the local thermal hydraulic phenomena but the incidents at Dampierre 1 showed that this knowledge is still imperfect. The results of these instrumentations are also used for the design of the future units by the use of the feedback of several cycles of acquisition on the 900 MW units, but also 1300 MW and 1450 MW since similar instrumentations were installed on the auxiliary lines in Golfech 2 and Chooz B1

  4. Failures modes in model feedwater heater tubing

    International Nuclear Information System (INIS)

    Steam extracted from the turbine is used to preheat the boiler feedwater in fossil-fuel-fired steam plants in order to improve thermal efficiency. This is accomplished in a series of heaters between the condenser hot well and the boiler. The heaters usually consist of a shell containing a bundle of U-bend tubes through which the feedwater is circulated. The heaters closest to the boiler handle water at high pressure and temperature. Because of the severe service conditions, high-pressure feedwater heaters are frequently tubed with drawn-and-stress relieved Monel 400, a nickel-base alloy containing 35 percent copper. As part of a study designed to reduce the rate of tube failure in high-pressure feedwater heaters, a number of failed drawn-and-stress-relieved Monel 400 U-bend tubes removed from three high-pressure feedwater heaters were examined at Battelle to determine the causes of failure. The results of this examination are described

  5. Residual heat removal in a PWR using a passive system

    International Nuclear Information System (INIS)

    The present work is made in the frame of the studies that are performed at the French Atomic Energy Commission on the innovative safety systems. The system which is discussed here is devoted to the residual heat removal. It can be used for a current french 3 loops PWR in place of the combination auxiliary feedwater system - atmospheric relief valve. A blackout transient, without auxiliary feedwater, is calculated, using the CATHARE code, in order to assess the capabilities of the system. Some complementary scenarios are calculated, assuming the intervention of other systems after a while, for example restart of the primary pumps and manual opening of the atmospheric relief valves. The influence of non condensable gases is also discussed. 7 refs., 17 figs

  6. 46 CFR 58.01-35 - Main propulsion auxiliary machinery.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Main propulsion auxiliary machinery. 58.01-35 Section 58... AUXILIARY MACHINERY AND RELATED SYSTEMS General Requirements § 58.01-35 Main propulsion auxiliary machinery. Auxiliary machinery vital to the main propulsion system must be provided in duplicate unless the...

  7. Derivation of the mass factors for decommissioning cost estimation of low contaminated auxiliary systems

    International Nuclear Information System (INIS)

    Ignalina NPP was operating two RBMK-1500 reactors. Unit 1 was closed at the end of 2004, and Unit 2 - at the end of 2009. Now they are under decommissioning. Decommissioning has been started from the reactor's periphery, with dismantling of non-contaminated and low contaminated equipment and installations. This paper discusses a methodology for derivation of mass factors for preliminary decommissioning costing at NPP when the number of inventory items is significant, and separate consideration of each inventory item is impossible or impractical for preliminary decommissioning plan, especially when the level of radioactive contamination is very low. The methodology is based on detailed data analysis of building V1 taking into account period and inventory based activities, investment and consumables and other decommissioning approach- related properties for building average mass factors. The methodology can be used for cost estimation of preliminary decommissioning planning of NPP auxiliary buildings with mostly very low level contamination. (authors)

  8. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  11. Controlling the feedwater flow in a BWR. Examples from Forsmark 2

    International Nuclear Information System (INIS)

    An investigation of the feedwater controller at Forsmark 2 has been performed. The investigation is based on signal analysis of measurement signals recorded during operation of the plant during different tests. The feedwater controller consists of the water level controller, the flow controller and the condenser balance controller. The overall goal of the feedwater control is to maintain constant water level (level controller) in the reactor and at the same time balance the water levels in the two condensers (condenser balance controller) to avoid that one condenser is full of water while the other one is operated with too low level. There is also a feed forward of the difference between steam flow and feedwater flow (flow controller) for each turbine system with the aim to reduce the fluctuation in reactor water level. The relation in strength between the three controllers is such that the level controller is the strongest followed by the condenser balance controller and finally the flow controller. Tests with trip of the feedwater pump and automatic start of the spare pump in each turbine system indicates a fast reduction in reactor water level that is restored after the transient in the control system. The transient in water level is stable without oscillations. However, it takes about 100 s before the reactor water level is restored. The function of the flow controller has been questioned by the authors. It does not take the action that is expected when a disturbance takes place in the difference between steam and feedwater flow. In addition to this principal weakness there is an offset in the feedwater controller output for feedwater flow 22 that reduces the contribution in flow control that is expected during the introduction of a disturbance. This offset should be adjusted during instrument maintenance of the feedwater controller. The PIP parameters for the level controller are gain factors and time constants. These have been evaluated with the aid of

  12. High-reliability feedwater heater design and construction

    International Nuclear Information System (INIS)

    Feedwater heater failures have significantly reduced the availability and thermal efficiency of fossil and nuclear power plants. Utilities must spend up to $1 million when replacing a high-pressure feedwater heater in a 400-MW fossil plant. An earlier Electric Power Research Institute report analyzed the causes of feedwater heater failure and identified six major problem areas. Improvements in feedwater heater design and construction should help correct these problems. This paper summarizes the root causes of feedwater heater problems. Design and construction features are recommended to avoid these problems. 6 refs., 10 figs

  13. Pilot application of risk informed safety margin characterization to a total loss of feedwater event

    International Nuclear Information System (INIS)

    In this paper we present the results of application of a risk-informed safety margin characterization (RISMC) approach to the analysis of a loss of feedwater (LOFW) event at a pressurized water reactor (PWR). This application considered a LOFW event with the failure of auxiliary feedwater (AFW) for which feed and bleed cooling would be required to prevent core damage. For this analysis the main parameters which impact core damage for the scenario were identified and probability distributions were constructed to represent the uncertainties associated with the parameter values. These distributions were sampled using a Latin Hypercube Sampling (LHS) technique to generate sets of sample cases to simulate using the MAAP4 code. Simulation results were evaluated to determine the safety margins relative to those obtained using typical probabilistic risk assessment (PRA) modeling (success criteria) assumptions. -- Highlights: • We apply a risk-informed approach to characterize safety margins. • This approach was applied to of a loss of feedwater event at a pressurized water reactor. • Probability distributions for the important parameter values were constructed. • These distributions were sampled to generate MAAP4 code simulation cases. • Simulation results were evaluated to determine the safety margins

  14. Teleporting an unknown quantum state with unit fidelity and unit probability via a non-maximally entangled channel and an auxiliary system

    Science.gov (United States)

    Rashvand, Taghi

    2016-08-01

    We present a new scheme for quantum teleportation that one can teleport an unknown state via a non-maximally entangled channel with certainly, using an auxiliary system. In this scheme depending on the state of the auxiliary system, one can find a class of orthogonal vectors set as a basis which by performing von Neumann measurement in each element of this class Alice can teleport an unknown state with unit fidelity and unit probability. A comparison of our scheme with some previous schemes is given and we will see that our scheme has advantages that the others do not.

  15. 46 CFR 61.20-3 - Main and auxiliary machinery and associated equipment, including fluid control systems.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Main and auxiliary machinery and associated equipment... SECURITY (CONTINUED) MARINE ENGINEERING PERIODIC TESTS AND INSPECTIONS Periodic Tests of Machinery and Equipment § 61.20-3 Main and auxiliary machinery and associated equipment, including fluid control...

  16. CPR1000非能动应急给水系统瞬态特性分析%Transient Analysis of Passive Emergency Feedwater System of CPR1O00

    Institute of Scientific and Technical Information of China (English)

    张亚培; 田文喜; 秋穗正; 苏光辉

    2011-01-01

    The transient thermal hydraulic characteristics of CPR1000 were analyzed by using RELAP5/MOD3. 4 code to verify the capability of the passive emergency feedwa-ter system (PEFWS) for accident mitigation under the condition of station blackout accident (SBO). The calculation results show that the PEFWS of CPR1000 can supply the water to steam generator immediately and remove the core residual heat after the SBO successfully, and it also shows that the design of the PEFWS of CPR1000 is successful.%利用RELAP5/MOD3.4程序对CPR1000压水堆在全厂断电事故下一回路主要参数的瞬态热工水力特性进行分析,验证CPR1000非能动应急给水系统(PEFWS)对事故的缓解能力.计算结果表明,CPR1000在发生全厂断电事故后,PEFWS完全可及时向蒸汽发生器补水,同时导出堆芯余热,保证反应堆处于安全状态,从而验证CPR1000 PEFWS的设计成功.

  17. Start-up of a power unit of a thermal power plant auxiliary system with supply from a hydropower plant

    OpenAIRE

    Zbigniew Lubośny; Krzysztof Dobrzyński; Jacek Klucznik

    2013-01-01

    This article discusses the issues related to a power unit of a thermal power plant start-up with the use of a hydropower plant. Hydropower plant can supply and will enable start-up of auxiliary equipment in a power unit of a thermal power plant. Due to high capacity of auxiliary drives, startup of auxiliaries in a thermal power plant after blackout (and boiler shutdown) is not possible from emergency energy sources in the power plant. In such a case an external electricity source with high ca...

  18. Start-up of a power unit of a thermal power plant auxiliary system with supply from a hydropower plant

    Directory of Open Access Journals (Sweden)

    Zbigniew Lubośny

    2013-09-01

    Full Text Available This article discusses the issues related to a power unit of a thermal power plant start-up with the use of a hydropower plant. Hydropower plant can supply and will enable start-up of auxiliary equipment in a power unit of a thermal power plant. Due to high capacity of auxiliary drives, startup of auxiliaries in a thermal power plant after blackout (and boiler shutdown is not possible from emergency energy sources in the power plant. In such a case an external electricity source with high capacity is required.

  19. Application of Relap-5 to the analysis of the Doel-2 steam generator tube rupture accident and studies of the Doel 1-2 loss of feedwater and feedwaterline break accidents

    International Nuclear Information System (INIS)

    By bilateral agreement between the US-NRC and TRACTIONEL, the Relap-5/Mod-2 code was obtained and installed on a CYBER-176 in March 1985. The code assessment work performed at TRACTIONEL is based mainly on operating plant data such as start-up and transients and complemented by data from the LOBI test facility. The simulation of the Doel 2 steam generator tube rupture accident by means of the mod-2 version of the code illustrates some important improvements over the mod-1 version such as smaller mass error and better performance while the comparison with plant data still shows some important shortcomings. Inadequate modeling of the heat and mass transfer for condensation and evaporation in stagnant flow conditions is probably responsible for the larger anomalies in the steam generator pressure and water level. The studies of a loss of feedwater and a feedwaterline break accident, using a best-estimate code with conservative initial and boundary conditions and identical nodalisation as the SGTR simulation, illustrates the high efficiency of the auxiliary feedwater system due to sustained thermal coupling between the primary and the secondary system, for low steam generator inventories

  20. Equipment for heat storage in feedwater for steam transformers

    International Nuclear Information System (INIS)

    The equipment consists of hot water storage tanks connected in parallel to feedwater heaters for steam transformers. One outlet connects the hot water storage anks via a first and a second valves to the supply pipe branch before the feedwater heaters while another outlet connects them via a third valve to the supply pipe branch after the feedwater heaters. The former storage tank outlet is connected via the first valve to the supply pump suction chamber while it is connected to the pump delivery part via the second valve. The cooled condensate outlet from the feedwater heater is connected to the condensate pipe before the regenerating economizers of the steam turbine while the first outlet of hot condensate from the steam transformer is connected via a fourth valve to the feedwater pipe after the regenerating economizers. A first outlet from the steam transformer is connected via a fifth valve to the outlet of the feedwater heater. (B.S.)

  1. Turbine generator and its auxiliaries

    International Nuclear Information System (INIS)

    The turbine generator and its auxiliary systems in Tarapur Atomic Power Station (TAPS) have been performing well and further their performance and availability has increased due to timely assessment of the problems anticipated in the systems by a close co-ordination among the concerned staff. Continued efforts are on for further improvements. (author)

  2. Modifications done in the IPR-R1 reactor and their auxiliary systems

    International Nuclear Information System (INIS)

    The improvements done in the IPR-R1 reactor for adequateness of operation conditions and increase of irradiation sample capability. The cooling systems, reactor pool, system of control rods were substituted. The optimization of transfer pneumatic system was done. (M.C.K.)

  3. Solar-auxiliary Coal-fired Power Generation System Thermal Economic Analysis%太阳能辅助燃煤发电系统经济性分析

    Institute of Scientific and Technical Information of China (English)

    葛晓霞; 邵娜; 邵成; 钱晨; 姜晨峰

    2015-01-01

    介绍了太阳能辅助锅炉受热面替代部分省煤器作用和太阳能辅助给水回热加热的两种发电系统。应用等效热降法对这两种太阳能辅助燃煤发电集成方案的热经济指标进行了计算与比较,选择了太阳能辅助给水回热加热为优化的集成方案。对槽式集热器的换热效率,光热电转换效率及投资节煤比3个技术经济性相关指标进行研究,在太阳能辅助给水回热加热的方案中,通过综合比较利用太阳能产生的汽替换各级抽汽的计算结果后,得出了替换第六级抽汽最为合理的结论。%Two power generation systems were introduced about solar assisted part replace of boiler economizer heating surface effects and the solar-assisted heating feed water regenerator .It was calculated and compared these two types of solar assisted heat economic indicators coal-fired integrated solutions by using of Equivalent Heat Drop .It was selected a solar-assisted water heating for the optimization of regenerative integrated solution ., Three related indicators of technical and economic were studied On heat transfer efficiency of trough collector and the light thermoelectric conversion efficiency as well as investment in coal saving ratio .In the solar thermal heating auxiliary feedwater back scheme , By comprehensive comparison of the use of solar energy to produce steam to replace the calculation of results at all levels extraction .It was obtained the most reasonable conclusion of replacing sixth stage extraction .

  4. Optimal Scheduling of a Battery Energy Storage System with Electric Vehicles’ Auxiliary for a Distribution Network with Renewable Energy Integration

    Directory of Open Access Journals (Sweden)

    Yuqing Yang

    2015-09-01

    Full Text Available With global conventional energy depletion, as well as environmental pollution, utilizing renewable energy for power supply is the only way for human beings to survive. Currently, distributed generation incorporated into a distribution network has become the new trend, with the advantages of controllability, flexibility and tremendous potential. However, the fluctuation of distributed energy resources (DERs is still the main concern for accurate deployment. Thus, a battery energy storage system (BESS has to be involved to mitigate the bad effects of DERs’ integration. In this paper, optimal scheduling strategies for BESS operation have been proposed, to assist with consuming the renewable energy, reduce the active power loss, alleviate the voltage fluctuation and minimize the electricity cost. Besides, the electric vehicles (EVs considered as the auxiliary technique are also introduced to attenuate the DERs’ influence. Moreover, both day-ahead and real-time operation scheduling strategies were presented under the consideration with the constraints of BESS and the EVs’ operation, and the optimization was tackled by a fuzzy mathematical method and an improved particle swarm optimization (IPSO algorithm. Furthermore, the test system for the proposed strategies is a real distribution network with renewable energy integration. After simulation, the proposed scheduling strategies have been verified to be extremely effective for the enhancement of the distribution network characteristics.

  5. Troubleshooting and treatment of declining boiler feedwater quality

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Qiuyang; Qian, Zhouhai; Zhu, Liwei [Zheijiang Electric Company, Hangzhou (China). Electric Power Research Inst.

    2013-03-15

    The conductivity of aqueous solutions is one of the most commonly used analytic readings in the field of water chemistry. In thermal power plants, to avoid the interference of ammonia, cation conductivity is used to show the purity of feedwater, boiler water, and steam instead of specific conductivity. The cation conductivity of the feedwater of Unit 2 at Changxing Power Plant increased substantially and the sodium content in the feedwater fluctuated frequently. This paper describes the process of troubleshooting and treating the deteriorating feedwater quality of Unit 2 at Changxing Power Plant. (orig.)

  6. Solutions of system of P1 equations without use of auxiliary differential equations coupled

    International Nuclear Information System (INIS)

    The system of P1 equations is composed by two equations coupled itself one for the neutron flux and other for the current. Usually this system is solved by definitions of two integrals parameters, which are named slowing down densities of the flux and the current. Hence, the system P1 can be change from integral to only two differential equations. However, there are two new differentials equations that may be solved with the initial system. The present work analyzes this procedure and studies a method, which solve the P1 equations directly, without definitions of slowing down densities. (author)

  7. Design and Development of a Small Heat Exchanger as Auxiliary Cooling System for Domestic and Industrial Applications

    Directory of Open Access Journals (Sweden)

    L.O.Ogunleye

    2013-11-01

    Full Text Available The epileptic supply of power from the national grid in Nigeria has made many industries to engage Internal Combustion Engine generators as alternative to providing energy required for production. The excessive use of these machines has mostly altered their effective performance, thereby necessitating more frequent maintenance or repair than recommended by the manufacturers. Frequent break-downs of these machines reduce rate of production of these industries and by extension, this adversely affects the economy development of the country. A known engineering enterprise in Kano; North West region of Nigeria due to the same factor stated above, subjected her 30 kvagenerator to run almost throughout the working hours of the week. Initially, the generator run perfectly within the manufacturer recommended 100 hours of operation before conducting maintenance works. After sometime, due to excessive use, the generator hardly met half the required service hours before overheating and this resulted in frequent damage of the gasket and repair of the valve outlets, consequently increased the cost of maintenance. A Small Tube and Shell Heat Exchanger with parallel/counter flow that would serve as an auxiliary cooling system for the radiator was designed and developed.

  8. BIOFEAT: Biodiesel fuel processor for a vehicle fuel cell auxiliary power unit. Study of the feed system

    Science.gov (United States)

    Sgroi, M.; Bollito, G.; Saracco, G.; Specchia, S.

    An integrated auxiliary power unit (APU) based on a 10 kW e integrated biodiesel fuel processor has been designed and is being developed. Auto-thermal reforming (ATR) and thermal cracking (TC) were considered for converting the fuel into a hydrogen-rich gas suitable for PEM fuel cells. The fuel processor includes also a gas clean-up system that will reduce the carbon monoxide in the primary processor exit gas to below 10 ppm via a new heat-integrated CO clean-up unit, based on the assembly of catalytic heat exchange plates, so as to meet the operational requirements of a PEMFC stack. This article is devoted to the study and selection of the proper feed strategy for the primary fuel processor. Different pre-treatment and feed alternatives (e.g. based on nozzles or simple coils) were devised and tested for the ATR processors, which turned out to be the preferred primary processing route. A nozzle-based strategy was finally selected along with special recommendations about the constituent materials and the operating procedures to be adopted to avoid coking and nozzle corrosion as well as to allow a wide turn down ratio.

  9. Dispersant application: (1) during steam generator wet layup for removal of existing deposits, and (2) during the long-path recirculation cleanup process of the condensate/feedwater system to reduce startup corrosion product transport to the steam generators

    International Nuclear Information System (INIS)

    management operations (e.g., sludge lancing, chemical cleaning, etc.). In 2009 EPRI worked with Exelon Corporation to develop and implement a plan for a trial dispersant application during the SG wet layup at Three Mile Island Unit 1 (TMI-1). Because the SGs were scheduled for replacement during the Fall 2009 outage, this trial represented a unique opportunity to evaluate the efficacy of a dispersant wet layup application with minimal risk to the SGs. This paper discusses the technical bases supporting the addition of dispersant during wet layup at TMI-1 and the results of the Fall 2009 trial application. Additional applications, under EPRI sponsorship, have either just taken place (Doel 3 in Summer 2010) or are planned (Braidwood 1 in Fall 2010). It is anticipated that addition of a dispersant during the long-path recirculation cleanup process will more readily clean up transportable corrosion products from the system and increase their retention time in solution, thereby increasing the amount of iron removed from the condensate and feedwater systems prior to initiation of flow to the SGs. Evaluations preparing for applications of this type include: Reviews of the procedure and general characteristics of the long-path recirculation cleanup process at three plants: Byron Unit 1, Millstone Unit 2 and Three Mile Island (TMI) Unit 1; A laboratory test program to assess dispersant efficacy under the conditions present during the long-path recirculation cleanup process. A number of different dispersant chemicals, including the polyacrylic acid (PAA) used in online applications, were investigated; Generic and plant-specific evaluations (for the above three units) of the compatibility of PAA with the secondary system components anticipated to be wetted during the long-path recirculation cleanup process. Based on the results of this study, inputs concerning the recommended dispersant chemical, concentration, application schedule, and cleanup criteria were generated to aid interested

  10. Simplified analysis of PRISM RVACS [Reactor Vessel Auxiliary Cooling System] performance without liner spill-over

    International Nuclear Information System (INIS)

    Simplified analysis of the performance of the PRISM RVACS decay heat removal system under off-normal conditions, i.e., without the liner spill-over, is described. Without the spilling of hot-pool sodium over the liner and the resultant down-flow along the inside of the reactor vessel wall, the RVACS system performance becomes dominated by the radial heat condition and radiation. Simple estimates of the resulting heat conduction and radiation processes support GE's contention that the RVACS performance is not severely impacted by the absence of spillover, and can improve significantly if sodium has leaked into the region between the reactor and containment vessels. 7 refs

  11. Auxiliary control system of the safety parameters for IPR-R1 reactor

    International Nuclear Information System (INIS)

    This paper deals with the description of the control of three cooling water parameters, as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, a permanent and accurate control of the cooling water is needed. This is achieved through this system, which allows the simultaneous measurement of the water parameters such as: conductivity, temperature and the maximum and minimum water levels. The monitoring of a fourth parameter, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author)

  12. Auxiliary control system of the safety parameters for IPR-R1 reactor

    International Nuclear Information System (INIS)

    This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author)

  13. 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...

  14. Support on water chemistry and processes for nuclear power plant auxiliary systems

    International Nuclear Information System (INIS)

    In particular PHWRs have a system devoted to the purification and upgrading of the collected heavy water leaks. The purification train is fed with different degradation ratios (D2O/H2O), activities and impurities. The water is distilled in a packed bed column filled with a mesh type packing. The mesh wire is made of a bronze substrate covered by copper oxides whose current composition has been determined by Moessbauer spectroscopy. With the purpose of minimizing the column stack corrosion, the water is pre-treated in a train consisting of an activated charcoal bed-strong cationic-anionic resin and a final polishing mixed bed resin. Ionic chemicals like acetic acid (whose provenance is suspected to come from the air treatment/D2O recovery system where the regeneration is performed at high temperature) are detected by the conductivity and ion chromatography when they concentrate at the column bottom. Traces of oils are retained by the charcoal bed but some compounds extracted by the aqueous phase are suspected to be responsible for the resins fouling or precursors of potentially aggressive agents inside the distillation column. Those species have been detected and identified by gaseous chromatography-mass spectrometry (GC-MS). In the present work, the identification, evaluation of alternatives for the retention and results compared to the original products present in the water upgrading purification train have been summarized. (authors)

  15. Cost Effective, High Efficiency Integrated Systems Approach To Auxiliary Electric Motors

    Energy Technology Data Exchange (ETDEWEB)

    Roy Kessinger; Kanchan Angal; Steve Brewer; Steve Kraihanzel; Lenny Schrank; Jason Wolf

    2003-07-15

    The CARAT program, carried out by Kinetic Art & Technology Corporation (KAT), has been one of the most commercially successful KAT R&D programs to date. Based on previous development of its technology, KAT designed, constructed and tested a highly efficient motor and controller system under this CARAT program with supplemental commercial funding. Throughout this CARAT effort, the technical objectives have been refined and refocused. Some objectives have been greatly expanded, while others have been minimized. The determining factor in all decisions to refocus the objectives was the commercial need, primarily the needs of KAT manufacturing partners. Several companies are employing the resulting CARAT motor and controller designs in prototypes for commercial products. Two of these companies have committed to providing cost share in order to facilitate the development. One of these companies is a major manufacturing company developing a revolutionary new family of products requiring the ultra-high system efficiency achievable by the KAT motor and controller technologies (known as Segmented ElectroMagnetic Array, or SEMA technology). Another company requires the high efficiency, quiet operation, and control characteristics afforded by the same basic motor and controller for an advanced air filtration product. The combined annual production requirement projected by these two companies exceeds one million units by 2005.

  16. Checking earthquake resistance of feedwater pump

    International Nuclear Information System (INIS)

    The dynamic behavior of rotors of feedwater pumps for WWER-1000 nuclear power plants was analyzed with respect to seismic excitation. Mathematical determination of the transient state and of the maximum amplitude of the forced rotor vibration is described in detail. The results of the analysis give evidence that the pump rotor is earthquake-resistant. Its first critical revolutions of bending-circulation oscillations have a value of 75 Hz, which is well above the seismic excitation band. The forced vibration amplitude does not exceed the value of 0.02 mm, which is lower than the clearance of the runner packing rings. (Z.M.). 4 figs., 5 refs

  17. Definition of an auxiliary processor dedicated to real-time operating system kernels

    Science.gov (United States)

    Halang, Wolfgang A.

    1988-01-01

    In order to increase the efficiency of process control data processing, it is necessary to enhance the productivity of real time high level languages and to automate the task administration, because presently 60 percent or more of the applications are still programmed in assembly languages. This may be achieved by migrating apt functions for the support of process control oriented languages into the hardware, i.e., by new architectures. Whereas numerous high level languages have already been defined or realized, there are no investigations yet on hardware assisted implementation of real time features. The requirements to be fulfilled by languages and operating systems in hard real time environment are summarized. A comparison of the most prominent languages, viz. Ada, HAL/S, LTR, Pearl, as well as the real time extensions of FORTRAN and PL/1, reveals how existing languages meet these demands and which features still need to be incorporated to enable the development of reliable software with predictable program behavior, thus making it possible to carry out a technical safety approval. Accordingly, Pearl proved to be the closest match to the mentioned requirements.

  18. Improved plant economics through accurate feedwater flow measurement with the crossflow ultrasonic flowmeter

    International Nuclear Information System (INIS)

    The crossflow ultrasonic flowmeter (UFM) improves nuclear power plant performance through more accurate and reliable feedwater flow measurement. Reactor power levels are typically monitored via secondary-side calorimetric calculations that depend on the accurate measurement of feedwater flow . The feedwater flow is measured with calibrated venturis in most plants. These are subject to chemical fouling and other mechanical problems. If the loss in accuracy of the feedwater flow measurement overstates the actual flow rate, the result is a direct loss in megawatts generated by the plant. This paper describes a new, innovative ultrasonic technique to improve the accuracy, stability and repeatability of ultrasonic flow measurements. By employing this advanced technology to provide a continuous correction to the venturi-measured feed water flow rate, plants have reported the recovery of between 5 and 25 MWe. This technology has been implemented in a new flowmeter called CROSSFLOW. The CROSSFLOW meter utilizes a mathematical process called cross-correlation to process the ultrasonic signal, which is modulated by the flow eddys to determine the velocity of the feedwater. It replaces the older, less accurate transit-time methodology. Comparisons with weigh tank test, calibrated plant instrumentation, and chemical tracer tests have demonstrated a repeatable accuracy of 0.21% or better with this advanced cross-correlation technology. The paper discusses the history of the cross-correlation technique and its theoretical basis, illustrates how this technique addresses the measurement sensitivities for various parameters, demonstrates the calculation of the accuracy of the meter, and discusses the recently completed NRC review of the CROSSFLOW System and methodology. The paper also discusses recent precision flow measurement applications being performed with CROSSFLOW at nuclear plants worldwide. Among these applications are the measurement of Reactor Coolant System flow and the

  19. 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)

    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)

  20. 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.

  1. Optimization of relay protection for a auxiliary power system%厂用电系统继电保护优化

    Institute of Scientific and Technical Information of China (English)

    李子峰

    2015-01-01

    基于对厂用电系统继电保护中存在的配置不完整、后备保护动作时间过长等问题的分析,提出了配置6 kV 母线专用主保护、在低压厂用电系统变压器高压侧增加限时电流速断保护装置、优化后备保护之间配合的方案,并以此对厂用电系统的继电保护配置进行了优化,以600 MW机组为例,进行了保护整定计算,结果表明,优化后的厂用电系统继电保护配置完整,保护范围合理,后备保护动作时间显著缩短。从而,提高了厂用电系统继电保护动作的快速性和厂用电系统运行安全性。%On the basis of analysis on some questions of relay protection in auxiliary power systems,like the imperfection in configuration and the actuation time is too long,a scheme was proposed for relay protection design,such as configuring a specialized main protection for 6 kV busbar,adding a time limited current fast-trip protection on the high-voltage side of low-voltage transformers in auxiliary power system,optimizing the cooperation of reserve protections and then designing the configuration of relay protection in auxiliary power system according to the optimizing rule.Taking a 600 MW unit as the example,settings calculation was carried out.The results proved that the configuration of relay protection in auxiliary power system is integrated perfectly due to the optimization,the coverage of protection is reasonable and the actuation time of reserve protection has been remarkably shortened.Thus,the actuation speed and security of the auxiliary power system are promoted.

  2. The Marginalized Auxiliary Particle Filter

    OpenAIRE

    Fritsche, Carsten; Schön, Thomas; Klein, Anja

    2010-01-01

    In this paper we are concerned with nonlinear systems subject to a conditionally linear, Gaussian sub-structure. This structure is often exploited in high-dimensional state estimation problems using the marginalized (aka Rao-Blackwellized) particle filter. The main contribution in the present work is to show how an efficient filter can be derived by exploiting this structure within the auxiliary particle filter. Based on a multisensor aircraft tracking example, the superior performance of the...

  3. Theoretic analysis on separation efficiency of wire mesh mist eliminator of high-temperature gas-cooled reactor helium purification and auxiliary system

    International Nuclear Information System (INIS)

    Helium purification and helium auxiliary system is one of important systems guaranteeing the safe operation of high-temperature gas-cooled reactor. Wire mesh mist eliminator in this system is one of the key components. It is used to separate waste water containing tritium, and remove moisture after reactor accident. Base on the ideal fluid model and packing pad model developed by Carpenter, a calculation model was presented for separation efficiency of mist eliminator. The calculation program SEP-WMME was developed based on the model. The calculation results fit well with experiment results. Theoretic analysis was carried out for the mist eliminator of regeneration system in HTR-PM helium purification system engineering validation test loop. The analysis results show that the inlet velocity is an important parameter for mist eliminator in regeneration system. When the inlet velocity is above 3.0 m/s, high separation efficiency will be obtained. The number of wire mesh layers also affects the separation efficiency remarkably. When the number of layers increases further to some extent, the separation efficiency increase becomes insignificant. The number of layers should be chosen properly by considering pressure loss. Additionally, the diameter of wire is an important parameter related to separation efficiency. The separation efficiency increases with the decrease of the wire diameter. The analysis is significant for structure design, optimization and safe operation of mist eliminator in helium purification and helium auxiliary system. (authors)

  4. Study on application of T-S fuzzy neural method in once through steam generator feedwater control

    International Nuclear Information System (INIS)

    Considering the problems of small water inventory and inferior heat accumulation capacity of the secondary side of the once through steam generator (OTSG) , and the need of rapid match of feedwater flowrate and the reactor power to ensure the steam quality, the T-S fuzzy neural method is incorporated into the OTSG feedwater control system to design a T-S fuzzy neural controller based on input and output data acquired by numerical simulation. The control results are verified by simulation method. The simulation results show that the control results under this control manner are better than those under the PID control manner and meet the control requirements satisfactorily. (authors)

  5. Optimal control strategy in the design of a PWR feedwater controller

    International Nuclear Information System (INIS)

    A design approach and available preliminary results are presented of work currently in progress to develop a modern alternative to the classical (present day) three-element analog steam generator water level control system for pressurized water reactors. The new control scheme is a digital feedwater controller design based on concepts of modern control theory. A basic approach is outlined which can be readily understood by industry engineers briefly acquainted with the modern control theory. A nonlinear dynamic steam generator model is developed to reasonably simulate the transient behavior of steam generator water level during typical control system transients. Deterministic optimal linear regulator theory is used to develop a multivariable feedwater controller for the steam generator and feedwater system linearized about several operating points. This controller design is now directly applicable for high power control of dry and saturated feedring-type steam generators only. Transient control performance of the feedback portion of the optimal controller is tested in computer simulation with a high-order nonlinear full plant model. Preliminary results show that the design has potential for better control performance than that of the present day control system

  6. 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....

  7. 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.

  8. Development of A Conservative Method for A Feedwater Pipe Break Analysis of An Integral Type Reactor

    International Nuclear Information System (INIS)

    The development of advanced small and medium sized nuclear power plants for multipurpose appears before the footlights, and some of them are ready for construction. The SMART, which is an integral pressurized water reactor is one of those advanced types of small sized nuclear reactors. The basic design of SMART was completed at the Korea Atomic Energy Research Institute. A new phase in order to test and verify the SMART design is currently underway in Korea. The results of these tests and verifications will be fed back into the SMART design for a further improvement of the safety and reliability. The integral type reactor can be mitigated design basis events by a reactor protection system, or engineered safety features. The consequences of design basis events must be less than the established acceptance limits and provide an acceptable margin to protect the health and safety. The design basis events are divided into general categories corresponding to their effect on a plant. One of these categories is a decrease in a heat removal by the secondary system. There are a turbine trip, a main steam isolation valve closure, a loss of the primary component cooling system, and a feedwater pipe break for the decrease in the heat removal by the secondary system. The feedwater pipe break accident is one of the most important accidents in the safety of the integral type reactor. Decrease in the feedwater supply to the steam generators causes a decrease in the heat extraction rate from the reactor coolant system, resulting in an increase of a primary coolant temperature and a pressure, and the nuclear power decreases due to a reactivity feedback. Performed sensitivity analysis to find parameters affecting seriously in the integral reactor's feedwater pipe break accident. According to these parametric analysis results, a power level, an initial system pressure, a moderator reactivity coefficient and a break size are major parameters for the maximum system pressure. The detailed

  9. Auxiliary building structures

    International Nuclear Information System (INIS)

    Five types of auxiliary structures are described such as were used during the construction of the Dukovany nuclear power plant, namely a portable staircase tower, a stable staircase tower, mobile tower scaffolding, mobile scaffolding on a crane track and a scaffold cradle. Basic technical data for all types of scaffolding are given. (Pu)

  10. Performance evaluation of a state-of-the-art solar air-heating system with auxiliary heat pump. Final report, 1 October 1978-30 September 1979

    Energy Technology Data Exchange (ETDEWEB)

    Karaki, S.; Brisbane, T.E.; Waterbury, S.S.; Lantz, T.G.

    1980-01-01

    The solar heating system in Solar House II, evaluated during the heating season of 1978-1979, consists of 57.9 m/sup 2/ of Solaron Series 3000 Collectors, 10.3 m/sup 3/ 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. Descriptions of the building and the system, modes of system operation, the data acquisition system, data processing, and performance - thermal, collector, storage, and heat pump - are included. (MHR)

  11. Optimizationin Operational Analysis of Auxiliary Steam System in Thermal Power Plant%火力发电厂辅汽系统优化运行分析

    Institute of Scientific and Technical Information of China (English)

    王乃军; 白秀春; 王俊俊

    2015-01-01

    For choosing different steam source for the auxiliary steam system in Inner Mongolia Daihai Electric Power Generation Co.,Ltd., carry on the quantitative analysis to the influence of the unit efficiency, the result is the most economical when choosing No.4 extraction steam of second unit supplying, while slightly economical as using second unit cold reheater as first unit. According to the results of the analysis, it provides a theoretical basis and guidance for selecting the source of auxiliary steam in power plants, so as to achieve the purpose of saving energy and reducing consumption, improve the operation efficiency. At the same time, based on the actual operating mode, unit start-up and shutdown, accident conditions, combined with different seasons and temperature, auxiliary steam consumption, it proposes operation precautions.%对内蒙古岱海发电有限责任公司辅助蒸汽系统选择不同供汽汽源时,机组运行经济性受到的影响进行了定量分析,认为由二期机组四段抽汽供汽经济性最优,由二期机组冷段再热蒸汽供汽的经济性略优于由一期机组冷段再热蒸汽供汽。并结合机组实际运行方式、机组启停方式及机组发生事故时的运行情况等,根据不同季节环境温度及各辅汽用户用汽量情况,提出辅汽系统优化运行注意事项。

  12. Methodology for carrying out energy diagnosis in auxiliaries systems in thermal electrical central stations; Metodologia para realizar un diagnostico energetico en sistemas auxiliares de centrales termoelectricas

    Energy Technology Data Exchange (ETDEWEB)

    Nebradt Garcia, Jesus [Comision Federal de Electricidad (CFE), Mexico, D. F. (Mexico); Rojas Hidalgo, Ismael; Huante Perez, Liborio [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1998-12-31

    One of the potential areas for energy saving in Central Electric Power Plants are the auxiliaries system, so as to based in a preliminary energy diagnosis and considering that energy saving measures would be taken, going from the instrumentation, operational changes in equipment, as well as in using velocity variators in motors, it turns out to be that the energy consumption of auxiliaries at 75% load in a 150 MW thermal power plant varies from 3% to 4% and for the case of a 350 MW power plant the energy consumption of the auxiliaries represents 2 to 3.5%. Nowadays this consumption are above 6%. Considering that the country has 40 units with capacities varying from 150 to 350 MW, the economical and the fuel saving would be substantial. This paper will present a summary of the methodology to be used to carry out this type of projects. [Espanol] Una de las areas potenciales de ahorro de energia en centrales termoelectricas son los sistemas auxiliares, de tal manera que basados en un diagnostico energetico preliminar y considerando que se aplicarian las medidas de ahorro de energia que van desde la instrumentacion, cambios operativos en equipos, asi como el uso de variadores de velocidad en motores, se tienen que los consumos de auxiliares para un 75% de carga en una central termoelectrica de 150 MW varian desde un 3% hasta un 4% y para el caso de una central termoelectrica de 350 MW, el consumo de auxiliares representa del 2 al 3.5%. Hoy en dia dichos consumos estan por encima del 6%. Si consideramos que el pais cuenta con 40 unidades que varian desde 150 MW hasta 350 MW, entonces los ahorros economicos y de combustible serian impactantes. La presente ponencia mostrara un resumen de la metodologia a emplear para la realizacion de este tipo de proyectos.

  13. Total loss of feedwater analysis of UCH 3 and 4

    International Nuclear Information System (INIS)

    The Ulchin nuclear unit 3 and 4 have a safety depressurization system (SDS) to mitigate the beyond design basis event of the total loss of feedwater (TLOFW). The SDS provides a manual means of rapidly depressurizing the reactor coolant system, (RCS) for the highly unlikely event of a TLOFW to both steam generators. The reduced RCS pressure allows high pressure safety injection (HPSI) flow to replenish and eventually exceed the mass flow rate out through the SDS prior to uncovering the core with two feet margin. Thus, extended decay heat removal is provided by a feed (HPSI) and bleed (SDS) process. The purpose of this report is to assess the UCN 3 and 4 SDS bleed capability in mitigating the TLOFW, and to demonstrate that the SDS would prevent core uncovery. This report presents the results of a best-estimate analysis of the TLOFW event with feed and bleed. Also presented are brief descriptions of the SDS design basis, system configuration, and analytical tools and methods. For the best-estimate analysis of the TLOFW event, the CEFLASH-4AS/REM code, that had been verified against experimental data, was used. The results of analysis based on the minimum SDS flow capacity demonstrated that the SDS in conjunction with the HPSI system would prevent core uncovery with two feet margin even if a single failure is assumed, if initiated at the time of pressurizer safety valves lift following TLOFW. 3 tabs., 25 figs., 10 refs. (Author) .new

  14. RETRAN analysis results of feedwater pump trip transient for Lungmen ABWR Plant

    International Nuclear Information System (INIS)

    Highlights: → The RETRAN model was used to predict one feedwater pump trip (FWPT) transient. → The result shows that the margin sustains at least 30 cm above the L3 setpoint. → The unavailable motor driven pump case eventually actuates the low level scram signal. → The lowest load line case without motor driven pump still actuates the L3 scram. - Abstract: The RETRAN model of Lungmen ABWR was used to simulate one feedwater pump trip (FWPT) transient of the Lungmen start-up test program. The purpose of this test is to verify the capability of one surviving Turbine Driven Reactor Feedwater Pump (TDRFP) plus a Motor Driven Feedwater Pump (MDRFP) to continue operating the reactor stably following the incident. There are three major control systems implanted in Lungmen RETRAN model (LRM), which include Recirculation Flow Control System (RFCS), Steam Bypass and Pressure Control System (SBPCS), and Feedwater Control System (FWCS). The reactor water level margin with respect to the low level scram setpoint in the transient is monitored to confirm whether the acceptance criteria has been satisfied, which depends on the responses of the control systems to the FWPT transient. The analysis result of base case at 100% rated power/100% rated core flow with automatic start of MDRFP demonstrates that the acceptance criteria are met, which shows that the water level still sustains ample margin of 30 cm above the low level setpoint, and the reactor does not scram. To get more insight into the function of MDRFP, a set of sensitivity studies with the assumption of unavailable MDRFP, and with a different initial condition which extended to the maximum allowable core flow of 111% rated at rated power, was conducted to verify the superior capability of power coastdown due to the RIPs runback logic under the lowest load line, and also the delay time of the Reactor Internal Pumps (RIPs). Finally, it is concluded that FWPT transient without start of MDRFP eventually actuates the low

  15. 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.

  16. Dental Auxiliary Occupations. Interim Report.

    Science.gov (United States)

    Kingston, Richard D.

    As part of a dental auxiliaries project, a Dental Auxiliary National Technical Advisory Committee was established, and its major undertaking was to assist in the development of a functional inventory for each of the three dental auxiliary occupations (dental assisting, dental hygiene, and dental laboratory technology). The analysis consisted of…

  17. Comparative Analysis on Several Operation Schemes for Raw Water Heater of Boiler Feedwater Treatment System%几种锅炉补给水处理系统原水加热器运行方案的分析比较

    Institute of Scientific and Technical Information of China (English)

    童晓凡; 潘灯

    2016-01-01

    Due to low temperature of raw water in boiler feedwater treatment system, high-quality steam is used for heating that consumes too much energy. Therefore, the paper proposes to use waste heat from power plant circulating cooling water to heat raw water of boiler feedwater treatment system. The paper analyzes four operation schemes for raw water heater system, aiming to recycle part of the waste heat. Through technical and economic comparison, the paper presents a scheme of combing hot water heater with steam heater to save energy, reduce emissions as well as meet requirements of different operating conditions. The scheme is char-acterized by its higher integrated index of economical efficiency and performance.%针对目前锅炉补给水处理系统因原水温度低,而利用高品质蒸汽加热耗费能源的现象,提出利用发电厂循环冷却水的余热,加热锅炉补给水处理系统的原水。分析了4种原水加热器系统运行方案,目的在于回收部分废弃的热量。通过技术和经济性比较,给出采用热水加热器和蒸汽加热器配合使用方案,达到节能减排,同时满足各种运行工况需求,综合经济性能指标较高。

  18. Fracture of feedwater pipe in Surry unit-2 reactor

    International Nuclear Information System (INIS)

    A trip occurred in the Surry unit-2 reactor at 14 : 21 on December 9, 1986, when the reactor was being operated at full power. Recently, the power company published a detailed report on the accident and an inspection report was released by the U.S. Nuclear Regulator Commission (NRC), suggesting that the fracture had resulted from corrosion in an elbow. Thus, the present report discusses in detail the rupture of the elbow, which was part of the secondary cooler feedwater pipe. The rupture was of the guillotine mode and took place instantaneously. Immediately after the reactor trip started, the three main steam valve systems closed and then the rupture of the elbow occurred. Visual examination has shown that the wall of the elbow had become smaller in thickness. A dimple pattern was found in the fracture surface, indicating that the rupture was of the ductile breaking type. It has been concluded that this extreme decrease in wall thickness was the result of corrosion. An ultrasonic test has detected some local pitting corrosion portions in the neighborhood of the longitudinal welding seam. No unusual microscopic metal structures were detected by microscopy. Macroscopic observation revealed that small defects had been formed along the fracture surface. (Nogami, K.)

  19. Nuclear reactors with auxiliary boiler circuit

    International Nuclear Information System (INIS)

    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

  20. 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.

  1. Auxiliary power unit for moving a vehicle

    Science.gov (United States)

    Akasam, Sivaprasad; Johnson, Kris W.; Johnson, Matthew D.; Slone, Larry M.; Welter, James Milton

    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.

  2. Controlling the feedwater flow in a BWR. Examples from Forsmark 2; Regleringen av matarvattenfloedet i en BWR. Med exempel fraan Forsmark 2

    Energy Technology Data Exchange (ETDEWEB)

    Bergdahl, Bengt-Goeran; Oguma, Ritsuo (GSE Power Systems AB, Nykoeping (Sweden))

    2009-03-15

    An investigation of the feedwater controller at Forsmark 2 has been performed. The investigation is based on signal analysis of measurement signals recorded during operation of the plant during different tests. The feedwater controller consists of the water level controller, the flow controller and the condenser balance controller. The overall goal of the feedwater control is to maintain constant water level (level controller) in the reactor and at the same time balance the water levels in the two condensers (condenser balance controller) to avoid that one condenser is full of water while the other one is operated with too low level. There is also a feed forward of the difference between steam flow and feedwater flow (flow controller) for each turbine system with the aim to reduce the fluctuation in reactor water level. The relation in strength between the three controllers is such that the level controller is the strongest followed by the condenser balance controller and finally the flow controller. Tests with trip of the feedwater pump and automatic start of the spare pump in each turbine system indicates a fast reduction in reactor water level that is restored after the transient in the control system. The transient in water level is stable without oscillations. However, it takes about 100 s before the reactor water level is restored. The function of the flow controller has been questioned by the authors. It does not take the action that is expected when a disturbance takes place in the difference between steam and feedwater flow. In addition to this principal weakness there is an offset in the feedwater controller output for feedwater flow 22 that reduces the contribution in flow control that is expected during the introduction of a disturbance. This offset should be adjusted during instrument maintenance of the feedwater controller. The PIP parameters for the level controller are gain factors and time constants. These have been evaluated with the aid of

  3. Evaluation of piping integrity in thinned main feedwater pipes

    International Nuclear Information System (INIS)

    Significant wall thinning due to flow accelerated corrosion(FAC) was recently reported in main feedwater pipes in 3 Korean pressurized water reactor (PWR) plants. The main feedwater pipes in one plant were repaired using overlay weld method at the outside of pipe, while those in 2 other plants were replaced with new pipes. In this study, the effect of the wall thinning in the main feedwater pipes on piping integrity was evaluated using finite element method. Especially, the effects of both the overlay weld repair and the stress concentration in notch-type thinned area on the piping integrity were investigated. The results are as follows; (1) The piping load carrying capacity may significantly decrease due to FAC. In special, the load carrying capacity of the main feedwater pipes was reduced by about 40 percent during about 140 months operation in Korean PWR plants. (2) By performing overlay weld repair at the outside of pipe, the piping load carrying capacity can increase and the stress concentration level in the thinned area can be reduced. (author)

  4. 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.

  5. MATLAB GUI Design of Auxiliary Teaching System for Multivariate Statistics Course%多元统计课程教辅系统的 MATLAB GUI 设计

    Institute of Scientific and Technical Information of China (English)

    周志刚; 陈丽红

    2013-01-01

      针对目前多元统计方法课程教学重点内容涉及大量数据的统计处理难以在课堂上现场演示,学生接受多元统计知识感觉困难、枯燥的问题,文章利用 MATLAB GUI (Graphical User Interfaces,GUI)设计了一个多元统计方法课程教学辅助系统,用于多元统计方法课程的教学和学生实验。教学辅助系统的设计从教学层面和系统设计的技术层面进行了探讨,设计的教学辅助系统具有使用较方便、良好交互性和可扩展性的特点。同时 MATLAB GUI 开发环境开发出的教辅系统便于学生做二次开发,可以提高学生动手实践能力及多元统计方法课程的教学质量。%In view of the problems in the traditional multivariate statistics course teaching,which contain it is difficult to demonstrate in class about principle of key teaching content and statistic treatment of large quantities of data,students feel difficult to receive multiple sta-tistical knowledge,a multivariate statistics course teaching auxiliary system is designed by using MATLAB GUI (Graphical User Inter-faces,GUI). The system is used for multivariate statistical methods course teaching and students' experiment. Auxiliary teaching system was discussed from two aspects about teaching and technology and has advantages of convenient use,good interactivity and expansibility. At the same time,this system also facilitate students to do the secondary development,can improve students' practical ability and multiva-riate statistical methods course teaching quality.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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. 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.

  9. Economic Optimization of Solar-Electric Auxiliary Heating System%太阳能-电辅热供暖系统的经济优化

    Institute of Scientific and Technical Information of China (English)

    程磊; 田琦

    2011-01-01

    集热器面积的选择是太阳能供暖系统设计的关键,本文对太阳能一电辅热供暖系统进行了技术经济分析,建立了系统费用年值与集热器面积函数关系的数学模型,分析探讨了使费用年值最小的集热器面积的计算方法。并以太原为例(100m^2住宅)逐日进行了计算,结果表明当集热器面积为21.34m^2,整个系统的费用年值最小;整个采暖季的太阳能保证率为60.24%。%Collector area is the key to a solar heating system while designing. In this paper, the technical-economic analysis of a solar-electric auxiliary heating system was carried. The mathematical model of the functional relation between the annual system cost and the collector area was established, and the calculating method of collector area based on the minimum annual cost was discussed. Taking a 100m^2 residential in Taiyuan as an example, the result of daily calculation shows that the annual cost of the system is minimum when the collector area is 21.34m^2. The solar fraction of the whole heating season is up to 60.2%.

  10. Auxiliary nRules of Quantum Mechanics

    OpenAIRE

    Mould, Richard A

    2005-01-01

    Standard quantum mechanics makes use of four auxiliary rules that allow the Schrodinger solutions to be related to laboratory experience, such as the Born rule that connects square modulus to probability. These rules (here called the sRules) lead to some unacceptable results. They do not allow the primary observer to be part of the system. They do not allow individual observations (as opposed to ensembles) to be part of the system. They make a fundamental distinction between microscopic and m...

  11. Decommissioning of the ASTRA research reactor: Dismantling the auxiliary systems and clearance and reuse of the buildings

    Directory of Open Access Journals (Sweden)

    Meyer Franz

    2008-01-01

    Full Text Available The paper presents work performed in the last phase of the decommissioning of the ASTRA research reactor at the Austrian Research Centers Seibersdorf. Dismantling the pump room installations and the ventilation system, as well as the clearance of the buildings is described. Some conclusions and summary data regarding the timetable, material management, and the cost of the entire project are also presented.

  12. Research of auxiliary energy storage system in hybrid electric vehicles%混合电动汽车辅助储能系统的研究

    Institute of Scientific and Technical Information of China (English)

    武琼; 滕青芳

    2013-01-01

    针对混合电动汽车在城市交通中频繁加速减速的特点,设计了基于超级电容储能的电动汽车辅助储能系统,选择两相交错式半桥拓扑双向DC/DC变换器作为超级电容的充放电电路.重点设计双向DC/DC变换器对超级电容的充放电控制,采用平均电流控制的两个电感电流内环和一个电压外环的控制策略,并对电动汽车辅助储能系统进行了Simulink仿真,从而有效验证了超级电容在电动汽车中应用的优势.%Hybrid electric vehicles have the characteristics of frequent acceleration and slowdown in urban traffic. A bi-directional DC-DC converter in the super-capacitor energy system was designed. A topology for bidirectional interleaved DC/DC converter was used in the super-capacitor energy system. The design of control loop of charge mode and discharge mode for the super-capacitor was emphasized, using two inner inductor current loop and one outer voltage loop of average-current-controlled. The auxiliary energy storage system for hybrid electric vehicles was simulated and the advantage of super-capacitor in hybrid electric vehicles could be effectively verified.

  13. Replacement of electrical protection of generation (main generator, main transformer and auxiliary transformers) and new associated monitoring system

    International Nuclear Information System (INIS)

    The replacement of the electrical protection of generation is a technological quantum leap, since moving from an analog system (known by) (all, intuitive and visual) to a digital (integrates the hardware on a single computer, much more powerful and programmable). The keys to overcoming the challenge are know to manage the technological leap, the operational limitations of plant (keep operating the)preferred sources of energy) and make a good design (including a review independent of the configuration of the relays, taking into account the experience (operational available).

  14. WMO Selected, Supplemenatary, Auxiliary Ships

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — World Meteorological Organization International List of Selected, Supplementary and Auxiliary Ships, recognized as Publication 47. 1973-1998 editions, gathered from...

  15. Auxiliary services offered to operator of transmissive system - the primary and secondary regulation of active capacity in ENO, o.z

    International Nuclear Information System (INIS)

    The change of outmoded analogue regulators of turbogenerators of Novaky Thermal Power Plant (TG ENO) B1 and B2 to new digital assigned the basic condition for fulfilment of strict rules for providing of auxiliary services - of primary and secondary regulation of active capacity and regulation of turns. The paper deals with technical and program sources for realisation of marked auxiliary services on the level of direct regulation of bloc, with fulfilment and evaluating of auxiliary services according to valid norm PNE 34-01/2002 in condensation and power take off operation of blocks after realisation of extraction of heat in hot water from ENO B. The paper argues, that regulation attributes of block - regulation zone of active capacity for primary and secondary regulation as like as the parameters of these regulations have been not changed by power take off operation. (author)

  16. Investigations in the modelling and control of a medium voltage hybrid inverter system that uses a low voltage /low power rated auxiliary current source inverter

    OpenAIRE

    Papadopoulos, Savvas; Rashed, Mohamed; Klumpner, Christian; Wheeler, Patrick

    2016-01-01

    Hybrid converters consist of a main inverter processing the bulk of the power with poor waveform performance and a fast and versatile auxiliary inverter to correct the distortion. In this paper, the main converter is a medium voltage NPC inverter and the auxiliary inverter is a low-voltage and low-current rated current source inverter (CSI), with series capacitor being used to minimize the CSI voltage stress. The result is a high output current quality which is obtained with a very low switch...

  17. Design of Lean Shipbuilding Auxiliary System%精益造船辅助系统的设计

    Institute of Scientific and Technical Information of China (English)

    汪德庆; 李宗骍; 赵李飞; 尚鑫; 王玉涛; 曹晶

    2013-01-01

    本文在总结精益制造的发展与应用基础上,通过企业的实地调研分析,结合管理和生产诸多方面的因素,针对当前企业内用于套料的材料信息管理,与优化套料,设计与开发了将零件信息、材料信息与套料板信息集成化管理的精益造船辅助系统。%This paper summarized the lean manufacturing and agile manufacturing’s development and applies. Through survey and research particular in this enterprise, combined management and production factors, author set up two ideas:ifrstly, integrated management in parts data, materials information and nesting information. Taking nesting work as core, complete all nesting work in this system, and it could search and statistically analyze nesting data.

  18. 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...

  19. 燃气热水器辅助太阳能热水系统的应用%The application of gas water heater auxiliary solar water heating system

    Institute of Scientific and Technical Information of China (English)

    桑松表

    2015-01-01

    This paper introduced the component of solar hot water system,made comparison and selection to boiler heating,electric heating,gas water heater heating three traditional auxiliary heat sources,pointed out that the gas water heater auxiliary heat source had the advantages of sav-ing the space,short system restart time,convenient to intermittent.%介绍了太阳能热水系统的组成要素,对锅炉加热、电加热、燃气热水器加热三种传统的辅助热源进行了比选,指出选用燃气热水器辅助热源具有节省空间,重启时间短,便于间断使用等优点。

  20. 新时期辅警制度的现实困局与探索路径%Dilemma and Exploration of the Auxiliary Police System in the New Era

    Institute of Scientific and Technical Information of China (English)

    刘显峰

    2015-01-01

    As a useful complement to the regular police force, the Auxiliary Police has played an important role in maintaining security and stability and become an indispensable auxiliary security force. However, in the process of system construction and actual operation of auxiliary police system, there appeared many problems, such as the diversity of auxiliary police titles, lack of security, awkwardness of identity positioning, etc. How to further strengthen the management and construction of the auxiliary police force and actively promote its sustainable and healthy development is the issue which public security organs cannot be avoided. It is believed that in order to vigorously promote the normalization and institutionalization of the auxiliary police, the management and construction should be standardized, the assessment should be lfexible, the skills should be specialized and the work should be transparent.%作为正规警力的有益补充,辅警在协助警察维护社会治安等方面发挥了重要作用,成为维护安全稳定不可或缺的治安辅助力量。不过,辅警体系的制度构建和现实运作也出现了辅警称谓多样化、成员复杂、缺乏保障、身份定位尴尬等一系列问题。加强辅警队伍的管理和建设,积极推动辅警队伍持续健康发展,是公安机关不容回避的现实问题。对于辅警队伍的管理与建设应从管理规范化、考核灵活化、技能专业化、工作透明化四个方面入手,大力推进辅警队伍的正规化、制度化建设。

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

    International Nuclear Information System (INIS)

    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%

  2. Auxiliary nRules of Quantum Mechanics

    CERN Document Server

    Mould, R A

    2005-01-01

    Standard quantum mechanics makes use of four auxiliary rules that allow the Schrodinger solutions to be related to laboratory experience, such as the Born rule that connects square modulus to probability. These rules (here called the sRules) lead to some unacceptable results. They do not allow the primary observer to be part of the system. They do not allow individual observations (as opposed to ensembles) to be part of the system. They make a fundamental distinction between microscopic and macroscopic things, and they are ambiguous in their description of secondary observers such as Schrodingers cat. The nRules are an alternative set of auxiliary rules that avoid the above difficulties. In this paper we look at a wide range of representative experiments showing that the nRules adequately relate the Schrodinger solutions to empirical experience. This suggests that the sRules should be abandoned in favor of the more satisfactory nRules, or a third auxiliary rule-set called the oRules. Keywords: brain states, c...

  3. Feedwater heater life optimization at Peach Bottom Atomic Power Station

    International Nuclear Information System (INIS)

    Many papers published over the last 15 years have strongly emphasized the need for an ongoing program of inspection and testing with subsequent failure cause analysis of feedwater heaters. With deregulation of the electric utility industry in various phases of implementation, utilities must decrease costs, both O ampersand M and capital, while optimizing plant efficiency. In order to accomplish this coal, utility engineers must monitor feedwater heater performance in order to recognize degradation, correct/eliminate failure mechanisms, and prevent in-service failures while optimizing availability. Periodic tube plugging without complete analysis of the degraded/failed area resolves the immediate need for return for service, however, heater life will not be graded/failed area resolves optimized. This paper illustrates a complete inspection, testing, and maintenance program implemented at PECO Energy's Peach Bottom Atomic Power Station (PBAPS). Concerns that tubes may have been too conservatively plugged due to insufficient data justified a program that included: (1) Removal of previously installed plugs. (2) Videoprobe inspection of failed areas. (3) Extraction of tube samples for further analysis. (4) Eddy current testing of selected tubes. (5) Evaluation of the condition of open-quotes insuranceclose quotes plugged tubes for return to service. (6) Hydrostatic testing of selected tubes. (7) 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: (1) The extent of actual degraded conditions, (2) The source(s) of failure mechanisms, (3) The details of repair. It is a combination of all gathered data that affords the best chance in arresting problems and optimizing feedwater heater life

  4. 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)

    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

  5. Optimization of feedwater pump configuration for the M310 nuclear power station

    International Nuclear Information System (INIS)

    After reviewing the selection of the feed water pump configuration for Ling'ao I Nuclear Power Station and its operation, and based on the study results for several feedwater pump configurations which have been used in domestic and foreign nuclear power stations, several typical feed water pump configurations are compared and analyzed technically and economically. Considering the economics and reliability, the authors think the configuration of 2 x 50% turbine driven feedwater pump plus 2 x 25% electrical driven feedwater pump is better than that of 2 x 75% turbine driven feedwater pump plus 2 x 50% electrical driven feedwater pump. If the capability of one motor of feedwater pump can be increased from 10 MW to 14-15 MW and the voltage can be increased from 6.6 kV to 11 kV, the capacity of one electrical driven feedwater pump can reach 75% of the total capacity, then the configuration of 3 x 50% electrical driven feedwater pump is also accepted. (authors)

  6. The chemistry of feedwater for boiling-water and pressurized-water reactors

    International Nuclear Information System (INIS)

    In a nuclear power plant the purity of the feedwater depends largely on whether a condensate polishing plant is provided, whether the loop is conditioned and on the presence of corrosion products originating in the materials from which the loop is made. The feedwater specification depends on the type of steam generator used. The article defines the characteristic parameters of a condensate polishing plant (CPP), such as the 'degree of polishing' and 'practical exchange capacity of the resins' and indicates how they can be determined. In pressurized-water reactors (PWR) the feedwater is normally conditioned with hydrazine. Measurements are quoted to demonstrate that, in contrast to conventional plants, the point of injection is immaterial as regards the copper content of the feedwater. Moreover, the iron content of the feedwater of a PWR can be reduced by using cyclic amines. The feedwater chemistry of a BWR is discussed by referring to oxygen, iron and copper measurements. The authors show that in loops in which the feed-heater condensate is pumped forwards and where a feedwater tank is provided, the stipulated purity of the feedwater can be attained by suitable measures (such as mechanical filtration, prevention of erosion-corrosion, and so on). (Auth.)

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

    International Nuclear Information System (INIS)

    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

  8. 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

  9. Improving Energy Efficiency of Auxiliaries

    International Nuclear Information System (INIS)

    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

  10. Abnormality transient analysis of Monju using a plant system code

    International Nuclear Information System (INIS)

    The objectives of the present study are to analyze plant transients caused by small abnormalities and to find plant parameters by which operators can recognize these small abnormalities. In order to evaluate the plant transient during an abnormal situation in the water system using the plant system code NETFLOW++, the turbine and feedwater systems should be analyzed with good precision. The code is validated using the measured data at Monju. Several abnormalities in the water system are candidates of the present study, e.g., feedwater control valve degradation, feedwater pump degradation, heat transfer degradation due to fouling on heat transfer tubes of the evaporator, loss-of-feedwater-heating, etc. All major components in the tertiary system are included in the calculation model such as the steam generators, the high-pressure turbine, the deaerator, the feedwater pump, the feedwater heaters, the feedwater control valves, the steam control valve, extraction lines and drainpipes. (author)

  11. Sectional replacement of high pressure feedwater heater tubing

    Energy Technology Data Exchange (ETDEWEB)

    Bolton, J.A.; Bowes, P.D. [TransAlta Utilities Corp., Duffield, Alberta (Canada). Plant Engineering Services

    1994-12-31

    TransAlta Utilities is a Canadian Corporation which owns and operates the coal fired Sundance Generating Station located in central Alberta. Sundance is fitted with vertical channel down, carbon steel tubed, high pressure feedwater heaters. The primary mode of failure of these HP feedwater heaters on the six generating units is steam inlet area tube erosion and vibration damage. This damage is initiated with the deterioration of the desuperheating inlet shroud and backing plate, primarily due to thermal fatigue, thus allowing direct impingement of high velocity steam and entrained condensate upon the tubing. Topics discussed are: review of the design and conditions of the heater which allowed re-conditioning; cutting, lifting and supporting of the shell at an elevation sufficient to allow free access of the entire desuperheating zone; damage observed within the desuperheating and drains cooler zones; bundle reconditioning through damage tube section replacement and support plate repair techniques; design/installation of the desuperheating, drains-cooling zone shrouds and backing plates; benefits that this type of approach may offer; conclusions.

  12. Hydraulic turbines and auxiliary equipment

    Energy Technology Data Exchange (ETDEWEB)

    Luo Gaorong [Organization of the United Nations, Beijing (China). International Centre of Small Hydroelectric Power Plants

    1995-07-01

    This document presents a general overview on hydraulic turbines and auxiliary equipment, emphasizing the turbine classification, in accordance with the different types of turbines, standard turbine series in China, turbine selection based on the basic data required for the preliminary design, general hill model curves, chart of turbine series and the arrangement of application for hydraulic turbines, hydraulic turbine testing, and speed regulating device.

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

    International Nuclear Information System (INIS)

    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 m2 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)

  14. Decontamination of the reactor pressure vessel and further internals and auxiliary systems in the German boiling water reactor Isar-1; Dekontamination des RDB inkl. der Einbauten wie Dampftrockner und Wasserabscheider sowie der angeschlossenen Hilfssysteme im deutschen Siedewasserreaktor ISAR 1

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Michael; Sempere Belda, Luis; Basu, Ashim; Topf, Christian [AREVA GmbH, Erlangen (Germany). Abt. Chemistry Services; Erbacher, Thomas; Hiermer, Thomas; Schnurr, Bernhard; Appeldorn, Thomas van [E.ON Kernkraft GmbH, Kernkraftwerk ISAR, Essenbach (Germany). Abt. Maschinentechnik; Volkmann, Christian [ESG Engineering Services GmbH, Greifswald (Germany)

    2015-12-15

    The German nuclear power plant ISAR 1 (KKI 1), a 878 MWe boiling water reactor of KWU design, was shut down on March 17{sup th}, 2011. With the objective to minimize the plants activity inventory accompanied by the reduction of contact dose rates of systems and components the project 'decontamination of the RPV incl. steam dryer and water separator and the connected auxiliary systems' was implemented in the first quarter of 2015. One major focus within the project was the specific in-situ decontamination of the steam dryer.

  15. Loss of a main feedwater pump test at 100% power simulation using Korean standard nuclear plant analyzer (KSNPA)

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Won Sang; Kim, Shin Whan; Sung, Kang Sik; Seo, Jong Tae; Lee, Sang Keun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    The Loss of a Main Feedwater Pump test at 100% Power for YGN 4 was simulated in order to verify and validate the KSNPA. The comparison of the test data with the KSNPA prediction results showed reasonable agreement in the trends of the major plant parameters. All plant control systems including NSSS and T/G control systems are properly actuated and stabilized the plant conditions to a new steady state conditions in the KSNPA. From the comparison results, the KSNPA showed its capability to simulate the LOMFP event for the KOrean Standard Nuclear Power Plant.

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

    International Nuclear Information System (INIS)

    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 (θ1). In this case, the effect of the Carnot temperature ratio (Φ1), absolute temperature ratio (ξ) and the phase-change absolute temperature ratio for the feedwater heater (Φ2) 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 (Φ2), lower the cost at the minima. As feedwater heater to hot end unit cost ratio decreases, the minimum total conductance required increases

  17. Efficiency improvement strategies for the feedwater heaters network designing in supercritical coal-fired power plants

    International Nuclear Information System (INIS)

    Coal will continue playing a major role in worldwide electricity generation during next years. This trend will augment CO2 emission to the atmosphere. Improving power plant efficiency could alleviate the negative effect of coal consumption on CO2 emissions. Main efforts have been focused on supercritical boiler technology (once-through units) and materials development (austenitic steels) with the aim of accomplishing high steam parameters that allow this efficiency enhancement. However, improvement in supercritical steam parameters should be followed by an exhaustive review of the steam cycle design. This paper shows a strategy for the optimization of the feedwater heaters network and the flue gas heat recovery system design. Starting with the lay-out of a supercritical steam cycle using the best available technology, this paper analyses not only the steam cycle itself, but also its integration with the boiler cold-end. By means of thermodynamic optimization it is possible to propose new feedwater heat exchanger network configurations and reducing steam consumption from turbine bleeds, achieving optimum power plant efficiency. Simulations have been carried out using Aspen Plus software and optimization procedure is based on a sequential quadratic programming method that maximized overall plant efficiency taking turbine bleeds pressure as independent variables. Results show a feasible improvement of the overall plant efficiency of 0.7 points in comparison with state-of-the-art reference plant. This increase implies a direct reduction of CO2 emissions of about 1.3% compared with the best plant currently available. Moreover, an economic analysis confirms the feasibility of the proposals analysed and shows important additional yearly incomes. - Highlights: • A strategy is presented for steam cycle design in supercritical units. • Thermodynamic and economic approaches are accomplished. • A case application study shows a power plant efficiency increase up to 0

  18. 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...

  19. 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...

  20. Ligand-Enabled β-C–H Arylation of Alpha-Amino Acids Using a Simple and Practical Auxiliary

    OpenAIRE

    Chen, Gang; Shigenari, Toshihiko; Jain, Pankaj; Zhang, Zhipeng; Jin, Zhong; He, Jian; Li, Suhua; Mapelli, Claudio; Miller, Michael M.; Poss, Michael A.; Scola, Paul M.; Yeung, Kap-Sun; Yu, Jin-Quan

    2015-01-01

    Pd-catalyzed β-C–H functionalizations of carboxylic acid derivatives using an auxiliary as a directing group have been extensively explored in the past decade. In comparison to the most widely used auxiliaries in asymmetric synthesis, the simplicity and practicality of the auxiliaries developed for C–H activation remains to be improved. We previously developed a simple N-methoxyamide auxiliary to direct β-C–H activation, albeit this system was not compatible with carboxylic acids containing α...

  1. Feedwater heater life optimization at Peach Bottom Atomic Power Station

    International Nuclear Information System (INIS)

    This paper illustrates a complete inspection, testing, and maintenance program implemented at PECO Energy's Peach Bottom Atomic Power Station (PBAPS). Concerns that tubes may have been too conservatively plugged due to insufficient data justified a program that included: removal of previously installed plugs; videoprobe 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 tubes; final repair plan based on the results of the above program. This paper concludes that no single method of inspection or testing should be solely relied upon in establishing: the extent of actual degraded conditions; the source(s) of failure mechanisms; and the details of repair. It is a combination of all gathered data that affords the best chance in arresting problems and optimizing feedwater heater life

  2. 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.).

  3. Unavailability Analysis of Digital Engineered Safety Feature Actuation System

    International Nuclear Information System (INIS)

    This paper quantitatively presents the results of the fault tree analysis of Digital Engineered Safety Feature Actuation System which is one of the most important signal generation systems in nuclear power plant because it generates the signal for mitigating possible accidents. In this paper, as an example, we explore the case of auxiliary feedwater actuation signal. Based on the analysis results, we quantitatively explain the relationship between the important characteristics of digital systems and the system unavailability. We find out some factors which remarkably affect the system unavailability. They are the common cause failures and the coverage of fault tolerant mechanisms. Human operator's backup also plays very important role. In this analysis we ignore the effect of software failure. We also compare the result with the PSA result of conventional analog Engineered Safety Feature Actuation System. The result of Digital ESFAS is about 27% lower than in the analog system

  4. 太阳能-蒸汽源与锅炉辅助供暖系统节能效益分析%Analysis on Energy Efficiency of Solar-Steam Sourceand Boiler Auxiliary Heating System

    Institute of Scientific and Technical Information of China (English)

    高越; 高丽娜; 孙清

    2015-01-01

    针对呼伦贝尔地区某工业区民用建筑热水供暖、电热水器为洗浴供热工程存在的缺点,提出太阳能-蒸汽源与锅炉辅助供暖系统,对改造后的工程进行实际测试,用DeST模拟探讨室内热环境和节能效益。结果表明,使用太阳能-蒸汽源与锅炉辅助供暖系统的室内环境得到明显改善,节能率可达到66.46%。%Aiming at the disadvantages existing in the project of civil construction hot water heating and bath heating by electric hotwater heater in a certain industrial area in Hulunbeir, it raised solar-steam source and boiler auxiliary heating system, and did practicaltest for the project after improving, used DeST to simulate to discuss indoor thermal environment and energy efficiency. The result showsthat the indoor environment is obviously improved after using solar-steam source and boiler auxiliary heating system, the energy savingrate can achieve 66.46%.

  5. Pre-test calculations of SPES experiment - a loss of main feedwater transient

    International Nuclear Information System (INIS)

    Results of a pre-test calculation of international standard experiment ISP-22 SPES are shown in this paper. SPES facility represents a model of three-loop PWR power plant which was used to perform an experimental loss of main feedwater transient with emergency feedwater delayed. calculation was performed by RELAP5/MOD2/36.1 computer code which we had converted to VAX computers. (author)

  6. 汽车辅助制动系统及试验方法研究%Study on Automotive Auxiliary Braking System and Test Method

    Institute of Scientific and Technical Information of China (English)

    王勇; 张雄; 刘翰东

    2013-01-01

    The authors introduce the type and principles of the auxiliary braking types, and describe the require-ments and test methods of typeⅡand typeⅡA tests. Through simulating the ramp test, they verify the meeting of the exhaust braking and the retarder braking to the standard, and put forward the relative recommendations.%介绍辅助制动装置的类型及原理,并对Ⅱ型和ⅡA型试验的要求和试验方法进行阐述。通过模拟坡道试验,验证排气制动和缓速器对标准的满足性,并给出相应建议。

  7. Hovercraft auxiliary power units (APUs)

    Energy Technology Data Exchange (ETDEWEB)

    Russell, B.J.

    1983-08-01

    Auxiliary power units (APU) manufactured by British firms for use in hovercraft are characterized. Both diesel and gas-turbine APUs are found to be well suited to the demands of this application. The design features, dimensions, performance data, and installation requirements are discussed for the SS 90, SS 923, DA-1, BA-1, HM 5, and Gevaudan 9 APUs, as well as the TRS 18 gas-turbine smoke generator. The progress made in improving the fuel efficiency of gas turbines and reducing the weight of diesel engines is considered significant.

  8. An assessment of RELAP5/MOD2 applicability to loss-of-feedwater transient analysis in a Babcock and Wilcox reactor plant

    International Nuclear Information System (INIS)

    The applicability and scaling capability of RELAP5/MOD2 when applied to a Babcock and Wilcox (B and W) loss-of-feedwater transient is assessed using a code applicability methodology. A loss-of-feedwater test with a feed-and-bleed recovery was selected from the once-through integral system (OTIS) test data as a reference transient. Nondimensional comparisons are made between code assessment calculations and code applications calculations using computer code models scaled according to scaling criteria derived from the work of Ishii and others. The results indicate that RELAP5/MOD2 can scale the phenomena observed in the experiment and that the code is applicable for transients for which phenomena are within this envelope. The results also demonstrate the usefulness of the code applicability methodology for interpreting and verifying code calculations. 21 refs., 59 figs., 12 tabs

  9. The limitations of a feed/water based heat-stable vaccine delivery system forNnewcastle disease-control strategies for backyard poultry flocks in sub-Saharan Africa.

    Science.gov (United States)

    Oakeley, R D

    2000-12-01

    Backyard poultry are a major contributor to egg and meat consumption in sub-Saharan Africa and an important source of income for many rural producers. Production throughout Africa is severely constrained by continuing outbreaks of Newcastle disease. The livestock-service sector lacks the resources and infrastructure to control Newcastle disease in extensive flocks without the active participation of producers. The development of 'heat-stable' Newcastle disease vaccines offers a potential solution. Trials over the last two decades have examined the effectiveness of heat-stable vaccines in both controlling Newcastle disease and in involving the rural community in control strategies. Constraints highlighted include the reliability of the vaccines using alternative delivery methods and the capacity of rural communities to apply those methods. The search for appropriate Newcastle disease-control strategies in extensive poultry systems should focus on policies and methodologies that incorporate the wider concerns and priorities of extensive producers. PMID:11087958

  10. The solar energy hot water system paired with auxiliary heat system of air source heat pump%以空气源热泵辅助加热的太阳能热水系统

    Institute of Scientific and Technical Information of China (English)

    施龙; 刘刚; 杨丰畅

    2013-01-01

    介绍了太阳能热泵热水系统的运行原理及控制方式,分析了太阳能热泵热水系统在上海某高校 学生宿舍的应用情况.以每吨热水的加热成本为指标,对几种热源热水系统的经济性进行了比较,分析了太阳能热泵热水系统的节能效果.%An operating principles and control strategies of solar energy hot water system with air source heat pump for auxiliary heating were introduced. Then an application in the students' dormitories of one university in Shanghai was analyzed. Basing on an index which was the cost of heating 1 ton water, the economic efficiency of several heat sources was compared, and higher energy-saving effect of current system was concluded.

  11. Auxiliary bearing design considerations for gas cooled reactors

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  13. Perfects, resultatives and auxiliaries in early English

    OpenAIRE

    McFadden, Thomas; Alexiadou, Artemis

    2008-01-01

    In this paper, we will argue for a novel analysis of the auxiliary alternation in Early English, its development and subsequent loss which has broader consequences for the way that auxiliary selection is looked at cross-linguistically. We will present evidence that the choice of auxiliaries accompanying past participles in Early English differed in several significant respects from that in the familiar modern European languages. Specifically, while the construction with have became a full-fle...

  14. Bispinor Auxiliary Fields in Duality-Invariant Electrodynamics Revisited

    OpenAIRE

    Ivanov, E. A.; Zupnik, B. M.

    2012-01-01

    Motivated by a recent progress in studying the duality-symmetric models of nonlinear electrodynamics, we revert to the auxiliary tensorial (bispinor) field formulation of the O(2) duality proposed by us in arXiv:hep-th/0110074, arXiv:hep-th/0303192. In this approach, the entire information about the given duality-symmetric system is encoded in the O(2) invariant interaction Lagrangian which is a function of the auxiliary fields V_{\\alpha\\beta}, \\bar V_{\\dot \\alpha\\dot \\beta}. We extend this s...

  15. Brane worlds in gravity with auxiliary fields

    International Nuclear Information System (INIS)

    Recently, Pani et al. explored a new theory of gravity by adding nondynamical fields, i.e., gravity with auxiliary fields (Phys Rev D 88:121502, 2013). In this gravity theory, higher-order derivatives of matter fields generically appear in the field equations. In this paper we extend this theory to any dimensions and discuss the thick braneworld model in five dimensions. Domain wall solutions are obtained numerically. The stability of the brane system under tensor perturbations is analyzed. We find that the system is stable under tensor perturbations and the gravity zero mode is localized on the brane. Therefore, the four-dimensional Newtonian potential can be realized on the brane. (orig.)

  16. Brane worlds in gravity with auxiliary fields

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Bin; Liu, Yu-Xiao; Yang, Ke [Lanzhou University, Institute of Theoretical Physics, Lanzhou (China)

    2015-02-01

    Recently, Pani et al. explored a new theory of gravity by adding nondynamical fields, i.e., gravity with auxiliary fields (Phys Rev D 88:121502, 2013). In this gravity theory, higher-order derivatives of matter fields generically appear in the field equations. In this paper we extend this theory to any dimensions and discuss the thick braneworld model in five dimensions. Domain wall solutions are obtained numerically. The stability of the brane system under tensor perturbations is analyzed. We find that the system is stable under tensor perturbations and the gravity zero mode is localized on the brane. Therefore, the four-dimensional Newtonian potential can be realized on the brane. (orig.)

  17. Economizer water-wall damages initiated by feedwater impurities

    Directory of Open Access Journals (Sweden)

    Vidojković Sonja M.

    2014-01-01

    Full Text Available The main causes of efficiency loss in thermal power plants are boiler tube failures that diminish unit reliability and availability, and raise the cost of the electric energy. For that reason, regular examination of boiler tubes is indispensable measure for prevention future malfunctions of power units. Microscopic examination of economizer inner wall microstructure, analysis of chemical composition of deposit using x-ray diffraction (XRD and scanning electron microscopy/energy dispersive spectroscopy (SEM/EDS has been performed in a subcritical power plant. Stress corrosion cracking, pitting corrosion, destroyed protective magnetite layer, presence of magnetite and hematite in deposit and corrosive impurities within the cracks were indicated the effect of inadequate quality of feedwater that can not entirely ensure reliable operation of the boiler. It may be stated that maintenance of present boiler does not provide its reliable operation. Extensive chemical control of water/steam cycle was recommended. [Projekat Ministartsva nauke Republike Srbije, br. III 43009 i br. III 45012

  18. Effects of auxiliary heat sources on energy efficiency of active solar heating systems%辅助热源对主动式太阳能供暖系统节能性的影响

    Institute of Scientific and Technical Information of China (English)

    于国清; 周继瑞

    2015-01-01

    根据太阳能贡献率和辅助加热设备的能源效率,将系统的能耗折算成一次能源,分析不同辅助热源对太阳能供暖系统节能性的影响,认为主动式太阳能供暖系统要达到节能效果必须满足一定的前提条件。研究结果表明:辅助热源方式和太阳能贡献率对系统的节能性影响很大;相对于燃气锅炉单独供暖,电加热辅助太阳能供暖系统只有在太阳能贡献率高于65.4%时才节能,而大多数系统的太阳能贡献率很难达到这么高,因此要尽量避免使用;热泵辅助太阳能供暖系统在大多数情况下都是节能的(热泵的平均 COP =2.5时,太阳能贡献率需高于7%);采用平均COP 高于3.0的热泵或采用燃气锅炉作为辅助热源时,系统都是节能的,太阳能贡献率越高,节能效果越明显。%According to the solar fraction and the energy efficiency of auxiliary heating equipments, analyses the effects of different auxiliary heat sources on the energy efficiency of active solar heating systems by converting energy consumption to primary energy,and believes that the active solar heating system must satisfy certain conditions for achieving energy saving effect.The results indicate that the auxiliary heat source and the solar fraction have a great impact on the energy efficiency of solar heating systems.Compared to the gas-fired boiler heating alone,the solar heating system assisted by electric heating is energy efficient only when the solar fraction is greater than 65.4%,which is very difficult to achieve for many real systems.The primary energy consumption of the solar heating system assisted by heat pump is lower in general instances (when the average COP of heat pump is 2.5,the solar fraction should be greater than 7%).When heat pumps with average COP above 3.0 or gas-fired boilers are used as auxiliary heat sources,the solar heating system is energy efficient,and the higher the solar

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

    International Nuclear Information System (INIS)

    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)

  20. Auxiliary space preconditioners for linear elasticity based on generalized finite element methods

    OpenAIRE

    Brannick, James; Cho, Durkbin

    2010-01-01

    We construct and analyze a preconditioner of the linear elastiity system discretized by conforming linear finite elements in the framework of the auxiliary space method. The auxiliary space preconditioner is based on discretization of a scalar elliptic equation with Generalized Finite Element Method.

  1. Auxiliary-field quantum Monte Carlo methods in nuclei

    CERN Document Server

    Alhassid, Y

    2016-01-01

    Auxiliary-field quantum Monte Carlo methods enable the calculation of thermal and ground state properties of correlated quantum many-body systems in model spaces that are many orders of magnitude larger than those that can be treated by conventional diagonalization methods. We review recent developments and applications of these methods in nuclei using the framework of the configuration-interaction shell model.

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

    International Nuclear Information System (INIS)

    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)

  3. An integral effect test on the feedwater line break of APR1400 with the ATLAS facility

    International Nuclear Information System (INIS)

    An integral effect test on the feedwater line break (FLB) was performed with the ATLAS facility for an APR1400 as a typical secondary system transient. The objectives of the present FLB tests are to understand the accident progression of the FLB scenario based on the APR1400 preliminary safety analysis report (PSAR) and to assess the prediction capability of system analysis codes such as the MARS, RELAP5, TRACE and SPACE. The main concern of the present FLB test is the peak RCS pressure and the major parameters affecting the peak RCS pressure are the break size, the break location, potential for reverse flow, initial pressurizer level and the initial SG level. The present test is performed for a break on the pipe connected to the economizer with a typical break size. The initial steady-state conditions and the sequence of event of FLB scenario for the APR1400 were successfully simulated with the ATLAS facility. In the present paper, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate are presented and discussed. It could be concluded that the APR1400 has the capability of coping with the hypothetical FLB scenario with an adequate set of controlling devices and proper setpoints. This integral effect test data could also be used to evaluate the prediction capability of existing safety analysis codes of MARS, RELAP5 and SPACE and to identify any code deficiency for an FLB simulation. (author)

  4. An Improved Auxiliary Particle Filter Algorithm for Information Fusion in Multi-Agent System%改进的辅助粒子多Agent系统信息融合滤波算法

    Institute of Scientific and Technical Information of China (English)

    朱波; 陈贞翔

    2012-01-01

    In order to overcome the nonlinear problem and improve the collaborative capabilities in the multi-agent system, this paper proposes an improved auxiliary particle information fusion filter algorithm. The architecture of multi-agent system is analyzed, then the signal of agent is decomposed and reconstructed based on orthogonal wavelet multi-scale analysis theory, so the improved auxiliary particle information fusion filter algorithm is proposed based on the signal of agent, which can be taken as the important sampling density function,after that,the features are integrated into multi-agent system's features. Finally,this algorithm is applied to the maneuvering target tracking through computer simulation and compared with conventional algorithms,and simulation results verify its effectiveness.%为克服多Agent系统的非线性问题,提高多Agent系统对信息的协同合作能力,提出一种改进的辅助粒子信息融合滤波算法.该算法通过分析多Agent系统的结构,根据正交小波多尺度分析理论对系统内Agent的信息进行分解、重构,给出以Agent信息为重要采样密度函数的辅助粒子滤波算法,并对滤波结果进行特征融合,得到多Agent系统融合特征.将此算法应用到机动目标跟踪领域,并与传统的滤波算法进行对比,仿真结果验证了该算法的有效性.

  5. Orthodontic springs and auxiliary appliances: assessment of magnetic field interactions associated with 1.5 T and 3 T magnetic resonance systems

    Energy Technology Data Exchange (ETDEWEB)

    Kemper, J.; Priest, A.N.; Adam, G. [University Medical Center of Hamburg-Eppendorf, Clinic of Diagnostic and Interventional Radiology, Hamburg (Germany); Schulze, D. [University Hospital of Freiburg, Department of Oral and Maxillofacial Surgery, Freiburg (Germany); Kahl-Nieke, B.; Klocke, A. [University Medical Center of Hamburg-Eppendorf, Department of Orthodontics, Hamburg (Germany)

    2007-02-15

    The objective of this paper is to evaluate magnetic field interactions at 1.5 and 3 T for 20 orthodontic devices used for fixed orthodontic therapy. Twenty springs and auxiliary parts made from varying ferromagnetic alloys were tested for magnetic field interactions in the static magnetic field at 1.5 and 3 T. Magnetic translational force F{sub z} (in millinewtons) was evaluated by determining the deflection angle {beta} [American Society for Testing and Materials (ASTM standard test method)]. Magnetic-field-induced rotational force F{sub rot} was qualitatively determined using a five-point scale. {beta} was found to be >45 in 13(15) devices at 1.5(3) T and translational force F{sub z} exceeded gravitational force F{sub g} on the particular object [F{sub z} 10.17-261.4 mN (10.72-566.4 mN) at 1.5(3) T]. F{sub z} was found to be up to 24.1(47.5)-fold higher than F{sub g} at 1.5(3) T. Corresponding to this, F{sub rot} on the objects was shown to be high at both field strengths ({>=} +3). Three objects (at 1.5 T) and one object (at 3 T) showed deflection angles <45 , but F{sub rot} was found to be {>=} +3 at both field strengths. For the remaining objects, {beta} was below 45 and torque measurements ranged from 0 to +2. Of 20 objects investigated for magnetic field interactions at 1.5(3) T, 13(15) were unsafe in magnetic resonance (MR), based on the ASTM criteria of F{sub z}. The implications of these results for orthodontic patients undergoing MRI are discussed. (orig.)

  6. Orthodontic springs and auxiliary appliances: assessment of magnetic field interactions associated with 1.5 T and 3 T magnetic resonance systems

    International Nuclear Information System (INIS)

    The objective of this paper is to evaluate magnetic field interactions at 1.5 and 3 T for 20 orthodontic devices used for fixed orthodontic therapy. Twenty springs and auxiliary parts made from varying ferromagnetic alloys were tested for magnetic field interactions in the static magnetic field at 1.5 and 3 T. Magnetic translational force Fz (in millinewtons) was evaluated by determining the deflection angle β [American Society for Testing and Materials (ASTM standard test method)]. Magnetic-field-induced rotational force Frot was qualitatively determined using a five-point scale. β was found to be >45 in 13(15) devices at 1.5(3) T and translational force Fz exceeded gravitational force Fg on the particular object [Fz 10.17-261.4 mN (10.72-566.4 mN) at 1.5(3) T]. Fz was found to be up to 24.1(47.5)-fold higher than Fg at 1.5(3) T. Corresponding to this, Frot on the objects was shown to be high at both field strengths (≥ +3). Three objects (at 1.5 T) and one object (at 3 T) showed deflection angles rot was found to be ≥ +3 at both field strengths. For the remaining objects, β was below 45 and torque measurements ranged from 0 to +2. Of 20 objects investigated for magnetic field interactions at 1.5(3) T, 13(15) were unsafe in magnetic resonance (MR), based on the ASTM criteria of Fz. The implications of these results for orthodontic patients undergoing MRI are discussed. (orig.)

  7. 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...

  8. Retran kinetics studies for loss of feedwater anticipated transients without scram

    International Nuclear Information System (INIS)

    Results are presented for postulated loss of feedwater (LOFW) anticipated transient without scram (ATWS) analyses performed with the RETRAN-02 MOD003 computer code for a four-loop Westinghouse pressurized water reactor in order to demonstrate the capabilities of the RETRAN code to predict system transient responses similar to the Westinghouse LOFTRAN code and the Brookhaven National Laboratory RELAP code results. Differences between the RETRAN, LOFTRAN and RELAP code predictions can be attributed to differences in methodology and modelling techniques. These LOFW ATWS studies provide additional insight into the kind of detailed parametric analyses required for using RETRAN for ATWS studies or for the evaluation of any reactor transient dominated by moderator density feedback phenomena. This paper gives the results of point kinetics sensitivity studies and steam generator relief valve modelling effects. The effect of using a realistic steam generator model with multiple banked relief valve setpoints versus an ideal single plateau value is presented. The modelling of actual multiple banked opening and closing pressure setpoints result in higher primary system pressure due to the fast steam generator inventory depletion

  9. The centralized monitoring system solution for auxiliary equipment and environmental facilities in the station operation%厂站运行辅助设备及环境设备的集中监测系统解决方案

    Institute of Scientific and Technical Information of China (English)

    张振华

    2013-01-01

    With the gradual expansion of power system construction, unattended substation is imperative. The number and type of auxiliary equipment and environmental equipment in station operation are relatively large, so to realize the efficient and centralized monitoring and management of the equipment bears the brunt of difficulties. A new solution of plant station running auxiliary equipment and environmental equipment for centralized monitoring system is designed and proposed based on the relative switch information of automatic control and new sensing technology collect equipment. The system realizes centralized monitoring and management on auxiliary equipment and environment in the station operation by the dispatching terminal, after the communication network of electric power and the internal information network sent back the running state data of auxiliary equipment and environmental equipment in the station operation, which not only greatly improves the quality and efficiency of the equipment operation and maintenance, but also provides a new solution for substation communication network operation and maintenance. It can effectively realize unmanned operation and development of substation.%随着电力系统建设规模日趋壮大,变电站无人化已势在必行。厂站运行辅助设备、环境设备的数量和类型均相对较多,实现设备的高效、集中监控管理便成为了首当其冲的难题。基于自动控制和新型传感技术采集设备的相关开关量信息,设计并提出了一种新型的厂站运行辅助设备及环境设备的集中监测系统解决方案。该方案通过电力通信网和内部信息网络完成厂站运行辅助设备及环境设备运行状态数据的回传,从而实现了地调主站端对厂站运行的辅助设备和环境的集中实时监控和管理,极大地提高了设备运行维护的质量和效率。该研究为今后变电站通信网络的运行和维护提供了一种新的解决

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

    International Nuclear Information System (INIS)

    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

  11. Unavailability modeling and analysis of redundant safety systems

    International Nuclear Information System (INIS)

    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

  12. Loss-of-feedwater, steam generator tube rupture, and steam line break experiments: Steam generator transient response test program: Interim report

    International Nuclear Information System (INIS)

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results. Two LOF tests were analyzed in detail. Both tests were initiated from 100% power condition by shutting off the main feedwater flow. In LOF Test No. 1, the remaining boundary conditions were kept constant while in LOF Test No. 2, the power was rapidly reduced to 3%. The results show that the primary to secondary heat transfer becomes degraded when the collapsed water liquid level in the bundle region falls below approximately 50 inches. The SGTR test analyzed in detail - SGTR Test No. 2 - simulated the post-reactor-trip portion of the SGTR transient (T/sub prim/ = 5600F). The transient was initiated by starting the SGTR flow injection and simultaneously shutting off the auxiliary feedwater. The water level rose and flooded the dryer to its mid-elevation by the end of the test. The primary carry-over was shown to be less than 0.4% of the tracer mass injected into the secondary side by the SGTR flow. SGTR Test No. 3 investigated the response of the intact steam generator. Reverse heat transfer and low heat flow conditions were simulated. The results have demonstrated the occurrence of temperature stratification in the secondary water which lasted for about 800 seconds

  13. 多方式辅助加热太阳能集中供热水远程监测系统%Remote Multi-way Auxiliary Heating Monitoring System for Solar Central Heating

    Institute of Scientific and Technical Information of China (English)

    田志宏; 任立鹏; 董智

    2016-01-01

    在对太阳能供热水系统特点及空气源热泵工作原理进行研究的基础上,设计了以空气源热泵和电加热器为辅助热源的太阳能集中供热水远程监测系统,不仅解决了太阳能供热水系统单一热源的局限性问题,同时也实现了远程实时监测。以天津某高校培训中心采用的太阳能供热系统为例,结合当地气候,计算了工程需要的集热器面积和空气源热泵机组数量,详细介绍了采用模块化设计方法开发远程监测系统的过程,并与原天然气供热水系统的水、电和天然气实测消耗量进行了对比。结果表明,多方式辅助加热太阳能集中供热水系统的综合运行费用明显降低,具有较高的应用价值。%Based on the research of the characteristics of solar water heating systems and the working principle of air-source heating pump,a remote monitoring system for solar heating was designed,using air source heating pump and electric heater as the auxiliary heat source.It can not only solve the problem of the limitation of the solar heating system due to single heat-ing source,but also realize the remote and real-time monitoring of the solar water heating system.Taking the solar energy project of a university training center in Tianjin as an example,the required collector area and air source heating pump units were calculated against the local climate.The process of developing a remote monitoring system with modular designs was described in detail,and the actual consumption of water,electricity and natural gas were measured and compared with the natural gas heating system.The results show that the total operation cost of the multi-way auxiliary heating system is obvi-ously reduced,and the system is of higher value in application.

  14. Feedwater transient and small break loss of coolant accident analyses for the Bellefonte Nuclear Plant

    International Nuclear Information System (INIS)

    Specific sequences that may lead to core damage were analyzed for the Bellefonte nuclear plant as part of the US Nuclear Regulatory Commission's Severe Accident Sequence Analysis Program. The RELAP5, SCDAP, and SCDAP/RELAP5 computer codes were used in the analyses. The two main initiating events investigated were a loss of all feedwater to the steam generators and a small cold leg break loss of coolant accident. The transients of primary interest within these categories were the TMLB' and S2D sequences. Variations on systems availability were also investigated. Possible operator actions that could prevent or delay core damage were identified, and two were investigated for a small break transient. All of the transients were analyzed until either core damage began or long-term decay heat removal was established. The analyses showed that for the sequences considered the injection flow from one high-pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were able to prevent core damage in the S2D sequence; no operator actions were available to prevent core damage in the TMLB' sequence

  15. Establishment of Scenarios for a Hybrid SIT operation with an Active System

    International Nuclear Information System (INIS)

    After the Fukushima accident, passive nuclear safety was strengthened to prevent the recurrence of a severe accident. A hybrid safety injection tank (H-SIT) is suggested as one of the passive systems used to prevent a severe accident. Therefore, the development of an operation strategy for a H-SIT with an active system is required for increasing the level of safety. In this study, a master logic diagram, difference analysis, matching process, and accident case classification are performed to determine the proper parameters. Then, the conditions that require the use of a H-SIT are determined using a decision process. The operation strategy analysis indicates that a H-SIT can mitigate five failures, namely, the safety injection pump (SIP), passive auxiliary feedwater system (PAFS), depressurization system, shutdown cooling pump (SCP) and recirculation system failures and this strategy also indicates that each scenario has its own pressure range in which the H-SIT can be used

  16. Establishment of Scenarios for a Hybrid SIT operation with an Active System

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, In Seop; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2014-08-15

    After the Fukushima accident, passive nuclear safety was strengthened to prevent the recurrence of a severe accident. A hybrid safety injection tank (H-SIT) is suggested as one of the passive systems used to prevent a severe accident. Therefore, the development of an operation strategy for a H-SIT with an active system is required for increasing the level of safety. In this study, a master logic diagram, difference analysis, matching process, and accident case classification are performed to determine the proper parameters. Then, the conditions that require the use of a H-SIT are determined using a decision process. The operation strategy analysis indicates that a H-SIT can mitigate five failures, namely, the safety injection pump (SIP), passive auxiliary feedwater system (PAFS), depressurization system, shutdown cooling pump (SCP) and recirculation system failures and this strategy also indicates that each scenario has its own pressure range in which the H-SIT can be used.

  17. Modeling and simulation on feed-water heater of nuclear power plant

    International Nuclear Information System (INIS)

    The feed-water heater is one of the major equipment in the secondary loop of nuclear power plant (NPP), and its behavior has an important influence on the safe and economical operation of NPP. The research on the behavior of feed-water heater by means of modeling and simulation can provide important theoretical basis for its design and operation. In this paper, the distributed parameter dynamic models of NPP feed water heater were established, in which the nearly separated model was used to deal with two-phase flow. By simulating the behavior of actual NPP feed-water heaters under various operating conditions and comparing the differences between the simulation values and the actual values, the accuracy of the simulation models was proven to be higher than that of existing models. (authors)

  18. Global stratification effects in feedwater line to steam generator in PWR power plant

    International Nuclear Information System (INIS)

    Recent measurements recorded at Beaver Valley Power Station-Unit 1, a PWR plant, demonstrate the presence of significant global thermal stratification within horizontal sections of the feedwater line. Global thermal stratification has not been documented nor its effects quantified on pressurized water reactor feedwater piping. An investigation of the feedwater line at BV-1 was performed to determine the effects of global stratification. Plant specific operating data were gathered for a recent plant restart using temporary instrumentation installed to measure piping temperatures and displacements. Data reduction identified significant global thermal stratification and significant pipe movement during several operating modes. Steps were taken to incorporate its effects into the design calculations when the global thermal stratification was determined to be an unanalyzed condition which was not enveloped by the existing design basis calculations

  19. Control system for a nuclear power producing unit

    International Nuclear Information System (INIS)

    The invention deals with an improvement of a power/load control system for a PWR type reactor. A signal that regulates the feedwater supply is in proportion to the desired power of the reactor. Arrangements for the control of the heat transfer from the reactor and the control of the feedwater flow rate to the steam generator are both simultaneously affected by the feedwater supply signal. This coordinated control system embodies a fast response of power generation to power demand

  20. 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.).

  1. Conceptual Design of Passive Containment Cooling System Based on APR+

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Byongguk; No, Heecheon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-05-15

    Passive Auxiliary Feedwater System (PAFS) is expected to work well under extended SBO, but is vulnerable to extended SBO coupled with loss of coolant accident (LOCA). Various reactors have been developed, such as AP1000, ESBWR, and KERENA, with passive containment cooling systems (PCCSs) dealing with the accident scenario of SBO with LOCA. The performance of the PCCSs is already or almost validated. Though PCCS is well adopted into BWRs, there has been no success in PWRs with concrete containment. In this paper, we suggest a new PCCS based on APR+ and represent scoping analysis results. The Fukushima accident proved the importance of treating extended SBO. To deal with extended SBO with LOCA scenario, the PCCS based on APR+ is suggested and evaluated roughly for the first time as a PWR with concrete containment.

  2. An Investigation of Safety Improvements using Additional Safety Class Instrument Air System in Nuclear Power Plant

    International Nuclear Information System (INIS)

    An additional safety class instrument air (IA) system delivers air to safety class air operated valves (AOVs) such as pressurizer PORVs, the steam generator PORVs, and auxiliary feedwater modulation valves related to natural circulation, when a loss of offsite power (LOOP) occurs. This paper analyzes the risk change before and after installing the new IA system based on the probabilistic safety assessment (PSA) methodology. This approach identifies the cause of the risk reduction through a detailed cutset investigation and measures the importance of the added components and human error. Three main conclusions were elicited: there was a 64.6%∼99.1% reduction in the unavailability of the supporting system related to the safety AOVs, a 19.1% CDF reduction due to the new IA system, and the importance measures of human error were decreased

  3. An Investigation of Safety Improvements using Additional Safety Class Instrument Air System in Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yeon Kyoung; Chi, Moon Goo [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of); Kim, Hak Seon [Korea Hydro and Nuclear Power Co., Busan (Korea, Republic of)

    2011-08-15

    An additional safety class instrument air (IA) system delivers air to safety class air operated valves (AOVs) such as pressurizer PORVs, the steam generator PORVs, and auxiliary feedwater modulation valves related to natural circulation, when a loss of offsite power (LOOP) occurs. This paper analyzes the risk change before and after installing the new IA system based on the probabilistic safety assessment (PSA) methodology. This approach identifies the cause of the risk reduction through a detailed cutset investigation and measures the importance of the added components and human error. Three main conclusions were elicited: there was a 64.6%{approx}99.1% reduction in the unavailability of the supporting system related to the safety AOVs, a 19.1% CDF reduction due to the new IA system, and the importance measures of human error were decreased.

  4. Diagnosis of Feedwater Heater Performance Degradation using Fuzzy Approach

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  5. Participation of Empresarios Agrupados in engineering of the Tokamak systems and auxiliary buildings; Participacion de Empresarios Agrupados en la Ingenieria de los sistemas de tokamak y edificios auxiliares

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez del Palacio, V.

    2013-07-01

    The Architect Engineering works comprise approximately the design of 32 buildings and structures, some of them very simple and others extremely complex. 32 buildings include nuclear buildings, as the Toyama, tritium, or the building of emergency diesel and buildings conventional and the building of the site services. Systems including those related to systems within a building they reach conventional, while interfaces with process systems are countless. Grouped entrepreneurs led the design of the mechanical (PBS65) systems and Electric (PBS43) actively participating in the design of systems of fire proofing of buildings and site. During the development of engineering accident of Fukushima works directly affect in the design of the systems.

  6. Calculation of the limiting CESSAR Feedwater Line-Break and Steam Line-Break transients

    International Nuclear Information System (INIS)

    Argonne National Laboratory (ANL), under contract to the Nuclear Regulatory Commission, performed audit calculations of the limiting Feedwater Line Break (FLB) and Steam Line Break (SLB) transients presented in the CESSAR FSAR. The results of the FLB and SLB calculations are discussed

  7. Fission product transport in the reactor coolant system for a spectrum of interfacing system LOCA scenarios

    International Nuclear Information System (INIS)

    One of the most important potential severe accident sequences for any pressurized water reactor (PWR) is a loss of coolant accident (LOCA), or V-sequence, in one of the interfacing systems. As initially described in the reactor safety study WASH-1400, interfacing system LOCAs involved the failure of check valves in emergency core cooling systems (ECCS), but could also involve the residual heat removal (RHR) systems. The check valves protect the low-pressure portions of these systems from the high pressures of the reactor coolant system (RCS) to which they are connected to provide cold leg injection. A consequent break in the low-pressure piping outside the containment may result in core damage and a direct pathway for fission products to be transported from the core, through the RCS and ECCS or RHR to the auxiliary building, from which they can escape to the environment. This paper addresses the retention and transport of fission products (specifically, CsI) in the RCS in V-sequence scenarios. It summarizes some of the major differences between models resulting from the latest version of the industry degraded core rulemaking (IDCOR) MAAP Computer Program, MAAP 3.0B. Discussed are the differences in: fission product transport and retention in small, medium, and large ECCS pipe breaks, as well as the effect of ECCS and auxiliary feedwater (AFW) system operation and fission product retention in the various regions of the RCS as calculated by MAAP 3.0B and the STCP

  8. Contributions to the research programs in nuclear and industrial electronics, domestic production of instrumentation, safety and control systems and equipment for nuclear reactors and auxiliary installations

    International Nuclear Information System (INIS)

    Domestic production of component system and equipment for the control and safety of nuclear facilities was one of the priority objective of the Nuclear Research Institute Pitesti. The problems addressed were particularly related to design and production of analog and digital equipment for measurements, triggering and display of the values of process parameters as well as to regulating complex functions of this equipment. Associated to this effort were the research works concerning: - reliability and in-service life-time of the electronic components and equipment in the safety and control systems for nuclear processes; - radiation endurance of industrial electronic components; utilization of whirling currents in calandria tube testing; - expert systems and applications in nuclear reactor control and safety; design and testing methods of process real time software packages for safety in control critical systems for nuclear domain. There are presented characteristics of the following equipment: 1. amplifier for ionization chambers with triggering comparator circuits for the CANDU 600 reactor shut down system; 2. amplifier for ionization chambers without triggering comparator circuits for power regulating system; 3. safety and regulating computerized system for C9 and C5 cans; 4. acquisition system for dosimetric data in nuclear facilities; 5. program able digital comparator for the reactor shut down system; 6. stationary gamma areal monitors for CANDU 600 reactors and other nuclear facilities

  9. A simplified model for scheduling services on auxiliary bus lines

    OpenAIRE

    Codina Sancho, Esteve; Montero Mercadé, Lídia

    2014-01-01

    In this paper, a mathematical programming based model is described to assist with the schedule of services for a set of auxiliary bus lines that operate alleviating a disruption of the regular transportation system during a given time period. In contrast to other models, considered static, service schedules are set taking into account demand fluctuations that may happen in that time period. Passenger flows are represented with a multi-commodity structure and disseminate throug...

  10. Analysis of total loss of feedwater event for the determination of safety depressurization bleed capacity

    International Nuclear Information System (INIS)

    The Ulchin 3 and 4, which are 2825 MWt PWRs, adopted Safety Depressurization System (SDS) to mitigate the beyond design basis event of Total Loss of Feedwater (TLOFW). In this study the results and methodology of the analyses for the determination of SDS bleed capacity are discussed. The SDS design bleed capacity has been determined from the CEFLASH-4AS/REM simulation according to the following design criteria: 1) Each SDS flow path, in conjunction with one of two High Pressure Safety Injection (HPSI) pumps, is designed to have a sufficient capacity to prevent core uncovery if one SDS path is opened simultaneously with the opening of the Pressurizer Safety Valves (PSVs). 2) Both SDS bleed paths are designed to have sufficient total capacity with both HPSI pumps operating to prevent core uncovery if the Feed and Bleed (F and B) initiation is delayed up to thirty minutes from the time of the PSVs lift. To verify the results of CEFLASH-4AS/REM simulation a comparative analysis has also been performed by more sophisticated computer code, RELAP5/MOD3. The TLOFW event without operator recovery and TLOFW event with F and B are analyzed. The predictions by the CEFLASH-4AS/REM of the transient two phase system behavior are in good qualitative and quantitative agreement with those by the RELAP5/MOD3 simulation. Both of the results of analyses by CEFLASH-4AS/REM and RELAP5/MOD3 have demonstrated that decay heat removal and core inventory make-up can be successfully accomplished by F and B operation during TLOFW event for the Ulchin 3 and 4. 18 figs., 2 tabs., 14 refs. (Author)

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

    OpenAIRE

    竹田 武司

    2016-01-01

    The TR-LF-07 test simulated a loss-of-feedwater transient in a PWR. A 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 ...

  12. Feedwater piping guillotine breaks at 340 deg. C operation temperature

    International Nuclear Information System (INIS)

    The feed water pipe in power plant KARDIA 1, operated at p = 184 bar and T = 340 deg. C ruptured in five cross sections (guillotine breaks) after 130 000 operating hours. The burst feed pipe of 250 inside diameter and 18 mm wall thickness made of 15 NiCuMoNb 5 (according to specification 1972) was subjected to 1) mechanical-technological testing 2) stress analysis and 3) fractographic analysis, as a basis for final fracture mechanic assessment. The analysis results revealed that the state of the material found after the failure of the pipe is characterized by low ductility. In addition, numerous material intergranular micro-separations which occurred under operating conditions were found within individual fractured cross sections. The circumferential fractures parallel to the circumferential welding seam regions started in precracked regions and widened in ductile manner (shearing) with low ductility at fractured areas. The guillotine break in parent material started in an area on circumference which impacted the boiler auxiliary structure. The rest of this fracture is ductile by shear. The remaining circumferential partial cross section fractures found in the base material started from impacted zones and exhibit predominantly brittle fractures. The analysis of the findings reveals that fractures started in circumferential stress relief cracks adjoined to geometrical imperfections in highly stressed pipe cross sections. In the crack extension areas localities with intergranular cracks have been allocated. After the cracks reached the critical length and depth under sustained bending moment loading, the spontaneous rupture followed. (author)

  13. Orbiter Auxiliary Power Unit Flight Support Plan

    Science.gov (United States)

    Guirl, Robert; Munroe, James; Scott, Walter

    1990-01-01

    This paper discussed the development of an integrated Orbiter Auxiliary Power Unit (APU) and Improved APU (IAPU) Flight Suuport Plan. The plan identifies hardware requirements for continued support of flight activities for the Space Shuttle Orbiter fleet. Each Orbiter vehicle has three APUs that provide power to the hydraulic system for flight control surface actuation, engine gimbaling, landing gear deployment, braking, and steering. The APUs contain hardware that has been found over the course of development and flight history to have operating time and on-vehicle exposure time limits. These APUs will be replaced by IAPUs with enhanced operating lives on a vehicle-by-vehicle basis during scheduled Orbiter modification periods. This Flight Support Plan is used by program management, engineering, logistics, contracts, and procurement groups to establish optimum use of available hardware and replacement quantities and delivery requirements for APUs until vehicle modifications and incorporation of IAPUs. Changes to the flight manifest and program delays are evaluated relative to their impact on hardware availability.

  14. Systems reliability evaluation under aging by the device of stages optimized through genetic algorithms

    International Nuclear Information System (INIS)

    A methodology is developed based on the method of stages optimized through genetic algorithms to allow reliability evaluation of plant stand by systems. This methodology allows the transformation of on-Markovian models into equivalent Markovian ones by considering a set of fictitious states, called stages. Therefore, a state with a time-dependent transition rate is represented by a combination of stages (series, parallel or series/parallel) with constant transition rate. The optimization is necessary to overlap the difficulties presented to find the identification parameter of the device of stages, which is a complex optimization problem and requires a robust and efficient optimization tool, e.g the genetic algorithms. results concerning initial applications to Auxiliary Feedwater Systems (AFW) pumps are shown and commented, for which Weibull and lognormal distribution are employed respectively for modeling failures and repair times. (author)

  15. Comparison of single phase self excited induction generator excitation configurations using the main and auxiliary windings for a micro hydro system

    Energy Technology Data Exchange (ETDEWEB)

    Oh, T.; Freere, P. [Kathmandu Univ. (Nepal). Dept. of Electrical and Electronic Engineering

    2005-07-01

    Induction generators are one of the most popular generators for micro hydro systems. Three phase generators are often used even though only a single phase supply is needed. This paper investigates the characteristics of single phase induction generators. Many different winding and excitation capacitor configurations are investigated. To use these generators effectively, the desired output should be near the rated power and voltage. This paper includes laboratory test and simulation results. (Author)

  16. 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.

  17. Heat transfer equipment performance diagnosis of auxiliary systems in electric power stations; Diagnostico de comportamiento de equipo de transferencia de calor de sistemas auxiliares de centrales termoelectricas

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz Esparza Gutierrez, Rogelio [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1990-12-31

    In this article the methodology followed to diagnose the performance of the equipment where heat is transferred from the feed water, condensate and circulation water systems in fossil power plants (FPP). The data collection is made with the unit in normal operation, using local instrumentation without taking the equipment out of service for its installation. The equipment diagnosis is made through the analysis of the collected data in actual operation and the design data; for this purpose a thermal balance of the interested systems is performed to obtain all the conditions an operation data. Later on the performance indicative parameters (PIP) of actual operation and design are calculated and compared one against the other. Such a comparison reveals the performance deterioration and the possible equipment faults. The data obtained and the supplementary information are stored in a data base whose objective is that Comision Federal de Electricidad has on hand a prompt access to them in order to control the performance, compare them among similar units and power stations, and inclusively verify possible recurrent causes of low availability in the referred systems. [Espanol] En este articulo se presenta la metodologia seguida para diagnosticar el comportamiento de equipos en los que se transfiere calor de los sistemas de agua de alimentacion, condensado y circulacion de las centrales termoelectricas (CTE). La toma de datos se realiza con la unidad en operacion normal, utilizando instrumentacion local sin necesidad de sacar de servicio a los equipos para su instalacion, ya que se ocupan los mismos puntos para instrumentos con que cuentan por diseno. El diagnostico de los equipos se realiza mediante el analisis de los datos recopilados, tanto de operacion real como de diseno; para ello, se efectua un balance termico de los sistemas de interes para obtener todas las condiciones y los datos de operacion. Posteriormente, se calculan los parametros indicativos de

  18. Transport in Auxiliary Heated NSTX Discharges

    International Nuclear Information System (INIS)

    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

  19. 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.

  20. New energy conversion processes and neglected auxiliary systems - a treasure trove for energy service providers; Neue Techniken und vernachlaessigte Nebenanlagen - eine Fundgrube fuer Energiedienstleister

    Energy Technology Data Exchange (ETDEWEB)

    Jochem, E.

    1998-12-31

    The paper shows that apart from the risks, the deregulation of the energy sector also offered new chances, which recently materialized in three emerging independent trends, initiating innovative approaches for energy conversion and applications and demand for new knowledge-based, near-production energy services as a new field for engineers: Providers of energy efficiency-related technology widen their capabilities by adding the new business segment of energy service marketing. Industrial energy users more strongly consider outsourcing of energy conversion systems and the relevant contracting schemes as an option for reducing energy costs. Public enterprise as energy users started following this trend, all the more as insufficient financial resources do not permit them to reinvest. The paper discusses the conversion processes which currently open up promising chances for energy service producers. (orig./CB) [Deutsch] Neben Gefahren, die inhaerentes Merkmal der Liberalisierung sind, zeichnen sich drei unabhaengige Entwicklungen ab, die innovative Initiativen in der Energiewandlung und -nutzung staerken und wissensintensive produktionsnahe Dienstleistungen als zentrale Aufgabe fuer den Ingenieur initieren: - Die Energieeffizienz-Technologieanbieter nehmen die Energiedienstleistungen in ihre Angebotspalette auf. - Die betrieblichen Energieanwender beginnen zunehmend, ueber das Outsourcing energiewandelnder Anlagen Energiekosten zu senken. - Die oeffentliche Hand als Energieanwender beginnt aehnlich zu denken, zumal sie infolge der hohen Verschuldung unzureichende finanzielle Mittel fuer Reinvestitionen hat. Hierueber sei kurz berichtet, bevor auf die Techniken eingegangen wird, die sich besonders aus heutiger Sicht fuer Energiedienstleister eignen. (orig./RHM)

  1. 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)

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

    International Nuclear Information System (INIS)

    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

  3. Actual flow calibration of a feedwater flowmeter using a high Reynolds number facility at NMIJ

    International Nuclear Information System (INIS)

    The results of calibration tests of the feedwater flowrate of ultrasonic flowmeters used in a nuclear power plant for variety of upstream conditions obtained using the new high Reynolds number calibration facility at NMIJ are described. In this examination, the measurements are performed for five pattern pipe layouts with one or two elbows. The flow conditioners installed upstream of the flowmeter are the tube bundle type and the Mitsubishi, which are normally used in nuclear power plants. The calibration result for each flowmeter are largely different for each flow conditioner and each upstream pipe layout, except in some special cases. Moreover, the trend of the correction factor with Reynolds number is not uniform for each case. Furthermore, some differences were observed for individual flowmeters. It is recommended that the feedwater flowmeter, especially when used to perform measurement uncertainty recapture, is calibrated based on the actual pipe layout and the Reynolds number corresponding to the actual nuclear power plant conditions.

  4. Actual flow calibration of a feedwater flowmeter using a high Reynolds number facility at NMIJ

    Energy Technology Data Exchange (ETDEWEB)

    Furuichi, Noriyuki [Fluid Flow Division, National Metrology Institute of Japan, AIST, Tsukuba Central 3, 1-1-1 Umezono, Tsukuba, 305-8563 (Japan)], E-mail: furuichi.noriyuki@aist.go.jp; Terao, Yoshiya [Fluid Flow Division, National Metrology Institute of Japan, AIST, Tsukuba Central 3, 1-1-1 Umezono, Tsukuba, 305-8563 (Japan)], E-mail: yterao@ni.aist.go.jp; Takamoto, Masaki [Fluid Flow Division, National Metrology Institute of Japan, AIST, Tsukuba Central 3, 1-1-1 Umezono, Tsukuba, 305-8563 (Japan)], E-mail: m.takamoto@nifty.ne.jp

    2009-07-15

    The results of calibration tests of the feedwater flowrate of ultrasonic flowmeters used in a nuclear power plant for variety of upstream conditions obtained using the new high Reynolds number calibration facility at NMIJ are described. In this examination, the measurements are performed for five pattern pipe layouts with one or two elbows. The flow conditioners installed upstream of the flowmeter are the tube bundle type and the Mitsubishi, which are normally used in nuclear power plants. The calibration result for each flowmeter are largely different for each flow conditioner and each upstream pipe layout, except in some special cases. Moreover, the trend of the correction factor with Reynolds number is not uniform for each case. Furthermore, some differences were observed for individual flowmeters. It is recommended that the feedwater flowmeter, especially when used to perform measurement uncertainty recapture, is calibrated based on the actual pipe layout and the Reynolds number corresponding to the actual nuclear power plant conditions.

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

    International Nuclear Information System (INIS)

    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. Reactor feedwater flow rate control device in a BWR type power plant

    International Nuclear Information System (INIS)

    Purpose: To control reactor feedwater level stationarily in a case where steams are released from relief valves. Constitution: Flow rate of steams discharged from relief valves is determined by the reactor pressure and the number of opened relief valves, and the value is added to the main steam flow rate detected by a steam flow rate detector. Then, based on the sum and the feedwater flow rate, a flow rate deviation signal is obtained. Thus, in a case where steams are discharged from relief valves, the flow rate of steams discharged from the reactor can be estimated accurately with no negative errors, and reduction in the reactor water level can thereby be prevented. (Kamimura, M.)

  7. Improvement on the EDF monitoring system for the evaluation of thermal stratification

    International Nuclear Information System (INIS)

    The SYSFAC system has been developed earlier in the nineties by EDF. This led to a software computer program permitting to affect operating detected transient to a design transient, function not only of functional parameter but also of mechanical criteria. Another use, called 'fatigue-meter', is now to get an evaluation of the Fatigue damage evolution during NPP's life, using measured actual loads for the computation of local stress and strain and finally the usage factor online. EDF decided to use as less as possible transducer data which are source of incertitude and invalidity. Therefore the aim of the instrumentation described here is to get correlations between data easily and confidently measured flowrate, temperature...) and the temperature vertical distribution in piping submitted to thermal stratification. The instrumentation of a feedwater line of steam generator has been carried out for a CPY (900 MWe) NPP and more recently for a 1300 MWe NPP with a new instrumentation design improving significantly the precision of results. A ring of several thermocouples is installed around the pipe and an ultrasonic flowmeter is used to measure low flowrate (until 1 to 2 m3/h) of the auxiliary feedwater line, range of flow creating the more detrimental stratification states. The correlations obtained are presented and more generally the acquisitions are analyzed to improve the knowledge on thermal loads created by thermal stratification. (authors)

  8. Determination of intersecting curve between two surfaces of revolution with parallel axes by use of auxiliary planes and auxiliary spheres

    Directory of Open Access Journals (Sweden)

    Obradović Ratko

    2002-01-01

    Full Text Available In this paper the space intersecting curve between two surfaces of revolution with parallel axes of surfaces have been determined. Two mathematical models for determination of intersecting curve between two surfaces of revolution have been formed: auxiliary planes have been used in the first mathematical model and auxiliary spheres have been used in the second model (Obradović 2000. In the first case each auxiliary plane intersected with each surface of revolution on circle and two points of intersecting curve are obtained as intersecting points between these two circles. In the second case centres of two locks of auxiliary spheres are put on axes of surfaces of revolution (centre of first lock is on axis of the first surface of revolution and centre of second lock is on axis of the second surface of revolution on saine z coordinate (when axes of surfaces of revolution are parallel with z axis of coordinate system. First lock sphere intersects the first surface of revolution on w1 parallels and second lock corresponding sphere intersects the second surface of revolution on w2 circles. It is possible to find a relationship that for selected radius of the first lock sphere can determine the radius of second lock sphere and real points of intersecting curve have been determined by use of these two spheres. The points of intersecting curve between two surfaces of revolution are obtained by intersection between w1 circles from the first surface with w2 circles from the second surface (Obradović 2000.

  9. Assessment of the code RELAP5/MOD2 against loss of feedwater without scram

    International Nuclear Information System (INIS)

    The integral effect test L9-3 (loss of feedwater without reactor trip) performed at the LOFT facility was analyzed as part of an assessment of the RELAP5/MOD2 code with the aim of qualifying this simulation tool for analysis of pressurization transients in pressurized water reactors. The code proved suitable for analysis of this kind of transients. Some conclusions of relevance to simulation of anticipated transients without scram scenarios with forced circulation could be drawn. (orig.)

  10. Assessment of the code RELAP5/MOD2 against loss of feedwater without scram

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. (Union Fenosa, Madrid (Spain))

    1993-02-01

    The integral effect test L9-3 (loss of feedwater without reactor trip) performed at the LOFT facility was analyzed as part of an assessment of the RELAP5/MOD2 code with the aim of qualifying this simulation tool for analysis of pressurization transients in pressurized water reactors. The code proved suitable for analysis of this kind of transients. Some conclusions of relevance to simulation of anticipated transients without scram scenarios with forced circulation could be drawn. (orig.).

  11. 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.

  12. Vacuum switchgear for power station auxiliary switchboards

    International Nuclear Information System (INIS)

    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)

  13. Curricular Guidelines for Dental Auxiliary Radiology.

    Science.gov (United States)

    Journal of Dental Education, 1981

    1981-01-01

    AADS curricular guidelines suggest objectives for these areas of dental auxiliary radiology: physical principles of X-radiation in dentistry, related radiobiological concepts, principles of radiologic health, radiographic technique, x-ray films and intensifying screens, factors contributing to film quality, darkroom, and normal variations in…

  14. 基于模糊规则提升理论的马病辅助诊断专家系统%Equine diseases auxiliary diagnosis expert system based on fuzzy rule promotion theory

    Institute of Scientific and Technical Information of China (English)

    秦宏宇; 李建新; 高翔; 王欢; 肖建华; 王洪斌

    2016-01-01

    approach for providing the intelligence in the system for diagnosis of the equine diseases. This experiment was conducted to develop a remote auxiliary equine diseases diagnosis expert system. By collecting and analyzing the experiences of diagnosis and treatment from experts on equine disease, the numerical expression of the equine diseases diagnosis knowledge was developed. The knowledge of equine diseases was represented with the method called object-attribute-value triples act (referred as O-A-V act) that combined with the generative formula. As such, it was easy to extract knowledge rules and these rules were used for inference mechanism. Using the confidence factor, multi-valued logic was used to represent the rules of confidence level. In this paper, we suggested a new inference method which was based on use of a fuzzy rule promotion theory. This approach can enhance the intelligence of the disease diagnosis system. If a rule was repeatedly used in corrective diagnostic results, it was then promoted to a higher confidence factor by the rule promotion factor (PCF), and the PCF was the original confidence factor in the next diagnosis session. In short, the dynamic PCF which was generated in the past dialogue was used instead of static CF in the final decision making process. The dynamically promoted rules were derived from those diagnosis sessions, which resulted in successful decisions. This enabled more efficient decision making in the future sessions. With this approach, it was not only decreasing the number of interactive between the system and the users, but also leading to acceptable diagnostic results. Based on the research of knowledge representation and inference mechanism, an auxiliary diagnostic expert system of equine diseases based on Microsoft.Net and SQL Server 2008 was designed and developed. It provided online help to equine farmers and extension workers in China. For the inference engine of system, we used the fuzzy rule promotion methodology that

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

    International Nuclear Information System (INIS)

    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)

  16. Evaluation of two-phase flow and heat transfer in feed-water heater

    International Nuclear Information System (INIS)

    In this study, a mass transfer model for a noncondensable gas and a condensation heat transfer model were added to a three-dimensional two-phase flow analysis program, in order to evaluate heat transfer performance in a feed-water heater. The heat transfer model was verified by condensation heat transfer experiments with 4 X 32 rows of heat transfer tubes and thermal resistance due to noncondensable gas at a condensation surface was measured. The calculated overall heat transfer coefficients agreed with the data within +-5% under the inlet quality of 13-100%. Feed-water heater calculations indicated that the effect of noncondensable gas on the heat transfer was negligibly small because of its low concentration. Based on the calculated results, improved feed-water heaters were proposed in which tube support plates had holes to lead steam into neighboring sections and the shell had a two-phase flow inlet on its side to separate droplets from the steam flow

  17. Study of a loss of feedwater transient for an integral pressurized water reactor with a helical coil steam generator

    International Nuclear Information System (INIS)

    The Multi-Application Small Light Water Reactor (MASLWR) is a system-level test facility constructed by Oregon State University to examine the thermal hydraulic phenomena that are of importance to integral pressurized water reactors (IPWRs). These phenomena include natural circulation instability, helical coil steam generator (HCSG) heat transfer and coupled reactor-containment pressurization. In MASLWR, the steam generator, the pressurizer and the reactor coolant system are integrated in one reactor pressure vessel unit and a metal slab is used to conduct the heat transfer between the containment vessel and the cooling pool. MASLWR is a scaled model of the NuScale design, a small modular reactor (SMR), with 1:3 length scale, 1:254 volume scale, 1:1 time scale, and full operating pressure and temperature. An IAEA International Collaborative Standard Problem (ICSP) was conducted using MASLWR to generate experimental data for various natural circulation conditions. In the ICSP, two experiments were conducted – “Loss of feedwater transient with subsequent automatic depressurization system operation” and “Normal operating conditions at different power levels”. US Nuclear Regulatory Commission participated in this standard problem in order to validate the TRACE MASLWR model and to identify any potential challenges to the code in simulating this unique design. The aim of the study is to support the future licensing review of the NuScale-like integral reactor design. In this paper, the loss of feedwater transient simulation will be discussed. The modeling strategy for the IPWR's unique components, such as the helical coil steam generator and the coupled reactor and containment, will be described in details. TRACE results show qualitative agreement with the experimental data in the major parameters of interest. However, one important event time of the calculation - the pressure equalization between the reactor and the containment - deviates from the data

  18. Auxiliary field formulation of supersymmetric nonlinear sigma models

    International Nuclear Information System (INIS)

    Two dimensional N=2 supersymmetric nonlinear sigma models on hermitian symmetric spaces are formulated in terms of the auxiliary superfields. If we eliminate auxiliary vector and chiral superfields, they give D- and F- term constraints to define the target manifolds. The integration over auxiliary vector superfields, which can be performed exactly, is equivalent to the elimination of the auxiliary fields by the use of the classical equations of motion. (author)

  19. Effects of auxiliary source connections in multichip power module

    DEFF Research Database (Denmark)

    Li, Helong; Munk-Nielsen, Stig; Beczkowski, Szymon;

    2016-01-01

    Auxiliary source bond wires and connections are widely used to in the power module with paralleled MOSFETs or IGBTs. This paper investigates the working mechanism and the effects of the auxiliary source connections in multichip power modules. It reveals that the auxiliary source connections cannot...

  20. Software design and implementation of GPS-based gun directing radar auxiliary guide system%基于GPS的炮瞄雷达辅助引导系统软件设计与实现

    Institute of Scientific and Technical Information of China (English)

    黄信安; 刘刚

    2014-01-01

    空空导弹研制过程需要进行一系列地面飞行试验,而炮瞄雷达为高精度跟踪雷达,波束很窄,不具备大范围自动搜索功能;在地面试验中,雷达跟踪过程出会现丢失目标无法引导的情况,构不成试验条件,导致试验无法进行;该软件利用靶机提供的GPS定位数据,实时解算目标方位、俯仰信息传输给炮瞄雷达系统,实现了对目标的辅助引导、对靶机飞行轨迹的记录回放、以及对目标方位修正功能;实际应用结果表明,该软件具有较高的可靠性和可扩展性,完全满足辅助引导系统的需要。%During the development process of air-to-air missile,a serial of flight test have to be done. The gun directing radar has high precision in tracking,bu its beam is narrow,so it does not have wide range of automatic search function. In ra-dar tracking procedure,the situation of target missing can stop the test. The software can calculate the target bearing and pitch information in real-time,then transmit it to gun directing radar system by using GPS locator data from target drone,thus to realize the functions of assisted guidance for the target,recording and playback of target drone flight trajectory,and the modification the target bearing. The practical application result shows that the software has high reliability and scalability,and can fully meet the needs of auxiliary and guide system.

  1. Evaluation of the applicability of amines as feed-water control reagents in terms of equilibrium water chemistry

    International Nuclear Information System (INIS)

    All Volatile Treatment (AVT), using ammonia and hydrazine, has been employed in Japan to control the pH of PWR secondary water. However, this treatment may become less effective at high temperatures. In recent years, new methods for controlling pH values have been studied overseas. One such method in high-pH control, which employs high-concentration ammonia solution. Other methods involve treatment with organic chemicals such as ethanolamine (ETA) and dimethylamine (DMA). It is important to evaluate the pH control effects of these new treatments. The purpose of this study was to evaluate the effect of these treatments on pH control and on mitigation of iron and copper oxides transportation into a PWR secondary water system. The main results are as follows: 1. High-pH control treatment can decrease the transportation of iron oxide in both feed-water and drainage-water systems. However, it is necessary to exclude copper alloys and to devise methods for decreasing the condensate demineralizer load. 2. ETA treatment appears to be more effective in mitigating the transportation of iron and copper oxides than the conventional AVT treatment. ETA treatment is the most effective method for pH control, particularly in drainage-water system. (author)

  2. Modified Darboux transformations with foreign auxiliary equations

    International Nuclear Information System (INIS)

    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. Auxiliary facilities on nuclear ship 'MUTSU'

    International Nuclear Information System (INIS)

    The nuclear ship 'MUTSU' has been moored at SEKINEHAMA, MUTU City in AOMORI Prefecture and several tests and works are being carried out on the ship. The construction of the auxiliary facilities for these works on the ship was completed in safety in August 1988. After that the facilities have fulfilled their function. The outlines of design, fabrication and construction of the facilities are described in this paper. (author)

  4. New Massive Supergravity and Auxiliary Fields

    CERN Document Server

    Bergshoeff, Eric A; Parra, Lorena; Rosseel, Jan; Yin, Yihao; Zojer, Thomas

    2013-01-01

    We construct a supersymmetric formulation of linearized New Massive Gravity without introducing higher derivatives. Instead, we introduce supersymmetrically a set of bosonic and fermionic auxiliary fields which, upon elimination by their equations of motion, introduce fourth-order derivative terms for the metric and third-order derivative terms for the gravitino. Our construction requires an off-shell formulation of the three-dimensional supersymmetric massive Fierz--Pauli theory. We discuss the non-linear extension of our results.

  5. 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...

  6. Automatic control device for feedwater flow rate into reactor

    International Nuclear Information System (INIS)

    In automatic control for a water injection flow rate, an anticipated transient without screw (ATWS) signal is outputted upon judgement of the occurrence of ATWS event based on a reactor power signal and a scram demand signal, and a high pressure water injection system inactivation signal is outputted upon detection for the inactivation of a high pressure water injection system. An ATWS/high pressure water injection system inactivation judging section outputs a high pressure water injection system inactivation signal. A reactor pressure capable of water injection and a pressure change signal for setting opening/closing of a main steam relief valve corresponding thereto are calculated to output the same to a pressure control section for setting opening/closing of the main steam relief valve. Even if insertion of the entire control rods should fail upon scram by the loss of reactor water to disable the scram, and high pressure water injection system is not operated, the reactor pressure and the water level of the reactor are automatically controlled, and water is injected from a low pressure water injection system with no trouble, to suppress the reactor power. Then, the integrity of the reactor pressure vessel and the reactor container can be maintained. (N.H.)

  7. The dynamics of a turbine-driven reactor feedwater pump and identification of related parameters in the Kuosheng power station

    International Nuclear Information System (INIS)

    In the Kuosheng nuclear power station, a turbine-driven reactor feedwater pump (TDRFP) is used to drive feedwater to the reactor pressure vessel (RPV). The performance of the TDRFP plays an important role in the safe operation of the nuclear power station as it governs the water inventory in the RPV. The dynamics of the TDRFP in the Kuosheng nuclear power station are simulated, and related parameters are identified by the simplex search method. A reactor feedwater pump (RFP) trip test was performed during the startup test of the Kuosheng nuclear power station. This RFP trip transient was selected for this study because it simplifies the identification process. The ratio of rated torque to moments of inertia, parameters related to the characteristic H-Q curve, and parameters related to the characteristic T-Q curve are verified

  8. Solids filtration of high-temperature feedwater in a PWR secondary circuit: Final report

    International Nuclear Information System (INIS)

    Pressurized water reactor steam generators and turbines have experienced a variety of corrosion problems as a result of ionic, corrosion product and oxidizing species transport to the steam generators. Installation of high temperature filters on final feedwater, high pressure drains and moisture separator drains to reduce corrosion product ingress to the steam generators of a 1160 MWe design basis plant are specified and evaluated. Cost estimates for installing electromagnetic filters, and added operating and maintenance costs are given. 18 refs., 12 figs., 9 tabs

  9. 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.

  10. 49 CFR 230.57 - Injectors and feedwater pumps.

    Science.gov (United States)

    2010-10-01

    ... delivery systems required. Each steam locomotive must be equipped with at least two means of delivering water to the boiler, at least one of which is a live steam injector. (b) Maintenance and testing... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION STEAM LOCOMOTIVE INSPECTION AND MAINTENANCE STANDARDS Boilers...

  11. Key issues of oxygenated feed-water treatment for supercritical boiler%超临界锅炉给水加氧的关键问题

    Institute of Scientific and Technical Information of China (English)

    艾志虎; 刘定平

    2011-01-01

    The oxygenated feed-water treatment technology of supercritical boiler could effectively decrease the internal corrosion and leposition rate from the water condensation system to the boiler water feeding system.The steam-water quality and the operating efficiency of supercritical units are greatly promoted .Combined with the different feed-water operation modes in the supercritical power currently,the features of the operational ways and the fundamental principles have been described.The key issues in oxygenated treatment(OT)for feed-water of supercritical units were researched.Based on an accurate and reliable adding oxygen system,the precise oxygen feeding amount and the strict transistion conditions between AVT and OT,the effective implementation of the oxygenated treament has been proposed so as to provide a useful reference to popularized the technique of oxygen addition for the supercritical electricity generation units.%超临界锅炉给水加氧处理技术可以有效降低机组凝结水至炉水系统管内的腐蚀及沉积速率,提高汽水品质和机组运行效率.结合当前超临界锅炉不同的给水运行方式,阐述了给水加氧特点和基本原理;针对当前困扰超临界锅炉给水加氧技术应用的关键问题进行研究,提出了有效实施给水加氧技术必须选择正确可靠的加氧系统、把握精确的氧质量分数控制和严格控制全挥发性处理与给水加氧处理方式相互转换条件等关键环节,可为给水加氧技术在大型超临界火力发电机组的推广应用提供指导.

  12. Outdoor testing of solar water heaters - Effects of load pattern and auxiliary boosting

    Energy Technology Data Exchange (ETDEWEB)

    Morrison, G.L.; Gilliaert, D.; Tebaldi, P. (Commission of the European Communities, Ispra (Italy))

    1992-10-01

    The effect of load draw-off pattern and auxiliary boosting on the performance of vertical and horizontal tank solar water heaters is investigated. The selection of an appropriate load pattern to use in a performance rating test is shown to depend on the type of system. The primary factor governing the dependence of performance on load pattern in single tank systems is conduction between the auxiliary boosted zone and the solar preheat zone. A method of quantifying the effect of conduction in solar tanks is presented.

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

    International Nuclear Information System (INIS)

    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

  14. Analysis of the Pactel loss-of-feedwater experiments

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V.; Virtanen, E.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Kouhia, J. [VTT Energy, Espoo (Finland). Nuclear Energy

    1995-12-31

    The western thermal-hydraulic system codes, like RELAP5 and CATHARE2, are developed and validated to model the phenomena in the vertical steam generators. The differences between vertical and horizontal steam generators are so significant that the calculational models developed for the vertical steam generators are not directly applicable for the horizontal steam generators. The PACTEL loss of feed water experiments have been performed to study the overall behaviour of the horizontal steam generators and especially to produce experimental data for code assessment. In this paper the results from APROS, CATHARE2 and RELAP5 calculations of a series of experiments are presented. 6 refs.

  15. Solid Oxide Fuel Cell Auxiliary Power Unit

    International Nuclear Information System (INIS)

    Solid Oxide Fuel Cell (SOFC) is an attractive, efficient, clean source of power for transportation, military, and stationary applications. Delphi has pioneered its application as an auxiliary Power Unit (APU) for transportation. Delphi is also interested in marketing this technology for stationary applications. Its key advantages are high efficiency and compatibility with gasoline, natural gas and diesel fuel. It's consistent with mechanizations that support the trend to low emissions. Delphi is committed to working with customers and partners to bring this novel technology to market

  16. [Hospital auxiliary staff, between polyvalence and invisibility].

    Science.gov (United States)

    Veissier, Pascale

    2016-01-01

    Often underestimated, hospital auxiliary staff carry out on a daily basis a professional activity that may be difficult to define and/or recognize. What does their work consist in and what are the boundaries of the scope of their activity? Faced with a growing rate of absenteeism among these members of staff in a nursing home for elderly people attached to a hospital, an issue emerges: does the content of their professional activity have an impact on the causes and evolution of this phenomenon? PMID:26976318

  17. Evaluation of erosion corrosion damage of the original WWER 440 steam generator feedwater piping

    International Nuclear Information System (INIS)

    The effects of erosion corrosion on WWER 440 steam generator feedwater piping were evaluated by using the CHECMATE code (EPRI), involving data of temperature of the medium, Cr, Mo, and Cu contents of the steel, medium flow, component geometry, liquid and gaseous fractions of the medium, pH, and oxygen content of the medium. The assessment was performed for the Dukovany NPP. It was concluded that in view of the relatively low corrosion damage rate, additional development of corrosion defects will not be caused by erosion. The development of defects observed on the T-piece may be due to a combined effect of erosion and corrosion in points where additional corrosion takes place in defects that had developed before. The degradation mechanism causing a further development of existing defects seems to consist in the corrosion effect of the aqueous medium in operating conditions. The additional development of defects of the feedwater piping is a serious finding requiring adequate provisions. 4 tabs., 7 figs., 2 refs

  18. National Ignition Facility subsystem design requirements laser auxiliary subsystem SSDR 1.3.5

    International Nuclear Information System (INIS)

    This system design requirement document establishes the performance, design, development and test requirements for the NIF Laser Auxiliary Systems. The Laser Auxiliary Systems consist of: a. Gas Cooling System; b. Low conductivity cooling water system; C. Deionized cooling water system; d. Electrical power distribution system. The gas cooling system will be used for cooling the main laser amplifier flashlamps and some smaller quantities will be used for purging Pockels cells and for diode pumps in preamplifier. The low conductivity cooling water system will be used for cooling the capacitor banks. The deionized cooling water system will be used to cool the multi-pass amplifier in the OPG PAM. Electrical power will be required for the OPG systems, Pockels cells, power conditioning, and amplifier support equipment

  19. Analysis of the Loss of Feedwater scenario for RBMK-1500 by the coupled code Athlet-Quabox/Cubbox taking into account late reactor trip signals and ATWS conditions

    International Nuclear Information System (INIS)

    For the accident analysis of RBMK-1500 reactors the GRS system code ATHLET and the 3-dimensional-reactor core model QUABOX/CUBBOX have been adapted to the special requirements of this type of reactor and a coupled version of both codes for RBMK-1500 investigations has been developed. The coupled model was applied to analyse the event Loss of Feedwater by best estimate methods to investigate the consequences of failures of reactor trip initiations. Therefore, late trip signals where credited to determine the available time for a safe shutdown, relevant for the demands on the newly developed diverse shut down system for Ignalina NPP Unit 2. These results are also compared to cases assuming ATWS conditions. The results show that a reactor trip at t equals 143 s from the begin of the transient, neglecting two preceding trip initiations, is sufficient for a safe shut down. If the shutdown function fails (ATWS), the transient behaviour is fully determined by the reactivity feedback, which is mainly dominated by the change of fuel temperature. The loss of feedwater provides worse cooling conditions and the fuel temperatures are rising significantly causing a power reduction to about 10 %. This power reduction is not sufficient to avoid that cladding temperature limits of 1200 C degrees and pressure tube temperature limits of 650 C degrees are exceeded. The transient provides an example to demonstrate the advantage of the coupled code system in comparison to point kinetic models since local effects influencing the reactivity behaviour are taken into account and maximum loaded channels are described in an adequate manner. (authors)

  20. Dynamic Response during the Online Compensation of Misalignment Based on Auxiliary Electromagnetic Support

    Institute of Scientific and Technical Information of China (English)

    LI Hui-min; LI Yan; HE Yong

    2008-01-01

    A new support structure,named auxiliary electromagnetic support,is introduced to adjust the position of rotor shaft online.The auxiliary electromagnetic support consists of sliding bearing,iron core,radial spring,pedestal and electroraagnet that could adjust the rotor position.In a rotor system adopting two auxiliary electromagnetic supports,the online compensation of misalignment was achieved without stopping work.Using the proposed auxiliary electromagnetic support,the analytical solution of dynamic response was obtained by analyzing a simulated single-degree-of-freedom(SDOF)system.And the dynamic response of the multi-degrees-of-freedom(MDOF)system in the experiment is acquired using Newmark variable step-size integration method.Further the new design is tested in a symmetrical single-disk rotor-bearing system.With both the theory analysis and the experimental results,some conclusions are obtained about the relationship between the vibration and the system damping which are related to system stability.

  1. Auxiliary pattern for cell-based OPC

    Science.gov (United States)

    Kahng, Andrew B.; Park, Chul-Hong

    2006-10-01

    The runtime of model-based optical proximity correction (OPC) tools has grown unacceptably with each successive technology generation, and has emerged as one of the major bottlenecks for turnaround time (TAT) of IC data preparation and manufacturing. The cell-based OPC approach improves runtime by performing OPC once per cell definition as opposed to once per cell instantiation in the layout. However, cell-based OPC does not comprehend inter-cell optical interactions that affect feature printability in a layout context. In this work, we propose auxiliary pattern-enabled cell-based OPC which can minimize the CD differences between cell-based OPC and model-based OPC. To enable effective insertion of auxiliary pattern (AP) in the design, we also propose a post-placement optimization of a standard cell block with respect to detailed placement. By dynamic programming-based placement perturbation, we achieve 100% AP applicability in designs with placement utilizations of cell-based OPC with AP can match gate edge placement error (EPE) count of model-based OPC within 4%. This is an improvement of 90%, on average, over cell-based OPC without APs. The AP-based OPC approach can reduce OPC runtimes versus model-based OPC by up to 40X in our benchmark designs. We can also achieve reduction of GDSII file size and ORC runtimes due to hierarchy maintenance of cell-based OPC.

  2. The auxiliary elliptic-like equation and the exp-function method

    Indian Academy of Sciences (India)

    Hong Li; Jin-Liang Zhang

    2009-06-01

    The auxiliary equation method is very useful for finding the exact solutions of the nonlinear evolution equations. In this paper, a new idea of finding the exact solutions of the nonlinear evolution equations is introduced. The idea is that the exact solutions of the auxiliary elliptic-like equation are derived using exp-function method, and then the exact solutions of the nonlinear evolution equations are derived with the aid of auxiliary elliptic-like equation. As examples, the RKL models, the high-order nonlinear Schrödinger equation, the Hamilton amplitude equation, the generalized Hirota–Satsuma coupled KdV system and the generalized ZK–BBM equation are investigated and the exact solutions are presented using this method.

  3. An auxiliary equation technique and exact solutions for a nonlinear Klein-Gordon equation

    International Nuclear Information System (INIS)

    A mathematical technique based on an auxiliary equation and the symbolic computation system Matlab is developed to construct the exact travelling wave solutions to a nonlinear Klein-Gordon equation. Some new solutions including the Jacobi elliptic function solutions, the degenerated soliton-like solutions and the triangle function solutions to the equation are obtained.

  4. 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)

    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

  5. A theoretical investigation of the effect of steam generator on the passive residual heat removal system performance

    International Nuclear Information System (INIS)

    The Passive Residual Heat Removal System (PRHRS) of the AP600 is designed to remove the residual heat through heat exchangers located in the In-containment Refueling Water Storage Tank (IRWST) when Steam Genarators (S/Gs) lose their heat removal capabilities. In domestic area, PRHRS performance has been analyzed using the simple computer program and RELAP code and a heat transfer test through the heat exchanger tube has been planned with support from the CARR. In this study, the heat removal capability of the PRHRS has been evaluated by simulating the realistic system responses in conjunction with the NSSS and the secondary system. In addition, the effect of steam generators on the PRHRS performance has been investigated. For this study, the system transient analysis code has been developed based on the existing computer code which is capable of analyzing the transient related to the feedwater system and natural circulation in the primary system. The developed computer code simulates system responses for a Loss Of AC Power to the plant auxiliaries (LOAP) quite well and it has been shown that PRHRS satisfies the system design requirements. During the LOAP transient S/Gs are functioning as heat sinks until boiled dry, but, they act as heat sources after the primary system is cooled. Analyzed results for the reactor trip muamtime from 26 to 70 seconds after the loss of normal feedwater by adjusting S/G low level trip setpoints and for the PRHRS actuation delay time from 300 to 1,500 seconds show that there are minor effects on the PRHRS performance for the reactor trip time and for the actuation delay time until 1,200 seconds, respectively

  6. Bistable energy harvesting enhancement with an auxiliary linear oscillator

    International Nuclear Information System (INIS)

    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. (paper)

  7. Dynamic analysis of reactor auxiliary buildings

    International Nuclear Information System (INIS)

    A review of structural methods of generalized use in the dynamic analysis of Auxiliary Buildings and similar structures of Nuclear Power Plants is presented. Emphasis is placed on the structural response to blast and seismic loading studied from a global view point. Alternative models for the representation of both element and global stiffnesses are discussed. The assumption of rigid floor behaviour for lateral force excitation is studied. Advantages of using multi-stick models are referred and illustrated. The occurrence of torsional motions on the response is examined. The study evidences the importance of the low aspect ratio of these structures and shows its influence on parameters currently used in design of conventional buildings. (Author)

  8. Performance Analysis of APR+ PAFS for CDF evaluation

    International Nuclear Information System (INIS)

    The Advanced Power Reactor Plus (APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system (PAFS) has been adopted. The PAFS replaces the conventional active auxiliary feedwater system (AFWS) by introducing a natural driving force mechanism while maintaining the system functions of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional active auxiliary feedwater system (AFWS), it is necessary to verify its cooling capacity for the core damage frequency (CDF) evaluation. This paper discusses the cooling performance of the PAFS in transient accidents

  9. Performance Analysis of APR+ PAFS for CDF evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Sang Hee; Kim, Han Gon [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2011-05-15

    The Advanced Power Reactor Plus (APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system (PAFS) has been adopted. The PAFS replaces the conventional active auxiliary feedwater system (AFWS) by introducing a natural driving force mechanism while maintaining the system functions of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional active auxiliary feedwater system (AFWS), it is necessary to verify its cooling capacity for the core damage frequency (CDF) evaluation. This paper discusses the cooling performance of the PAFS in transient accidents

  10. 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.

  11. Auxiliary fields in the geometrical relativistic particle dynamics

    International Nuclear Information System (INIS)

    We describe how to construct the dynamics of relativistic particles, following either timelike or null curves, by means of an auxiliary variables method instead of the standard theory of deformations for curves. There are interesting physical particle models governed by actions that involve higher order derivatives of the embedding functions of the worldline. We point out that the mechanical content of such models can be extracted wisely from a lower order action, which can be performed by implementing in the action a finite number of constraints that involve the geometrical relationship structures inherent to a curve and by using a covariant formalism. We emphasize our approach for null curves. For such systems, the natural time parameter is a pseudo-arclength whose properties resemble those of the standard proper time. We illustrate the formalism by applying it to some models for relativistic particles

  12. Auxiliary fields in the geometrical relativistic particle dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Amador, A; Bagatella, N; Rojas, E [Departamento de Fisica, Facultad de Fisica e Inteligencia Artificial, Universidad Veracruzana, 91000 Xalapa, Veracruz (Mexico); Cordero, R [Departamento de Fisica, Escuela Superior de Fisica y Matematicas del I.P.N, Edificio 9, 07738 Mexico D.F (Mexico)], E-mail: aramador@gmail.com, E-mail: nbagatella@uv.mx, E-mail: cordero@esfm.ipn.mx, E-mail: efrojas@uv.mx

    2008-03-21

    We describe how to construct the dynamics of relativistic particles, following either timelike or null curves, by means of an auxiliary variables method instead of the standard theory of deformations for curves. There are interesting physical particle models governed by actions that involve higher order derivatives of the embedding functions of the worldline. We point out that the mechanical content of such models can be extracted wisely from a lower order action, which can be performed by implementing in the action a finite number of constraints that involve the geometrical relationship structures inherent to a curve and by using a covariant formalism. We emphasize our approach for null curves. For such systems, the natural time parameter is a pseudo-arclength whose properties resemble those of the standard proper time. We illustrate the formalism by applying it to some models for relativistic particles.

  13. 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 u

  14. 47 CFR 80.290 - Auxiliary receiving antenna.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 5 2010-10-01 2010-10-01 false Auxiliary receiving antenna. 80.290 Section 80... antenna. An auxiliary receiving antenna must be provided when necessary to avoid unauthorized interruption or reduced efficiency of the required watch because the normal receiving antenna is not...

  15. 基于多参数空气源热泵辅助太阳能热水机组的控制系统%A control system design for solar hot water system with auxiliary air source heat pump based on multiple parameters

    Institute of Scientific and Technical Information of China (English)

    丁鸿昌; 杨前明; 刘其会; 吕帅

    2013-01-01

    The solar hot water system aided by air source heat pump can overcome the drawback of weather influence on single solar hot water system.Based on PLC controller,a control system,which considers multiple parameters such as the sun conditions,water tank temperature,ambient temperature,pipe line pressure,etc.,is designed for a solar hot water system with auxiliary air source heat pump to achieve automatic all-weather hot water supply and save electric energy effectively.%采用了空气源热泵辅助的太阳能热水机组,以解决单一太阳能热水系统受天气条件影响的问题.以PLC控制器为核心,综合考虑光照条件、环境温度、水箱水温、供水压力等多个参数条件,设计了基于多参数的空气源热泵辅助太阳能热水机组的自动控制系统,实现了以太阳能为主热源的热水机组全天候稳定热水供应,有效地节约了能源.

  16. Indirect combustion noise of auxiliary power units

    Science.gov (United States)

    Tam, Christopher K. W.; Parrish, Sarah A.; Xu, Jun; Schuster, Bill

    2013-08-01

    Recent advances in noise suppression technology have significantly reduced jet and fan noise from commercial jet engines. This leads many investigators in the aeroacoustics community to suggest that core noise could well be the next aircraft noise barrier. Core noise consists of turbine noise and combustion noise. There is direct combustion noise generated by the combustion processes, and there is indirect combustion noise generated by the passage of combustion hot spots, or entropy waves, through constrictions in an engine. The present work focuses on indirect combustion noise. Indirect combustion noise has now been found in laboratory experiments. The primary objective of this work is to investigate whether indirect combustion noise is also generated in jet and other engines. In a jet engine, there are numerous noise sources. This makes the identification of indirect combustion noise a formidable task. Here, our effort concentrates exclusively on auxiliary power units (APUs). This choice is motivated by the fact that APUs are relatively simple engines with only a few noise sources. It is, therefore, expected that the chance of success is higher. Accordingly, a theoretical model study of the generation of indirect combustion noise in an Auxiliary Power Unit (APU) is carried out. The cross-sectional areas of an APU from the combustor to the turbine exit are scaled off to form an equivalent nozzle. A principal function of a turbine in an APU is to extract mechanical energy from the flow stream through the exertion of a resistive force. Therefore, the turbine is modeled by adding a negative body force to the momentum equation. This model is used to predict the ranges of frequencies over which there is a high probability for indirect combustion noise generation. Experimental spectra of internal pressure fluctuations and far-field noise of an RE220 APU are examined to identify anomalous peaks. These peaks are possible indirection combustion noise. In the case of the

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

    International Nuclear Information System (INIS)

    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 Ram = 102 to Ram = 105. 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)

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

    International Nuclear Information System (INIS)

    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

  19. Syntactic Analyses for Parallel Grammars Auxiliaries and Genitive NPs

    CERN Document Server

    Butt, M; Rohrer, C; Butt, Miriaam; Fortman, Christian; Rohrer, Christian

    1996-01-01

    This paper focuses on two disparate aspects of German syntax from the perspective of parallel grammar development. As part of a cooperative project, we present an innovative approach to auxiliaries and multiple genitive NPs in German. The LFG-based implementation presented here avoids unnessary structural complexity in the representation of auxiliaries by challenging the traditional analysis of auxiliaries as raising verbs. The approach developed for multiple genitive NPs provides a more abstract, language independent representation of genitives associated with nominalized verbs. Taken together, the two approaches represent a step towards providing uniformly applicable treatments for differing languages, thus lightening the burden for machine translation.

  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)

    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. Strategic analysis of a manufacturing company and the North American auxiliary power unit market

    OpenAIRE

    Fisher, Craig B.

    2005-01-01

    Power Systems has been the leader in the truck auxiliary power unit business since 2000 and has been making significant advancements in the price of the product as well as its reliability and robustness. Unfortunately the reliability and robustness requirement is increasing further and the profit margin is shrinking to the point that Power Systems needs to undertake a new product development program. The new product offering will result in zero increase in the price to the customer but provid...

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

    International Nuclear Information System (INIS)

    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

  3. Depressurization as an accident management strategy for Jose Cabrera nuclear plant loss of feedwater and station blackout events

    International Nuclear Information System (INIS)

    This paper reports on an evaluation of the efficiency of the operator initiated depressurization in the Spanish Westinghouse one loop Jose Cabrera nuclear power plant that has been developed. This operation is recommended in the present emergency procedure for the total loss of feedwater event in the bleed and feed mode. RELAP5/MOD2 analyses show that this is an effective measure to bring the plant to a cold and stable condition in a design-based accident scenario

  4. Generic evaluation of feedwater transients and small break loss-of-coolant accidents in combustion engineering designed operating plants

    International Nuclear Information System (INIS)

    The purpose of this report is to summarize the results of a generic evaluation of feedwater transients, small break loss-of-coolant accidents (LOCAs), and other TMI-2-related events in the Combustion Engineering (CE)-designed operating plants and to establish or confirm the bases for their continued operation. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas

  5. Next-Generation Evaporative Cooling Systems for the Advanced Extravehicular Mobility Unit Portable Life Support System

    Science.gov (United States)

    Makinen, Janice V.; Anchondo, Ian; Bue, Grant C.; Campbell, Colin; Colunga, Aaron

    2012-01-01

    The development of the Advanced Extravehicular Mobility Unit (AEMU) Portable Life Support System (PLSS) is currently underway at NASA Johnson Space Center. The AEMU PLSS features two new evaporative cooling systems, the Reduced Volume Prototype Spacesuit Water Membrane Evaporator (RVP SWME), and the Auxiliary Cooling Loop (ACL). The RVP SWME is the third generation of hollow fiber SWME hardware, and like its predecessors, RVP SWME provides nominal crewmember and electronics cooling by flowing water through porous hollow fibers. Water vapor escapes through the hollow fiber pores, thereby cooling the liquid water that remains inside of the fibers. This cooled water is then recirculated to remove heat from the crewmember and PLSS electronics. Major design improvements, including a 36% reduction in volume, reduced weight, and more flight like back-pressure valve, facilitate the packaging of RVP SWME in the AEMU PLSS envelope. In addition to the RVP SWME, the Auxiliary Cooling Loop (ACL), was developed for contingency crewmember cooling. The ACL is a completely redundant, independent cooling system that consists of a small evaporative cooler--the Mini Membrane Evaporator (Mini-ME), independent pump, independent feed-water assembly and independent Liquid Cooling Garment (LCG). The Mini-ME utilizes the same hollow fiber technology featured in the RVP SWME, but is only 25% of the size of RVP SWME, providing only the necessary crewmember cooling in a contingency situation. The ACL provides a number of benefits when compared with the current EMU PLSS contingency cooling technology; contingency crewmember cooling can be provided for a longer period of time, more contingency situations can be accounted for, no reliance on a Secondary Oxygen Vessel (SOV) for contingency cooling--thereby allowing a SOV reduction in size and pressure, and the ACL can be recharged-allowing the AEMU PLSS to be reused, even after a contingency event. The development of these evaporative cooling

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

    Institute of Scientific and Technical Information of China (English)

    Ma Xin; Liu Wei; Chen Ling; Li Xiao; Jia Zhenyuan; Shang Zhiliang

    2016-01-01

    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 param-eters 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 pre-sented. 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.

  7. Organization and scheduling of auxiliary components and large equipments erection and control

    International Nuclear Information System (INIS)

    The first part of this paper deals with the installation of auxiliary systems. FRAMATOME divides NSSS site construction activities into the following 3 packages: main primary system, electricity and instrumentation and control, and auxiliary systems. This paper presents an analysis of the organisation of auxiliary system installation with particular reference to piping. The term ''auxiliary systems'' covers all NSSS mechanical and electrical equipment with the exception of heavy Reactor Coolant System components (reactor, fuel, associated handling equipment and cooling loops). FRAMATOME, with the help of its associates working on the French nuclear programme, has been able to develop efficient methods of organisation as a result of the wide experience gained. The main field involved is that of piping, for which computerized task analysis, scheduling and follow-up methods have been developed. The second part deals with large components installation. The normal sequence of welding operations for the assembly of the reactor coolant loops of a PWR power plant begins with the simultaneous welding of hot and cold leg piping and is completed by the welding of the crossover leg between the steam generator and the reactor coolant pump, thus ''closing'' the loop. This is the traditional method. FRAMATOME has developed a different method whereby installation of the reactor coolant loops begins with the welding of the steam generator - reactor coolant pump connection. This is the ''crossover leg'' method. This report presents a description of both methods followed by the results of a comparative study of technical considerations and schedules; in conclusion, charts are presented to assist in pre-selection of the method which is best suited to the on-site delivery schedule of Reactor Coolant System components

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

    International Nuclear Information System (INIS)

    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.)

  9. Bayesian auxiliary variable models for binary and multinomial regression

    OpenAIRE

    Holmes, C C; HELD, L.

    2006-01-01

    In this paper we discuss auxiliary variable approaches to Bayesian binary and multinomial regression. These approaches are ideally suited to automated Markov chain Monte Carlo simulation. In the first part we describe a simple technique using joint updating that improves the performance of the conventional probit regression algorithm. In the second part we discuss auxiliary variable methods for inference in Bayesian logistic regression, including covariate set uncertainty. Fina...

  10. A Parallel Auxiliary Grid AMG Method for GPU

    OpenAIRE

    Wang, Lu; Hu, Xiaozhe; Cohen, Jonathan; Xu, Jinchao

    2012-01-01

    In this paper, we develop a new parallel auxiliary grid algebraic multigrid (AMG) method to leverage the power of graphic processing units (GPUs). In the construction of the hierarchical coarse grid, we use a simple and fixed coarsening procedure based on a region quadtree generated from an auxiliary grid. This allows us to explicitly control the sparsity patterns and operator complexities of the AMG solver. This feature provides (nearly) optimal load balancing and predictable communication p...

  11. New braneworld models in the presence of auxiliary fields

    CERN Document Server

    Bazeia, D; Menezes, R; Moreira, D C

    2014-01-01

    We study braneworld models in the presence of auxiliary fields. We use the first-order framework to investigate several distinct possibilities, where the standard braneworld scenario changes under the presence of the parameter that controls the auxiliary fields introduced to modify Einstein's equation. The results add to previous ones, to show that the minimal modification that we investigate contributes to change quantitatively the thick braneworld profile, although no new qualitative effect is capable of being induced by the minimal modification here considered.

  12. Alphabet Sizes of Auxiliary Variables in Canonical Inner Bounds

    OpenAIRE

    Jana, Soumya

    2008-01-01

    Alphabet size of auxiliary random variables in our canonical description is derived. Our analysis improves upon estimates known in special cases, and generalizes to an arbitrary multiterminal setup. The salient steps include decomposition of constituent rate polytopes into orthants, translation of a hyperplane till it becomes tangent to the achievable region at an extreme point, and derivation of minimum auxiliary alphabet sizes based on Caratheodory's theorem.

  13. SEPP-ZVS High Frequency Inverter Incorporating Auxiliary Switch

    Science.gov (United States)

    Ogiwara, Hiroyuki; Itoi, Misao; Nakaoka, Mutsuo

    This paper presents a novel circuit topology to attain ZVS operation of a high frequency inverter over a wide range output power regulation using a PWM control technique by connecting an auxiliary switch to the conventional single ended push-pull (SEPP) ZVS high frequency inverter. A switching current is injected into the main switches via the auxiliary switch only during the short period between its turn-on and off times to supply a current required for its ZVS operation.

  14. Soft switching DC/DC converter using auxiliary circuits

    OpenAIRE

    Jaroslav Dudrik; Vladimír Ruščin

    2008-01-01

    A novel auxiliary circuit for full-bridge PWMDC/DC converter with controlled secondary side rectifieris presented in this paper. Zero-current turn-on for allpower switches of the inverter is achieved for full loadrange from no-load to short circuit by using controlledrectifier and auxiliary circuit on the secondary side.Modified phase shift PWM control strategy is used for theconverter. The principle of operation is explained andanalyzed and the simulation results are presented.

  15. A full system decontamination of the Oskarshamn 1 BWR

    International Nuclear Information System (INIS)

    Due to environmentally assisted cracking in cold-worked bends in the primary piping, approximately 600 meters were subject to pipe replacement. The collective dose budget for the entire work was estimated to 6.5 manSv with traditional lead shielding. With a decontamination factor of five, the collective dose was estimated to approximately 3.4 manSv. It was decided that a decontamination would be cost effective. Because of up-coming inspection and repair work inside the reactor pressure vessel a decontamination of the lower part of the vessel and the main recirculation loops was demanded. A chemical system decontamination with the CORD-method was performed. The systems involved in the first phase were the residual heat removal system, the reactor water clean up system, pipe connections to the auxiliary condenser and a minor part of the final feedwater system. In the second phase, the four main recirculation loops and the lower part of the reactor pressure vessel were involved. For the first time ever, during a large decontamination, a major part of the decontamination chemicals were decomposed in-situ and the final waste volume was significantly reduced. (authors). 6 figs

  16. In-service inspection techniques for PWR steam generator feedwater and pressuriser nozzles

    International Nuclear Information System (INIS)

    Regular ultrasonic inspection of the steam generator feedwater and pressuriser nozzles of the Sizewell B Pressurised Water Reactor, under construction by the Central Electricity Generating Board (CEGB), will be carried out to detect and size any service-induced cracking in the nozzle corners and bores. External access only will be available for such inspections and, to achieve full inspection coverage, it may be necessary to scan probes under the outer blend radius and adjacent surfaces of each nozzle. As part of the PWR Safety Research Programme being conducted collaboratively by CEGB and the United Kingdom Atomic Energy Authority (UKAEA) in the UK, Risley Laboratories have been developing automated ultrasonic techniques to meet the stringent inspection standards demanded for these components. The geometrical complexity of the inspection has necessitated the use of mathematical modelling to optimise inspection techniques and coverage. Computer-based data collection, display and analysis methods have been developed for the combined pulse-echo and time-of-flight diffraction techniques selected for these inspections. In this paper, the development work performed at Risley is reviewed and examples of the application of the inspection techniques to simulated service-induced defects in full-scale test specimens presented. Work on the related problem of cracking in BWR nozzles (ref.1) has demonstrated detection capability but not accurate sizing. However sizing of defects in the nozzle corner areas has been achieved in this programme with the use of advanced reconstruction methods. (author)

  17. Loading conditions in horizontal feedwater pipes of LWRs influenced by thermal shock and thermal stratification effects

    International Nuclear Information System (INIS)

    In recent years, a particular form of crack formation has occurred in a number of pressurized and boiling water reactors on the integral surfaces of horizontal lengths of feedwater piping upstream of the steam generators and reactor pressure vessels. The fractographic evaluation and the orientation of the cracks show that these are to be attributed to cyclic stressing in the axial direction. Comparison of the stresses due to thermal shock and thermal stratification reveals that, on account of the associated load cycles, the cracks were in essence caused by thermal stratification. This is also indicated by the orientation of the cracks. The present results of the corrosion tests show that with high oxygen content (450 ppb) and temperature level (2100C) the strain rate at thermal stratification exerts an essential influence on the number of cycles to crack initiation. With the conservative test conditions described, the values may fall below those given in the ASME fatigue design curve. In the case of strain rates that apply to thermal shock, the cycles to crack initiation are on the safe side of the curve. The remedial measures taken by KWU by the installation of the siphon reduce the frequency of stress amplitudes. It can be concluded from corrosion tests simulating these conditions that with the strain rates occurring in this case, the number of cycles to crack initiation are on the safe side according to the ASME fatigue design curve. (orig.)

  18. Dismantling and decontamination of the tube bundle of a feedwater preheater of the Garigliano BWR

    International Nuclear Information System (INIS)

    The report deals with dismantling and decontamination of the tube bundle of a feedwater preheater of Garigliano-BWR. Decontamination is a common practice in decommissioning works and it can be used both for reducing radiation exposures, in order to save manrem, and for the unrestricted release of materials. In this latter field the decontamination of tube bundle is a particular case because of their large contaminated surfaces and relatively low weight; at the moment no decon technique was available on the market to decontaminate up to the unrestricted release the materials of tube bundles. In this context an innovative decon technique using aggressive chemicals together with ultrasounds in a tank, was developed by several laboratories and assessed with in-scale testings. A decon procedure considering two phases: first with ultrasounds applied in water at 600C, and second with ultrasounds applied in a solution of HF/HNO3 acids at 60-700C, was qualified. The demonstration of the performances of the new technique under real conditions was made by performing ten full-scale demo tests, each one on an assembly of 100 straight tubes, 1 m long each, for a total of 1000 meters

  19. GEOTHERMAL / SOLAR HYBRID DESIGNS: USE OF GEOTHERMAL ENERGY FOR CSP FEEDWATER HEATING

    Energy Technology Data Exchange (ETDEWEB)

    Craig Turchi; Guangdong Zhu; Michael Wagner; Tom Williams; Dan Wendt

    2014-10-01

    This paper examines a hybrid geothermal / solar thermal plant design that uses geothermal energy to provide feedwater heating in a conventional steam-Rankine power cycle deployed by a concentrating solar power (CSP) plant. The geothermal energy represents slightly over 10% of the total thermal input to the hybrid plant. The geothermal energy allows power output from the hybrid plant to increase by about 8% relative to a stand-alone CSP plant with the same solar-thermal input. Geothermal energy is converted to electricity at an efficiency of 1.7 to 2.5 times greater than would occur in a stand-alone, binary-cycle geothermal plant using the same geothermal resource. While the design exhibits a clear advantage during hybrid plant operation, the annual advantage of the hybrid versus two stand-alone power plants depends on the total annual operating hours of the hybrid plant. The annual results in this draft paper are preliminary, and further results are expected prior to submission of a final paper.

  20. 30 CFR 57.8534 - Shutdown or failure of auxiliary fans.

    Science.gov (United States)

    2010-07-01

    ... auxiliary fan failure due to malfunction, accident, power failure, or other such unplanned or unscheduled... 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...

  1. A PPS Sampling Scheme Using Harmonic Mean of An Auxiliary Variable

    OpenAIRE

    L.N. Sahoo; Dalabehera, M.; A.K. Mangaraj

    2012-01-01

    We consider a probability proportional to size sampling scheme by using harmonic mean of an auxiliary variable, when the correlation between study variable and auxiliary variable is highly negative. This is achieved by considering inverse transformation of the auxiliary variable and then utilizing the transformed auxiliary variable values at the design stage of the survey operation in selecting a sample.

  2. VLTI First Fringes with Two Auxiliary Telescopes at Paranal

    Science.gov (United States)

    2005-03-01

    [Preview - JPEG: 537 x 400 pix - 31k] [Normal - JPEG: 1074 x 800 pix - 555k] [HiRes - JPEG: 3000 x 2235 pix - 6.0M] ESO PR Photo 07d/05 ESO PR Photo 07d/05 [Preview - JPEG: 400 x 550 pix - 60k] [Normal - JPEG: 800 x 1099 pix - 946k] [HiRes - JPEG: 2414 x 3316 pix - 11.0M] Captions: ESO PR Photo 07b/05 shows VLTI Auxiliary Telescopes 1 and 2 (AT1 and AT2) in the early evening light, with the spherical domes opened and ready for observations. In ESO PR Photo 07c/05, the same scene is repeated later in the evening, with three of the large telescope enclosures in the background. This photo and ESO PR Photo 07c/05 which is a time-exposure with AT1 and AT2 under the beautiful night sky with the southern Milky Way band were obtained by ESO staff member Frédéric Gomté. However, most of the time the large telescopes are used for other research purposes. They are therefore only available for interferometric observations during a limited number of nights every year. Thus, in order to exploit the VLTI each night and to achieve the full potential of this unique setup, some other (smaller), dedicated telescopes were included into the overall VLT concept. These telescopes, known as the VLTI Auxiliary Telescopes (ATs), are mounted on tracks and can be placed at precisely defined "parking" observing positions on the observatory platform. From these positions, their light beams are fed into the same common focal point via a complex system of reflecting mirrors mounted in an underground system of tunnels. The Auxiliary Telescopes are real technological jewels. They are placed in ultra-compact enclosures, complete with all necessary electronics, an air conditioning system and cooling liquid for thermal control, compressed air for enclosure seals, a hydraulic plant for opening the dome shells, etc. Each AT is also fitted with a transporter that lifts the telescope and relocates it from one station to another. It moves around with its own housing on the top of Paranal, almost like a snail

  3. Dynamic behavior of a magnetic bearing supported jet engine rotor with auxiliary bearings

    Science.gov (United States)

    Flowers, George T.; Xie, Huajun; Sinha, S. C.

    1995-01-01

    This paper presents a study of the dynamic behavior of a rotor system supported by auxiliary bearings. The steady-state behavior of a simulation model based upon a production jet engine is explored over a wide range of operating conditions for varying rotor imbalance, support stiffness, and damping. Interesting dynamical phenomena, such as chaos, subharmonic responses, and double-valued responses, are presented and discussed.

  4. AMPA Receptors Commandeer an Ancient Cargo Exporter for Use as an Auxiliary Subunit for Signaling

    OpenAIRE

    Nadine Harmel; Barbara Cokic; Gerd Zolles; Henrike Berkefeld; Veronika Mauric; Bernd Fakler; Valentin Stein; Nikolaj Klöcker

    2012-01-01

    Fast excitatory neurotransmission in the mammalian central nervous system is mainly mediated by ionotropic glutamate receptors of the AMPA subtype (AMPARs). AMPARs are protein complexes of the pore-lining alpha-subunits GluA1-4 and auxiliary beta-subunits modulating their trafficking and gating. By a proteomic approach, two homologues of the cargo exporter cornichon, CNIH-2 and CNIH-3, have recently been identified as constituents of native AMPARs in mammalian brain. In heterologous reconstit...

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

    International Nuclear Information System (INIS)

    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

  6. Practical auxiliary basis implementation of Rung 3.5 functionals.

    Science.gov (United States)

    Janesko, Benjamin G; Scalmani, Giovanni; Frisch, Michael J

    2014-07-21

    Approximate exchange-correlation functionals for Kohn-Sham density functional theory often benefit from incorporating exact exchange. Exact exchange is constructed from the noninteracting reference system's nonlocal one-particle density matrix γ(r(->), r(->)'). Rung 3.5 functionals attempt to balance the strengths and limitations of exact exchange using a new ingredient, a projection of γ(r(->), r(->)') onto a semilocal model density matrix γ(SL)(ρ(r(->)), ∇ρ(r(->)), r(->) - r(->)'). γSL depends on the electron density ρ(r(->) at reference point r(->), and is closely related to semilocal model exchange holes. We present a practical implementation of Rung 3.5 functionals, expanding the r(->) - r(->)' dependence of γSL in an auxiliary basis set. Energies and energy derivatives are obtained from 3D numerical integration as in standard semilocal functionals. We also present numerical tests of a range of properties, including molecular thermochemistry and kinetics, geometries and vibrational frequencies, and bandgaps and excitation energies. Rung 3.5 functionals typically provide accuracy intermediate between semilocal and hybrid approximations. Nonlocal potential contributions from γSL yield interesting successes and failures for band structures and excitation energies. The results enable and motivate continued exploration of Rung 3.5 functional forms. PMID:25053297

  7. Green Propulsion Auxiliary Power Unit Demonstration at MSFC

    Science.gov (United States)

    Robinson, Joel W.

    2014-01-01

    In 2012, the National Aeronautics & Space Administration (NASA) Space Technology Mission Directorate (STMD) began the process of building an integrated technology roadmap, including both technology pull and technology push strategies. Technology Area 1 (TA-01)1 for Launch Propulsion Systems is one of fourteen TAs that provide recommendations for the overall technology investment strategy and prioritization of NASA's space technology activities. Identified within TA-01 was the need for a green propulsion auxiliary power unit (APU) for hydraulic power by 2015. Engineers led by the author at the Marshall Space Flight Center (MSFC) have been evaluating green propellant alternatives and have begun the development of an APU test bed to demonstrate the feasibility of use. NASA has residual APU assets remaining from the retired Space Shuttle Program. Likewise, the F-16 Falcon fighter jet also uses an Emergency Power Unit (EPU) that has similar characteristics to the NASA hardware. Both EPU and APU components have been acquired for testing at MSFC. This paper will summarize the status of the testing efforts of green propellant from the Air Force Research Laboratory (AFRL) propellant AFM315E based on hydroxyl ammonium nitrate (HAN) with these test assets.

  8. Vehicle Integrated Photovoltaics for Compression Ignition Vehicles: An Experimental Investigation of Solar Alkaline Water Electrolysis for Improving Diesel Combustion and a Solar Charging System for Reducing Auxiliary Engine Loads

    Science.gov (United States)

    Negroni, Garry Inocentes

    Vehicle-integrated photovoltaic electricity can be applied towards aspiration of hydrogen-oxygen-steam gas produced through alkaline electrolysis and reductions in auxiliary alternator load for reducing hydrocarbon emissions in low nitrogen oxide indirect-injection compression-ignition engines. Aspiration of 0.516 ± 0.007 liters-per-minute of gas produced through alkaline electrolysis of potassium-hydroxide 2wt.% improves full-load performance; however, part-load performance decreases due to auto-ignition of aspirated gas prior to top-dead center. Alternator load reductions offer improved part-load and full-load performance with practical limitations resulting from accessory electrical loads. In an additive approach, solar electrolysis can electrochemically convert solar photovoltaic electricity into a gas comprised of stoichiometric hydrogen and oxygen gas. Aspiration of this hydrogen-oxygen gas enhances combustion properties decreasing emissions and increased combustion efficiency in light-duty diesel vehicles. The 316L stainless steel (SS) electrolyser plates are arranged with two anodes and three cathodes space with four bipolar plates delineating four stacks in parallel with five cells per stack. The electrolyser was tested using potassium hydroxide 2 wt.% and hydronium 3wt.% at measured voltage and current inputs. The flow rate output from the reservoir cell was measured in parallel with the V and I inputs producing a regression model correlating current input to flow rate. KOH 2 wt.% produced 0.005 LPM/W, while H9O44 3 wt.% produced less at 0.00126 LPM/W. In a subtractive approach, solar energy can be used to charge a larger energy storage device, as is with plug-in electric vehicles, in order to alleviate the engine of the mechanical load placed upon it by the vehicles electrical accessories through the alternator. Solar electrolysis can improve part-load emissions and full-load performance. The average solar-to-battery efficiency based on the OEM rated

  9. The use of the vinegar as a chemical auxiliary in Endodontics: a literature review

    Directory of Open Access Journals (Sweden)

    Luis Eduardo Duarte IRALA

    2009-06-01

    Full Text Available Introduction: During the root canal instrumentation with the use of files, there is the formation of a layer called “smear layer”, which may hide among other elements microorganisms capable of perpetuating an endodontic infection, what can lead to endodontic treatment failure. Objective: To present a literature review of apple vinegar and its chemical properties as an auxiliary alternative substance in the permeability in dentin. Literature review: In the attempt to remove the smear layer, auxiliary substances that allow a better action of instruments are used, which also permit the permeability of the system of canals and the removal of organic remains and possible contaminants. Conclusion: The results had indicated that the use of the apple vinegar can be a viable alternative to assist chemistry in Endodontics.

  10. Evolving Neural Network Using Genetic Algorithm for Faults Diagnosis of Urban Rail Vehicle Auxiliary Inverter

    Directory of Open Access Journals (Sweden)

    Ji Changxu

    2013-07-01

    Full Text Available In this article, an efficient method is proposed to diagnose urban rail vehicle auxiliary inverter faults based on wavelet packet neural network and genetic algorithm. Firstly, the original signals are decomposed into different frequency subbands by wavelet packet. Secondly, the wavelet packet energy eigenvector is constructed. Finally, those wavelet packet energy eigenvectors are taken as fault samples to train neural network, In order to improve the function approximation accuracy and general capability of the neural network system, an efficient genetic algorithm approach is used to adjust the parameters of translation and weights functions. The experiment shows that the GA-ANN model gives superior result. This approach can be used as a useful tool for the auxiliary inverter fault diagnosis.

  11. A Scalable Auxiliary Space Preconditioner for High-Order Finite Element Methods

    CERN Document Server

    Lee, Young-Ju; Zhang, Chen-Song

    2012-01-01

    In this paper, we revisit an auxiliary space preconditioning method proposed by Xu [Computing 56, 1996], in which low-order finite element spaces are employed as auxiliary spaces for solving linear algebraic systems arising from high-order finite element discretizations. We provide a new convergence rate estimate and parallel implementation of the proposed algorithm. We show that this method is user-friendly and can play an important role in a variety of Poisson-based solvers for more challenging problems such as the Navier--Stokes equation. We investigate the performance of the proposed algorithm using the Poisson equation and the Stokes equation on 3D unstructured grids. Numerical results demonstrate the advantages of the proposed algorithm in terms of efficiency, robustness, and parallel scalability.

  12. Accuracy of marker-assisted selection with auxiliary traits

    Indian Academy of Sciences (India)

    P Narain

    2003-09-01

    Genetic information on molecular markers is increasingly being used in plant and animal improvement programmes particularly as indirect means to improve a metric trait by selection either on an individual basis or on the basis of an index incorporating such information. This paper examines the utility of an index of selection that not only combines phenotypic and molecular information on the trait under improvement but also combines similar information on one or more auxiliary traits. The accuracy of such a selection procedure has been theoretically studied for sufficiently large populations so that the effects of detected quantitative trait loci can be perfectly estimated. The theory is illustrated numerically by considering one auxiliary trait. It is shown that the use of an auxiliary trait improves the selection accuracy; and, hence, the relative efficiency of index selection compared to individual selection which is based on the same intensity of selection. This is particularly so for higher magnitudes of residual genetic correlation and environmental correlation having opposite signs, lower values of the proportion of genetic variation in the main trait associated with the markers, negligible proportion of genetic variation in the auxiliary trait associated with the markers, and lower values of the heritability of the main trait but higher values of the heritability of the auxiliary trait.

  13. Stochastic gradient algorithm for a dual-rate Box-Jenkins model based on auxiliary model and FIR model

    Institute of Scientific and Technical Information of China (English)

    Jing CHEN; Rui-feng DING

    2014-01-01

    Based on the work in Ding and Ding (2008), we develop a modifi ed stochastic gradient (SG) parameter estimation algorithm for a dual-rate Box-Jenkins model by using an auxiliary model. We simplify the complex dual-rate Box-Jenkins model to two fi nite impulse response (FIR) models, present an auxiliary model to estimate the missing outputs and the unknown noise variables, and compute all the unknown parameters of the system with colored noises. Simulation results indicate that the proposed method is effective.

  14. Improving Semi-Supervised Learning with Auxiliary Deep Generative Models

    DEFF Research Database (Denmark)

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

    Deep generative models based upon continuous variational distributions parameterized by deep networks give state-of-the-art performance. In this paper we propose a framework for extending the latent representation with extra auxiliary variables in order to make the variational distribution more...... expressive for semi-supervised learning. By utilizing the stochasticity of the auxiliary variable we demonstrate how to train discriminative classifiers resulting in state-of-the-art performance within semi-supervised learning exemplified by an 0.96% error on MNIST using 100 labeled data points. Furthermore...

  15. Auxiliary tensor fields for Sp(2,R) self-duality

    CERN Document Server

    Ivanov, Evgeny A; Zupnik, Boris M

    2014-01-01

    The coset Sp(2,R)/U(1) is parametrized by two real scalar fields. We generalize the formalism of auxiliary tensor (bispinor) fields in U(1) self-dual nonlinear models of abelian gauge fields to the case of Sp(2,R) self-duality. In this new formulation, Sp(2,R) duality of the nonlinear scalar-gauge equations of motion is equivalent to an Sp(2,R) invariance of the auxiliary interaction. We derive this result in two different ways, aiming at its further application to supersymmetric theories. We also consider an extension to interactions with higher derivatives.

  16. TMI-2 auxiliary building elevator shaft and pit decontamination

    International Nuclear Information System (INIS)

    Decontamination of the elevator pit and shaft in the auxiliary building at Three Mile Island Unit 2 (TMI-2) was performed to remove high radiation and contamination levels which prevented personnel from utilizing the elevator. The radiation and contamination levels in the TMI-2 auxiliary building elevator shaft have been reduced to the point where plant personnel are again permitted to ride in the elevator without a radiation work permit, with the exception of access to the 281-ft (basement) level. Based on the declassification and expanded use of the elevator, the task goal has been met. The tax expended 16.16 man-rem and 621 man-hours

  17. Vanishing auxiliary variables in PPS sampling - with applications in microscopy

    DEFF Research Database (Denmark)

    Andersen, Ina Trolle; Hahn, Ute; Jensen, Eva B. Vedel

    auxiliary variables are a common phenomenon in microscopy and, accordingly, part of the population is not accessible, using PPS sampling. We propose a modification of the design, for which an optimal solution can be found, using a model assisted approach. The optimal design has independent interest in......Recently, non-uniform sampling has been suggested in microscopy to increase efficiency. More precisely, sampling proportional to size (PPS) has been introduced where the probability of sampling a unit in the population is proportional to the value of an auxiliary variable. Unfortunately, vanishing...

  18. Response of the C-E owners group to the loss-of-feedwater event at the Davis-Besse plant

    International Nuclear Information System (INIS)

    As a result of the Davis-Besse loss of all feedwater event, C-E owners group (CEOG) agreed at its September, 1985 meeting to form a review group for the purpose of evaluating the event with respect to its impact on plants with a C-E nuclear steam supply system (NSSS). Objectives of the review group included review of the Davis-Besse event, determining generic concerns applicable to CEOG plants and development of an action plan to address these concerns. The review documents the inherent design differences between CEOG plants and the Davis-Besse plant. These differences are such that a similar LOF event at a CEOG plant would have a significantly different and improved safety margin relative to that which existed during the Davis-Besse event. Although the impact of this type of event on CEOG plants would be less adverse, the review of the event did identify a number of generic issues for CEOG plants. an action plan was created to address each of the applicable generic issues

  19. 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)

    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.)

  20. Analysis on damage cause of belt eyes for bypass feedwater isolation valves of Daya Bay and Ling'ao NPPs

    International Nuclear Information System (INIS)

    When all of bypass feedwater isolation valves of Daya Bay and Ling'ao NPPs were inspected under disassembled condition, it was found that the belt eyes were damaged in various degrees. The analysis on macro appearance, chemical composition, metallographic, Scanning Electron Microscope and fluid mechanics simulation were carried out on the damaged belt eyes. The results show that the flow rate of belt eye is well above the manufacturer's recommendation so that flow accelerated corrosion and cavitation erosion occur on the belt eyes. According to the damage causes of the belt eyes, corresponding corrective actions are proposed. (authors)