Technical features of advanced boiling water reactor (ABWR)

International Nuclear Information System (INIS)

Horiuchi, Tetsuo; Takashima, Yoshie; Yokomi, Michiro

1986-01-01

As the final stage of the development of ABWRs of 1300 MWe output class carried out since 1984, the design for optimizing the plants has been performed, but it was completed at the end of 1985. Hereafter, the ABWRs will advance to the development of the actual project toward the permission and approval and the construction. By the optimization, the simplification, compacting and the heightening of thermal efficiency of the ABWRs were further promoted. The above promotion was attained by a cylindrical concrete containment vessel constructed in one body with the structure of a reactor building, the optimization of the redundancy of facilities, the optimization of equipment size, the combination of a two-stage reheating, 52 in blade turbine and the recovery of heater drain and so on. The plant characteristics such as the capacity factor, operability and the radiation exposure dose of workers were examined in detail in the process of optimization. As the results, it was shown that the ABWRs have the excellent economical efficiency and the performance characteristics of high level. The technical features of the ABWRs are large capacity and high efficiency plants, the improved core adopting internal pumps and new control rod driving mechanism, rational waste treatment facilities and so on. (Kako, I.)

Feedwater processing method in a boiling water reactor

Energy Technology Data Exchange (ETDEWEB)

Izumitani, M; Tanno, K

1976-09-06

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

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

International Nuclear Information System (INIS)

Sun Shuhai; Li Huasheng; Zhang Hao

2012-01-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Minimum throttling feedwater control in VVER-1000 and PWR NPPs

International Nuclear Information System (INIS)

Symkin, B.E.; Thaulez, F.

2004-01-01

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Optimization of the pumping ring in a mechanical seal with an integrated cooler for feed-water pumps

International Nuclear Information System (INIS)

Buchdahl, D.; Martin, R.; Gueret, G.; Blanc, M.

1994-07-01

To simplify maintenance, E.D.F. along with its collaborators undertook the study of mechanical seal with integrated cooler used in feed-water pumps in the nuclear power plants. The cooler, integrated to the pump acts as a thermal barrier as well as a cooler of the mechanical seal. The water circulation in the cooler is assumed by an integrated pumping ring in the rotary part of the mechanical seal, with a matching screw thread in the pumping case. This assembly of mechanical seal/integrated cooler is tested in a test loop at the EDF/DER Laboratory. All working conditions are similar to that at site. Tests with different configurations of the rotor/stator profiles are performed, i.e.; different lengths and types of threading. Hydraulic performances and the global thermal balance of this assembly are studied. Our basic aim during these tests is to optimize the hydraulic performance of the pumping ring so as to best cool the mechanical seal faces. The different results obtained and the conclusions drawn during these tests are presented. (authors). 7 figs., 3 refs

Aging assessment of auxiliary feedwater pumps

International Nuclear Information System (INIS)

Greenstreet, W.L.

1987-01-01

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

The development and assessment of TRACE/PARCS model for Lungmen ABWR

Energy Technology Data Exchange (ETDEWEB)

Lin, Hao-Tzu, E-mail: htlin@iner.gov.tw [Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C., 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County, 32546, Taiwan, ROC (China); Wang, Jong-Rong, E-mail: jrwang@iner.gov.tw [Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C., 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County, 32546, Taiwan, ROC (China); Chen, Hsiung-Chih, E-mail: km8385@gmail.com [Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Section 2, Kuang Fu Road, HsinChu 30013, Taiwan, ROC (China); Shih, Chunkuan, E-mail: ckshih@ess.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Section 2, Kuang Fu Road, HsinChu 30013, Taiwan, ROC (China)

2014-07-01

Highlights: • The development of a TRACE/PARCS model of Lungmen (ABWR). • Thermal-hydraulics model of TRACE is coupling with 3-D neutronics model of PARCS. • The 3-D geometry vessel component of TRACE is used in this study. • The parameters responses of Lungmen TRACE/PARCS model are consistent with FSAR data. - Abstract: This study consists of two steps. The first step is the development of a TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS (Purdue Advanced Reactor Core Simulator) model of Lungmen nuclear power plant (NPP) which includes the vessel, reactor internal pumps (RIPs), main steam lines, and important control systems (such as the feedwater control system, steam bypass and pressure control system, and recirculation flow control system), etc. Key parameters were identified to refine the TRACE/PARCS model further in the frame of a steady state analysis. The second step is the performance of Lungmen NPP TRACE/PARCS model transient analyses. The transient data of Final Safety Analysis Report (FSAR) chapter 15th were used to compare with the results of Lungmen NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results were consistent with the FSAR data. It indicated that there was a respectable accuracy in the Lungmen NPP TRACE/PARCS model and it also depicted that the Lungmen NPP TRACE/PARCS model was satisfying for the purpose of Lungmen NPP safety analyses.

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

International Nuclear Information System (INIS)

Wang Xin; Han Weishi

2010-01-01

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

Development, operating experience and future plan of ABWR in Japan

International Nuclear Information System (INIS)

Ujihara, N.

2001-01-01

From 1974 to 1978 nuclear power plants' capacity factor decreased to a low level at the Tokyo Electric Power Company (TEPCO) for stress corrosion cracking of stainless steal pipe, thermal fatigue crack of feed water spurger and vibration trouble of LPRM, etc. So, for improving that situation, we gathered the most useful BWR technology in the world at that time and we started the ABWR project to develop an ideal BWR. The goals of this development were: (1) Enhanced safety and reliability; (2) Reduced occupational radiation exposure and radioactive waste; (3) Enhanced operability and maneuverability; and (4) Improved economy. Significant improvements were induced by the adoption of the reactor internal pump, fine motion control rod drive mechanism, integral type reinforced concrete containment vessel and digital control system. After about 10 years R and D, we succeeded in the development of the compact and economic plant, ABWR (1356MW(e)) whose building volume had been reduced to 76%, compared with the previous BWR (1100MW(e)). The construction of Kashiwazaki-Kariwa Nuclear Power Plant unit No. 6, which is the first ABWR plant in the world, was started September 1991, and that of unit No. 7 was started in February 1992. We used effective construction methods, such as large block construction method and all weather construction method, to shorten the construction duration from first concrete to commercial operation start to 51.5 months. The commercial operation start of Kashiwazaki-Kariwa Nuclear Power Plant unit No. 6 was November 1996 and that of unit No. 7 was July 1997. After that, unit No. 6 had three times the annual outage and unit No. 7 had two times the annual outage. Both plants have good operating experiences. From commercial operation start to end of August 2000, the capacity factor of unit No. 6 is 86.4% and that of unit No. 7 is 86.9%. Now, in Taiwan, ABWR plants are under construction at Lungmen unit No. 1 and 2. In Japan, ABWR plants are under

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

International Nuclear Information System (INIS)

Greenstreet, W.L.

1989-01-01

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

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

International Nuclear Information System (INIS)

Tanji, J.; Omori, T.

1982-01-01

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

Feedwater recycling system in BWR type reactor

International Nuclear Information System (INIS)

Shimamoto, Yoshiharu.

1980-01-01

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

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

International Nuclear Information System (INIS)

Wang Xin; Han Weishi

2010-01-01

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

Development of Innovative Construction Technologies for ABWRs

International Nuclear Information System (INIS)

Yamashita, T.; Fukuda, T.; Masuda, N.

2002-01-01

This paper describes an effort by Tokyo Electric Power Company (TEPCO) to shorten the construction time in a drastic manner for the Advanced Boiling Water Reactors (ABWR), thereby aiming at reducing construction costs. First an outline of the actual construction records for the five BWR Units and the two ABWR Units at the Kashiwazaki-Kariwa site is introduced along with the construction methods employed for these units. There is a continued trend in the reduction of construction time from Unit 1 to 7 owing to a number of improvements made to these units, and in particular it is noteworthy that the drastic reduction was accomplished due to the change in reactor type from BWR to ABWR. Explained next is an on-going effort for the next ABWR and the following next generation of ABWRs to further shorten construction time. In this effort an emphasis is laid on the development of innovative construction methods by the adoption of steel plate/concrete composite structures (SC Structure), and the application of those structures even to containment vessel (so-called SCCV). This work is being conducted as part of TEPCO's research and development for the next generation ABWRs (ABWR II). Presently it is expected that the construction time from bedrock inspection to fuel loading could be shortened in a stepwise fashion to 38 and 31.5 months for the next ABWR and finally to 23 months for the next generation ABWR II, thus enabling a greatly reduced power generating cost and enhanced safety during construction. In addition to the above effort, a preliminary study has been performed concerning the application of base isolation systems to ABWR plants. From this study which is aimed at standardization, lowered cost and mitigation of engineering work, it was found that base isolation system effectively reduces seismic response and is instrumental in achieving the above-mentioned objectives. Therefore it can be said that the use of this system gives more freedom in selecting sites

Development of ABWR inertia-increased reactor internal pump and thicker sleeve nozzle

International Nuclear Information System (INIS)

Takahashi, Shirou; Shiina, Kouji; Matsumura, Seiichi

2002-01-01

The conventional reactor internal pumps (RIPs) in the ABWR have an inertia moment coming from the shafts and Motor-Generator sets, enabling the RIPs to continue running for a few seconds, when a trip of all RIPs event occurs. It is possible to simplify the RIPs' power supply system without affecting the core flow supply when the above event occurs by eliminating M-G sets, if the rotating inertia is increased. This inertia increase due to an additional flywheel, which leads to gains in weight and length, requires the larger diameter nozzle with the thicker sleeve. However, too large a nozzle diameter may change the hydraulic performance. In authors' previous study, the optimum nozzle diameter (492 mm) was selected through 1/5-scale test. In this study, the 492 mm nozzle and the inertia-increased RIP were verified through the full-scale tests. The rotating inertia time constant on coastdown characteristics (behavior of the RIP speed in the event of power loss) for the inertia-increased RIP doubled compared with the current RIP. The casing and the shaft vibration were also confirmed to satisfy the design criteria. Moreover, hydraulic performance and heat increase in the motor casing due to the flywheel were evaluated. The inertia increased RIP with the 492 mm nozzle maintained good performance. (author)

ABWR construction experience in Japan

International Nuclear Information System (INIS)

Kawahata, Junichi; Miura, Jun; Saito, Fuyuki; Mori, Hiroya

2003-01-01

The construction of Kashiwazaki-Kariwa Nuclear Power Station Unit No. 6 and 7 (K6 and 7) owned by Tokyo Electric Power Company (TEPCO), have been in service since November 1996 and September 1997 respectively, and are the first ABWRs in the world. Hitachi, Ltd. and Toshiba Corporation took part in the projects and achieved safety construction in 52- month schedule from rock inspection to turnover in cooperation with TEPCO and other contractors with broad experience in BWRs. Taking advantage of experience of the Kashiwazaki construction, the ABWR construction methods realize to satisfy with requests from electric power companies and to fit site conditions. The Chubu Electric Power Company's Hamaoka Nuclear Power Station Unit No. 5 (Hamaoka 5) and Hokuriku Electric Power Company's Shika Nuclear Power Station Unit No. 2 (Shika 2), which are ABWRs, completed rock inspection in May 2000 and June 2001 respectively, and are under construction. The ABWR construction methods are proved their validity through four actual construction projects. Safe constructions with the highest quality have been advanced thus further aiming at the introduction of the latest technologies and contributing to reduce plant construction costs and also to shorten schedule. (author)

Modern technology applied in the advanced BWR (ABWR)

International Nuclear Information System (INIS)

Hucik, S.A.

1988-01-01

The advanced boiling water reactor (ABWR) represents the next generation of light water reactors (LWR) to be introduced into commercial operation in the 1990's. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technology and improvements based on worldwide experience, and extensive design and test and development programs. This paper discusses how the ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. A technical evaluation of the ABWR shows its superiority in terms of performance characteristics and economics relative to current LWR designs

Human-system interfaces for the ABWR control room

International Nuclear Information System (INIS)

Gutierrez, R.; O'Neil, T.J.

1993-01-01

The 1300-MW(electric) advanced boiling water reactor (ABWR) design has been developed by General Electric Co. (G.E.) and its technical associates, Hitachi Ltd. and Toshiba Corporation, under the sponsorship of the Tokyo Electric Power Company. Key features of the ABWR include simplification, improved safety and reliability, reduced occupational exposure and radwaste, improved maneuverability, and reduced construction, fuel, and operating costs. The ABWR incorporates the best proven features from BWR designs in Japan, Europe, and the United States, combined with the broad-scope application of leading-edge technology. The first application of the ABWR design will be at TEP-CO's Kashiwazaki-Kariwa nuclear power station units 6 and 7, which are scheduled to begin commercial operation in 1996 and 1997, respectively. The ABWR is also the lead plant in the U.S. standard plant design certification program. It is scheduled to receive final approval from the U.S. Nuclear Regulatory Commission in early 1994. The development of the human system interaces (HSI) for the ABWR is an extension of the operating experience of existing man-machine interfaces currently being used at GE-designed plants

Part-load operation of the boiler feedwater pumps for the new French PWR 1400 MW nuclear plants - a challenge for the designer

International Nuclear Information System (INIS)

Martin, R.; Canavelis, R.; Guilloiseau, P.

1988-01-01

The Boiler feedwater pumps in Electricite de France Power Stations have to work reliably for all flow rates between 33% and 134% of the flow corresponding to the best efficiency point (Q BEP ). Under transient conditions associated with load changing these limits increase to 33% to 147% Q BEP . Due to the high specific power of these pumps, the operating conditions heavily influence the hydraulic and mechanical dimensioning. This paper presents some particular aspects of their design and test results obtained on a pump model as well as on a full scale prototype concerning suction performance, head capacity curve stability, pressure pulsations and structural excitations. (author)

ABWR evolution program

International Nuclear Information System (INIS)

Omoto, A.; Tanabe, A.; Moriya, K.; Dillmann, C.W.

1993-01-01

The ABWR plant is becoming a commercial reality in Japan where the first two units are being built by the Tokyo Electric Power Company. Although these units are scheduled to come on line in 1996 and '97, the Tokyo Electric Power Company, jointly with NSSS vendors (General Electric, Toshiba and Hitachi) and Japanese BWR utilities, initiated a program for a new plant design. This program is aimed at the further evolution of the ABWR to take advantage of new technological developments and to meet possible social changes in the years to come. The expected time for the first-of-a-kind plant to come on line is in the 2010's when the first generation plants in Japan may approach the time for replacement. This paper presents the ouline of this program with focus on the utility requirements and candidate technologies. (orig.)

Full MOX core design in ABWR

International Nuclear Information System (INIS)

Ihara, Toshiteru; Mochida, Takaaki; Izutsu, Sadayuki; Fujimaki, Shingo

2003-01-01

Electric Power Development Co., Ltd. (EPDC) has been investigating an ABWR plant for construction at Oma-machi in Aomori Prefecture. The reactor, termed FULL MOX-ABWR will have its reactor core eventually loaded entirely with mixed-oxide (MOX) fuel. Extended use of MOX fuel in the plant is expected to play important roles in the country's nuclear fuel recycling policy. MOX fuel bundles will initially be loaded only to less than one-third of the reactor, but will be increased to cover its entire core eventually. The number of MOX fuel bundles in the core thus varies anywhere from 0 to 264 for the initial cycle and, 0 to 872 for equilibrium cycles. The safety design of the FULL MOX-ABWR briefly stated next considers any probable MOX loading combinations out of such MOX bundle usage scheme, starting from full UO 2 to full MOX cores. (author)

Features of ABWR operator training with a full-scope simulator

International Nuclear Information System (INIS)

Kondou, Shin'ichi

1999-01-01

Many innovations have been incorporated into the Advanced BWR (ABWR) type control panels. In the BWR Operator Training Center (BTC), we started ABWR operator training using an ABWR full-scope simulator prior to the first ABWR plant's commercial operation. In consideration of the features of the ABWR type control panels, BTC has been conducting ABWR operator training focusing on the following 2 points; (1) Operator training reflecting the differences in the Human-Machine Interface (HMI). The new HMI devices which have the touch-operation function were introduced. These devices have higher operability, however, they require new operational skills. We planned the training program so that operators can fully acquire these skills. Also the compact main console and the new HMI devices made it relatively difficult for the operator crews to grasp visually what an operator was doing. We provide the training to have proper communication skills, and check trainees' operation using monitoring systems for simulator training. (2) Operator training responding to the expanded operation automation system. The scope of the automation system was expanded to reduce the operators' burden. We provide the training to improve the trainees' competence for 'operation and monitoring' suitable to both manual and automatic operational modes. (author)

The status of development activities of ABWR-2

International Nuclear Information System (INIS)

Anegawa, T.; Aoyama, M.; Fennern, L.E.; Koh, F.

2001-01-01

This paper reports the current status of development activities for ABWR-II which is the next generation design based on ABWR. Six Japanese BWR utilities and three BWR plant venders, General Electric Company, Hitachi Ltd. and Toshiba Corporation, have jointly been promoting the program. The program has been progressing for a decade. In Phase I (1991-92), future technologies were discussed and several plant concepts were studied. In Phase II (1993-95), in order to establish a reference reactor concept, key design features were selected. In Phase III (1996-2000), based on the reference reactor concept, modifications and improvements have been done in order to satisfy the design requirements well. The start of commercial operation of ABWR-II is estimated to be in the late 2010 when replacements of the first generation of operating nuclear power plants are expected to start. Larger power output and some passive safety features are adopted in order to make ABWR-II competitive in power generation cost and attain public confidence on safety. The advantages of a large fuel bundle design are also discussed here. (author)

Alarm system for ABWR main control panels

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

Alarm system for ABWR main control panels

International Nuclear Information System (INIS)

Kobayashi, Yuji; Saito, Koji

1997-01-01

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

A review of pumped energy storage schemes

International Nuclear Information System (INIS)

Unsworth, G.N.

1975-07-01

The comparative advantages and costs of pumped water storage, steam accumulators, storage of boiler feedwater , and compressed air storage in caverns are described. Boiler feedwater storage in caverns and pumped water storage are most economical. All systems are costly enough to justify developing reactors with load following capabilities. (E.C.B.)

Application of ABWR construction database to nuclear power plant project

International Nuclear Information System (INIS)

Takashima, Atsushi; Katsube, Yasuhiko

1999-01-01

Tokyo Electric Power Company (TEPCO) completed the construction of Kashiwazaki-Kariwa Nuclear Power Station Unit No. 6 and No. 7 (K-6/7) as the first advanced boiling water reactors (ABWR) in the world successfully. K-6 and K-7 started their commercial operations in November, 1996 and in July, 1997 respectively. We consider ABWR as a standard BWR in the world as well as in Japan because ABWR is highly reputed. However, because the interval of our nuclear power plant construction is going to be longer, our engineering level on plant construction will be declining. Hence it is necessary for us to maintain our engineering level. In addition to this circumstance, we are planning to wide application of separated purchase orders for further cost reduction. Also there is an expectation for our contribution to ABWR plant constructions overseas. As facing these circumstances, we have developed a construction database based on our experience for ABWR construction. As the first step of developing the database for these use, we analyzed our own activities in the previous ABWR construction. Through this analysis, we could define activity units of which the project consists. As the second step, we clarified the data which are treated in each activity unit and the interface among them. By taking these steps, we could develop our database efficiently. (author)

Further improvement of human-machine interface for ABWR main control room

International Nuclear Information System (INIS)

Makino, S.

2001-01-01

Tokyo Electric Power Company (TEPCO) has developed main control room panels based on progress in C and I technology. ABWR type main control room panels (ABWR MCR PNLs) are categorized as third generation type domestic BWR MCR, that is, they are were developed step by step based on operating experience with the first and the second generation BWR. ABWR type main control room panels were applied to Kashiwazaki-Kariwa Nuclear Power Station Units Number 6 and 7 (K-6/7) for the first time. K-6/7 are the first advanced BWR (ABWR), which started commercial operation in November 1996 and July 1997, respectively. The concept of ABWR MCR design was verified through wooden mock-up panels, start-up tests and commercial operation. Though the K-6/7 design has borne fruit, we are planning to refine and standardize the design based on the following concepts: to maintain the plant operation and monitoring style of ABWR MCR PNLs; to introduce brand-new HMI technology and devices; to incorporate operators' advice in the design. This paper outlines the features and improvements of the K6/7 MCR PNLs design. (author)

Advanced boiling water reactor (ABWR). Design, construction, operation and maintenance experience

International Nuclear Information System (INIS)

Idesawa, M.

1998-01-01

The ABWR has experienced all phases of design, construction, operation and maintenance at Kashiwazaki-Kariwa Nuclear Power Station Units No.6 and 7 and confirmed that originally intended development targets have been achieved with highly satisfactory results. This is the fruit of a project that collected wisdom from various sources under a international cooperative organization, with Tokyo Electric Power Company taking the leading role from the onset. These two units have not only demonstrated that ABWRs have superior performance as the first standard units of advanced light water reactor but also aroused a hope for the big potential advantages that ABWRs can provide us. The ABWR has already been awarded a U.S. standard license for having proved that it can comply with the requirements of international regulatory systems with an ample margin. There are also many construction programs with ABWRs progressing both domestically and abroad, suggesting that it has won recognition as an international standard plant. We will do our utmost to perfect the operation and maintenance records of Kashiwazaki-Kariwa Units No.6 and 7, which is the top runner among ABWRs, and to make known the superiority of this reactor to the world. (J.P.N.)

Cost reduction and safety design features of ABWR-II. Annex 5

International Nuclear Information System (INIS)

Koh, F.; Moriya, K.; Anegawa, T.

2002-01-01

The ABWR-II, which is aimed to be the next generation reactor following the latest BWR: Advanced Boiling Reactor (ABWR), is now under development jointly by the Japanese BWR utilities, General Electric Company, Hitachi Limited, and Toshiba Corporation. The key objectives of ABWR-II development include improvement in economics and further sophistication in safety for commercialization in the late 2010's and after. This paper summarizes the current status of ABWR-II development focusing on economics and safety. Plant power rating, fuel size, CRD rationalization and outage period are discussed from a cost reduction perspective. In terms of safety, the features such as diversification in emergency power sources and passive system application against severe accidents are being introduced. (author)

Development of automated ultrasonic device for in-service inspection of ABWR pressure vessel bottom head

International Nuclear Information System (INIS)

Kojima, Y.; Matsuyama, A.

1995-01-01

An automated device and its controller have been developed for the bottom head weld examination of pressure vessel of Advanced Boiling Water Reactor (ABWR). The internal pump casings and the housings of control rod prevent a conventional ultrasonic device from scanning the required inspection zone. With this reason, it is required to develop a new device to examine the bottom head area of ABWR. The developed device is characterized by the following features. (1) Composed of a mother vehicle and a compact inspection vehicle. They are connected only by an electric wire without using the conventional arm mechanism. (2) The mother vehicle travels on a track and lift up the inspection vehicle to the vessel. (3) The mother vehicle can automatically attach the inspection vehicle to the bottom head, and detach the inspection vehicle from it. (4) Collision avoidance control function with a touch sensor is installed at the front of the inspection vehicle. The device was successfully demonstrated using a mock-up of reactor pressure vessel

Design of full MOX core in ABWR

International Nuclear Information System (INIS)

Kinoshita, Y.; Hirose, T.; Sasagawa, M.; Sakuma, T

1999-01-01

A Full MOX-ABWR, loaded with mixed-oxide (MOX) fuels of up to 100% of the core, is planned. Increased MOX fuel utilization will result in greater savings of uranium. Studies on the fuel rod thermal-mechanical design, the core design and the safety evaluation have been made, and the results are summarized in this paper. To sum it all up, the safety of the Full MOX-ABWR has been confirmed through design evaluations adequately considering the MOX fuel and core characteristics. (author)

Loss of feedwater heater analysis for the South Texas Project

International Nuclear Information System (INIS)

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

1987-01-01

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

Loss-of-feedwater transients in PWRs

International Nuclear Information System (INIS)

Burns, R.D. III.

1980-01-01

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

Safety system consideration of a supercritical-water cooled fast reactor with simplified PSA

International Nuclear Information System (INIS)

Lee, J.H.; Oka, Y.; Koshizuka, S.

1999-01-01

The probabilistic safety of the supercritical-water cooled fast reactor (SCFR) is evaluated with the simplified probabilistic safety assessment (PSA) methodology. SCFR has a once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure. There are no recirculation loops in the once-through direct cycle system, which is the most important difference from the current light water reactor (LWR). The main objective of the present study is to assess the effect of this difference on the safety in the stage of conceptual design study. A safety system configuration similar to the advanced boiling water reactor (ABWR) is employed. At loss of flow events, no natural recirculation occurs. Thus, emergency core flow should be quickly supplied before the completion of the feedwater pump coastdown at a loss of flow accident. The motor-driven high pressure coolant injection (MD-HPCI) system cannot be used for the quick core cooling due to the delay of the emergency diesel generator (D/G) start-up. Accordingly, an MD-HPCI system in an ABWR is substituted by a turbine-driven (TD-) HPCI system for the SCFR. The calculated core damage frequency (CDF) is a little higher than that of the Japanese ABWR and a little lower than that of the Japanese BWR when Japanese data are employed for initiating event frequencies. Four alternatives to the safety system configurations are also examined as a sensitivity analysis. This shows that the balance of the safety systems designed here is adequate. Consequently, though the SCFR has a once-through coolant system, the CDF is not high due to the diversity of feedwater systems as the direct cycle characteristics

Feedwater control system

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

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

Man-machine interface systems and operator training program for ABWR in Japan

International Nuclear Information System (INIS)

Kunito, Susumu

2004-01-01

The Tokyo Electric Power Company (TEPCO) has developed a new Main Control Room design for the Advanced Boiling Water Reactor (ABWR) to improve man-machine interface. New configuration of panels and enhanced automation are some of the features of the ABWR type Main Control Room design. Various technologies such as Cathode Ray Tubes (CRTs) and Flat Displays (FDs) with touch-sensitive operations are contributed to the development of the ABWR type control room design. This design will be first applied to Kashiwazaki-Kariwa Nuclear Power Station unit 6 (K-6). To train the operators sufficiently, TEPCO reviewed the operator training program. Compared with the conventional training, new training menu will be added and the training of ABWR operators will be started 6 months earlier. An ABWR simulator is under construction and training using this simulator is scheduled to be started in August 1994, which is 18 months before fuel loading of K-6. We are reviewing malfunction modes on the simulator. (author)

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

International Nuclear Information System (INIS)

Estrada, Herb; Hauser, Ernie

2007-01-01

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

Feedwater control method and device therefor

International Nuclear Information System (INIS)

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

1997-01-01

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

U.S. licensing process and ABWR certification

International Nuclear Information System (INIS)

Quirk, J.F.; Williams, W.A.

1996-01-01

Part 50 of Title 10 of the Code of Federal Regulation (CFR) establishes a two-step licensing process by which the U.S. Nuclear Regulatory Committee (NRC) authorizes nuclear reactor plant construction through issuance of a construction permit and authorizes operation by issuance of an operating license. At each stage, the NRC Staff conducts technical reviews and there is potential for public hearings. In 1989, the NRC issued a new, simplified licensing process: Part 52. The purpose of the Part 52 licensing process is to provide a regulatory framework that brings about earlier resolution of licensing issues. Because issues are not resolved early in the Part 50 licensing process, approval of an operating license is not assured until after a significant investment has been made in the plant. Part 52 increases the stability and certainty of the licensing process by providing for the early resolution of safety and environmental issues. The Part 52 licensing process features (1) early site permits, (2) design certification, and (3) combined construction permit and operating licenses. As part of the U.S. Advanced Light Water Reactor (ALWR) Program to revitalize the nuclear option through the integration of government/utility/industry efforts, GE undertook the role of applying for certification for its latest product line, the Advanced Boiling Water Reactor (ABWR), under the U.S. ABWR certification program. The ABWR design is an essentially complete plant. Initial application for design certification was in 1987 under Part 50. GE reapplied in late 1991 under the newly promulgated Part 52. Following seven years of intensive interactions with the NRC and ACRS, GE was awarded the first Final Design Approval (FDA) under Part 52. The Commission initiated rulemaking by publishing the proposed ABWR Certification Rule in the Federal Register in early 1995. Certification is anticipated mid-1996. (J.P.N.)

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

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

What kind of nuclear energy should be offered? Ge's ABWR nuclear plant

International Nuclear Information System (INIS)

Hucik, S.; Redding, J.

2000-01-01

GE is proud of the ABWR design and our accomplishments in providing this safe, cost effective means of generating electricity to our customers around the world. The ABWR defines the new generation of advanced nuclear plant designs that have achieved higher levels of safety, compete economically with other forms of power generation, and can contribute importantly to the goals of sustainable development. The GE-led team has amassed significant experience in licensing, designing, and constructing the ABWR and has established a strong track record of success in doing so. We believe the quality of the ABWR design and our experience will have a strong appeal to utility business managers and the financial community. Finally we believe in nuclear energy and it's ability to improve the quality of people's lives throughout the world. Nuclear energy can and should play an important role in meeting the worlds' growing needs for electricity while at the same time preserving our environment for future generations. (authors)

Evaluation of damages of airplane crash in European Advanced Boiling Water Reactor (EU-ABWR)

International Nuclear Information System (INIS)

Kamei, Kazuhiro; Tanoue, Tetsuharu; Kataoka, Kazuyoshi; Jimbo, Masakazu

2011-01-01

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash (APC), severe accident mitigation systems, N+2 principle in safety systems and a large output of 1600 MWe. Thanks to above mentioned features, EU-ABWR's design objectives and principles are consistent with safety requirements in an European market. In this paper, evaluation of damages induced by APC has been summarized. (author)

Design of a reactor core in the Oma Full MOX-ABWR

International Nuclear Information System (INIS)

Hama, Teruo

1999-01-01

The Electric Power Development Co., Ltd. has progressed a construction plan on an improved boiling-water reactor aiming at loading of MOX fuel in all reactor cores (full MOX-ABWR) at Oma-cho, Aomori prefecture, which is a last stage on application of approval on establishment at present. Here were described on outlines of reactor core in the full MOX-ABWR and its safety evaluation. For the full MOX-ABWR loading MOX fuel assembly into all reactor core, thermal and mechanical design analysis of fuel bars and core design analysis were conducted. As a result, it was confirmed that judgement standards in mixed core of MOX fuel and uranium fuel were also applicable as well as that in uranium fuel. (G.K.)

ABWR-II Core Design with Spectral Shift Rods for Operation with All Control Rods Withdrawn

International Nuclear Information System (INIS)

Moriwaki, Masanao; Aoyama, Motoo; Anegawa, Takafumi; Okada, Hiroyuki; Sakurada, Koichi; Tanabe, Akira

2004-01-01

An innovative reactor core concept applying spectral shift rods (SSRs) is proposed to improve the plant economy and the operability of the 1700-MW(electric) Advanced Boiling Water Reactor II (ABWR-II). The SSR is a new type of water rod in which a water level is naturally developed during operation and changed according to the coolant flow rate through the channel. By taking advantage of the large size of the ABWR-II bundle, the enhanced spectral shift operation by eight SSRs allows operation of the ABWR-II with all control rods withdrawn. In addition, the uranium-saving factor of 6 to 7% relative to the reference ABWR-II core with conventional water rods can be expected due to the greater effect of spectral shift. The combination of these advantages means the ABWR-II with SSRs should be an attractive alternative for the next-generation nuclear reactor

ABWR1. A Generation III.7 reactor after the Fukushima Daiichi accident

International Nuclear Information System (INIS)

Sato, Takashi; Matsumoto, Keiji; Kurosaki, Toshikazu; Taguchi, Keisuke

2015-01-01

PWR resulting in a smaller R/B. Each active safety division includes only one ECCS pump and one EDG. Therefore, a single failure of the EDG never causes multiple failures of ECCS pumps in a safety division. The iB1350 is based on the proven ABWR technology and ready for construction. No new technology is incorporated but design concept and philosophy are initiative and innovative. (author)

Simulation technology to develop the automatic control rod operation system in ABWR

Energy Technology Data Exchange (ETDEWEB)

Yoshihiko, Ishii; Atsushi, Fushimi; Naoyuki, Ishida [Hitachi Ltd. Power and Industrial Systems R and D Laboratory, Ibaraki (Japan); Hitoshi, Ochi; Masakazu, Yuzuki [Hitachi Ltd. Power Systems, Hitachi Works, Ibaraki (Japan); Hideki, Hanami [Hitachi Ltd. Information and Control System Div., Ibaraki (Japan)

2007-07-01

To design and verify the automatic power regulator system (APR) for an advanced boiling water reactor (ABWR), Hitachi has developed a plant simulation code system. The 3-dimensional kinetic analysis code STAND, based on a polynomial nodal expansion method, was effective in developing an APR automatic control rod operation algorithm in the critical mode. The point kinetic plant simulator AUTSIMU is an off-line and real time online plant simulation code that can simulate ABWR start-up and shutdown processes from a sub-critical condition with a water temperature of 20 degrees C to the rated operation power level at about 7 MPa. APR operation algorithms were designed using the off-line mode. In this mode, less than one minute in real time was needed to simulate a ten-hour transient with an ordinary personal computer. In the online mode, APR hardware was efficiently verified in a factory test before shipping. The codes were verified with ABWR start-up experimental data. It was confirmed that the simulators were effective in simulating ABWR plant start-up behavior. (authors)

Reactor feedwater device

International Nuclear Information System (INIS)

Igarashi, Noboru.

1986-01-01

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

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

International Nuclear Information System (INIS)

Bae, Yeon Kyoung; Lee, Eun Chan

2012-01-01

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

Speed control device for coolant recycling pump

International Nuclear Information System (INIS)

Kageyama, Takao.

1992-01-01

The present invention intends to increase a margin relative of the oscillations of neutron fluxes when the temperature of feedwater is lowered in a compulsory recycling type BWR reactor. That is, when the operation point represented by a reactor thermal power and a reactor core inlet flow rate is in a state approximate to an oscillation limit of the reactor power, the device of the present invention controls the recycling pump speed in the increasing direction depending on the lowering range of the feedwater temperature from a stationary state. With such a constitution, even if the reactor power is in the operation region near the oscillation limit in the BWR type reactor and a feedwater heating loss is caused, the speed of the coolant recycling pump is increased by 10% at the maximum depending on the extent of the reduction of the feedwater temperature, so that the oscillation of the reactor power can be prevented from lasting for a long period of time even if a reactivity external disturbance should occur in the reactor. (I.S.)

Safety design philosophy of the ABWR for the next generation LWRs

International Nuclear Information System (INIS)

Sato, Takashi; Akinaga, Makoto; Kojima, Yoshihiro

2009-01-01

The paper presents safety design philosophy of the advanced boiling water reactor (ABWR) to be reflected in developing the next generation light water reactors (LWRs). The basic policy of the ABWR safety design was to improve safety and reduce cost simultaneously by reflecting lessons learned of precursors, incidents and accidents that were beyond the design basis such as the Three Mile Island Unit 2 (TMI 2) accident. The ABWR is a fully active safety plant. The ABWR enhanced redundancy and diversity of active safety systems using probabilistic safety assessment (PSA) insights. It adopted a complete three division active emergency core cooling system (ECCS) and attained a very low core damage frequency (CDF) value of less than 10 -7 /ry for internal events. Only very small residual risks, if any, rather exist in external events such as an extremely large earthquake beyond the design basis. This is because external events can constitute a common cause that disables all the redundant active safety systems. Therefore, it is useless to add one more ECCS train and make a four division active ECCS for external events. Nowadays, however, fully passive safety LWRs are already established. Incorporating some of these passive safety systems we can also establish the next generation LWRs that are truly strong against external events. We can establish a plant that can survive a giant earthquake at least three days without AC power source, SA proof safety design that enables no containment failure and no evacuation to eliminate the residual risks. The same basic policy as the ABWR to improve safety and reduce cost simultaneously is again effective for the next generation LWRs. (author)

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Feedwater control system in nuclear power plants

International Nuclear Information System (INIS)

Masuyama, Hideo.

1981-01-01

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

Monitor for reactor feedwater systems

International Nuclear Information System (INIS)

Takizawa, Yoji; Tomizawa, Teruaki

1983-01-01

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

GE-Hitachi ABWR Design for 60 Years of Operation

International Nuclear Information System (INIS)

Hoang Xuan Hoa

2011-01-01

The GE Hitachi ABWR design benefits from the experiences of the prior generation BWRs in operation since the 1970s, and the knowledge learned from years of technical research on the aging effects on reactor systems, structures and components. The impact of exposure to neutrons over time to the reactor pressure vessel and internal components was evaluated to confirm their structural integrity. Other time-based plant analyses, such as fatigue vibration and corrosion effect, have also been performed on the ABWR to support a 60-year plant life assumption. Finally, operating experience and aging lessons learned have been used to verify the maintenance plans for plant aging management are acceptable to support operation up to 60 years. (author)

Auxiliary feedwater system aging study

International Nuclear Information System (INIS)

Kueck, J.D.

1992-01-01

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

Development of ABWR-2 and its safety design

International Nuclear Information System (INIS)

Takafumi, Anegawa; Kenji, Tateiwa

2002-01-01

This paper reports the current status of development project on ABWR-II, a next generation reactor design based on ABWR, and its safety design. This project was initiated over a decade ago and has completed three phases to date. In Phase I (1991-92), basic design requirements were discussed and several plant concepts were studied. In Phase II (1993-95), key design features were selected in order to establish a reference reactor concept. In Phase III (1996-2000), based on the reference reactor concept, modifications and improvements were made to fulfill the design requirements. By adopting large electric output (1 700 MW), large fuel bundle, modified ECCS, and passive heat removal systems, among other design features, we achieved a design concept capable of increasing both economic competitiveness and safety performance. Main focus of this paper will be on the safety design, safety performance, and further research needs related to safety. (authors)

Transient analysis of ABWR reactor using a best estimate code

International Nuclear Information System (INIS)

Mizokami, S.; Kitamura, H.; Mototani, A.; Ono, H.

2004-01-01

Since the recirculation pumps are mounted internally within the ABWR, core flow will decrease rapidly in the event of a loss of their driving force. A rapid reduction in core flow may cause the onset of boiling transition (BT). Therefore, in order to prevent the onset of BT, a motor-generator (MG) set is added to the power supply system of the reactor internal pump (RIP). Recent studies, however, have shown that dryout within a fuel assembly over a short time period will result in only a small rise in fuel cladding temperature and thus does not pose a threat to fuel integrity. In response to this finding, the standards committee of the Atomic Energy Society of Japan (AESJ) has proposed a post-BT standard which incorporates a cladding temperature criterion. If it is assumed that the MG-set is not added to the RIP power supply system, the result of the safety analysis shows the onset of BT with a subsequent rise in fuel cladding temperature. Although BT occurs under the conservative assumptions of this safety analysis, a possibility exists that BT will not occur under actual operating conditions. The best estimate code TRACG was used to show that BT does not occur and that fuel integrity can be sufficiently maintained under actual conditions. (author)

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

International Nuclear Information System (INIS)

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

1980-01-01

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

Reliability analysis of 2 types of auxiliary feedwater system for PWR

International Nuclear Information System (INIS)

Ekariansyah, Andi Sofrany

2002-01-01

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

A Smart Soft Sensor Predicting Feedwater Flow Rate

International Nuclear Information System (INIS)

Yang, Heon Young; Na, Man Gyun

2009-01-01

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

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.

Validations of BWR nuclear design code using ABWR MOX numerical benchmark problems

International Nuclear Information System (INIS)

Takano, Shou; Sasagawa, Masaru; Yamana, Teppei; Ikehara, Tadashi; Yanagisawa, Naoki

2017-01-01

BWR core design code package (the HINES assembly code and the PANACH core simulator), being used for full MOX-ABWR core design, has been benchmarked against the high-fidelity numerical solutions as references, for the purpose of validating its capability of predicting the BWR core design parameters systematically from UO 2 to 100% MOX cores. The reference solutions were created by whole core critical calculations using MCNPs with the precisely modeled ABWR cores both in hot and cold conditions at BOC and EOC of the equilibrium cycle. A Doppler-Broadening Rejection Correction (DCRB) implemented MCNP5-1.4 with ENDF/B-VII.0 was mainly used to evaluate the core design parameters, except for effective delayed neutron fraction (β eff ) and prompt neutron lifetime (l) with MCNP6.1. The discrepancies in the results between the design codes HINES-PANACH and MCNPs for the core design parameters such as the bundle powers, hot pin powers, control rod worth, boron worth, void reactivity, Doppler reactivity, β eff and l, are almost within target accuracy, leading to the conclusion that HINES-PANACH has sufficient fidelity for application to full MOX-ABWR core design. (author)

ABWR certification work brings US licensing stability nearer

International Nuclear Information System (INIS)

Wilkins, D.R.; Quirk, J.F.

1991-01-01

The Advanced Boiling Water Reactor (ABWR) is now approaching Final Design Approval by the US Nuclear Regulatory Commission (NRC) and will then proceed on to the certification phase of the NRC's new standard plant licensing process. Successful completion of this will usher in a new era of standardization and reactor licensing stability in the US. (author)

Analysis of Total Loss of Feedwater for APR1400 using SPACE

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Commercial operation and outage experience of ABWR at Kashiwazaki-Kariwa units Nos. 6 and 7

International Nuclear Information System (INIS)

Anahara, N.; Yamada, M.; Kataoka, H.

2000-01-01

drums and that of unit No. 3. Overhaul and inspection of new designs components, such as reactor internal pumps (RIPs), advanced type control rod drive mechanism (FMCRD) and large-capacity and high-efficiency turbine system, were conducted as planned, verifying their integrity after the commercial operation period. Operability during annual outage through ABWR-type main control room panels, which have a very new man-machine interface with touch-screen CRTs and flat displays, was also confirmed. In this paper, the commercial operation and annual outage experience of Kashiwazaki-Kariwa Nuclear Power Station units Nos. 6 and 7 are reported. (author)

ESBWR power maneuvering via feedwater temperature control

International Nuclear Information System (INIS)

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

2008-01-01

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

Overview of internal fire hazards aspects of ABWR design for United Kingdom

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Kazuhiro; Kawai, Hiroki [Hitachi-GE Nuclear Energy, Ltd., Ibaraki (Japan)

2015-12-15

The ABWR (Advanced Boiling Water Reactor) is a generation III+ reactor, the most modern operational generation of nuclear power plants. The UK ABWR design is proposed for development and construction in the United Kingdom (UK), and under review by the Office for Nuclear Regulation (ONR) through Generic Design Assessment (GDA). The UK ABWR design has mainly two types of the safety system: ''preventing'' and ''mitigating'' a fault and their consequences. The prevention of internal hazards starts with design processes and procedures. These processes lead to limiting the sources of potential hazards. The mitigative safety systems are required to ensure the fundamental safety functions (FSFs): control of reactivity, Fuel cooling, long term heat removal, confinement/containment of radioactive materials, and others. Implementation of the safety philosophy is based upon redundant and diverse safety systems that deliver the FSFs. Three mechanical divisions are provided, each of which contains redundant systems, structures, and components (SSCs) capable of carrying out all the FSFs. The safety divisions are separated by robust barriers which act to contain a hazard in an affected division and prevent the spread of the hazard to a different division. The deterministic assessments and the hazard schedule argue that the rooms containing SSCs providing the FSFs are located in different fire safety divisions. The approach to maintaining the FSFs during and after internal fires is to ensure fires do not spread beyond that division to affect redundant equipment in other divisions. During the GDA process, it is demonstrated that generally barrier compartmentation (the divisional barrier walls, ceilings and floors) is sufficient to contain the postulated fires. The UK ABWR design has sufficient capability of withstanding the postulated internal fire hazard to achieve the FSFs. Further development is being undertaken with feedback in the GDA

Feedwater heater

International Nuclear Information System (INIS)

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

1991-01-01

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

Lead corrosion and transport in simulated secondary feedwater

International Nuclear Information System (INIS)

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

1998-01-01

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

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)

Resolution of concerns in auxiliary feedwater piping

International Nuclear Information System (INIS)

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

1994-01-01

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

Effects of lower plenum flow structure on core inlet flow of ABWR

International Nuclear Information System (INIS)

Watanabe, Shun; Abe, Yutaka; Kaneko, Akiko; Watanabe, Fumitoshi; Tezuka, Kenichi

2010-01-01

The evaluation of coolant flow structure at a lower plenum of an advanced boiling water reactor (ABWR) in which there are many structures is very important in order to improve generating power. Although the simulation results by CFD (Computational Fluid Dynamics) codes can predict such complicated flow in the lower plenum, it is required to establish the database of flow structure in lower plenum of ABWR experimentally for the benchmark of the CFD codes. In the model of the lower plenum, we measured velocity profiles with LDV and PIV. And differential pressure of constructed model is measured with differential pressure instrument. It was identified that the velocity and differential pressure profiles also showed the tendency to be flat in the core inlet. Moreover, vortexes were observed around side entry orifice by PIV measurement. (author)

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)

CORE DESIGNS OF ABWR FOR PROPOSED OF THE FIRST NUCLEAR POWER PLANT IN INDONESIA

Directory of Open Access Journals (Sweden)

Yohannes Sardjono

2015-04-01

Full Text Available Indonesia as an archipelago has been experiencing high growth industry and energy demand due to high population growth, dynamic economic activities. The total population is around 230 million people and 75 % to the total population is living in Java. The introduction of Nuclear Power Plant on Java Bali electricity grid will be possible in 2022 for 2 GWe, using proven technology reactor like ABWR or others light water reactor with nominal power 1000 MWe. In this case, the rated thermal power for the equilibrium cycles is 3926 MWt, the cycle length is 18 month and overall capacity factor is 87 %. The designs were performed for an 872-fuel bundles ABWR core using GE-11 fuel type in an 9×9 fuel rod arrays with 2 Large Central Water Rods (LCWR. The calculations were divided into two steps; the first is to generate bundle library and the other is to make the thermal and reactivity limits satisfied for the core designs. Toshiba General Electric Bundle lattice Analysis (TGBLA and PANACEA computer codes were used as designs tools. TGBLA is a General Electric proprietary computer code which is used to generate bundle lattice library for fuel designs. PANACEA is General Electric proprietary computer code which is used as thermal hydraulic and neutronic coupled BWR core simulator. This result of core designs describes reactivity and thermal margins i.e.; Maximum Linear Heat Generation rate (MLHGR is lower than 14.4 kW/ft, Minimum Critical Power Ratio (MCPR is upper than 1.25, Hot Excess Reactivity (HOTXS is upper than 1 %Dk at BOC and 0.8 %Dk at 200 MWD/ST and Cold Shutdown Margin Reactivity (CSDM is upper than 1 %Dk. It is concluded that the equilibrium core design using GE-11 fuel bundle type satisfies the core design objectives for the proposed of the firs Indonesia ABWR Nuclear Power Plant. Keywords: The first NPP in Indonesia, ABWR-1000 MWe, and core designs.   Indonesia adalah sebagai negara kepulauan yang laju pertumbuhan industri, energi, penduduk

Software failure events derivation and analysis by frame-based technique

International Nuclear Information System (INIS)

Huang, H.-W.; Shih, C.; Yih, Swu; Chen, M.-H.

2007-01-01

A frame-based technique, including physical frame, logical frame, and cognitive frame, was adopted to perform digital I and C failure events derivation and analysis for generic ABWR. The physical frame was structured with a modified PCTran-ABWR plant simulation code, which was extended and enhanced on the feedwater system, recirculation system, and steam line system. The logical model is structured with MATLAB, which was incorporated into PCTran-ABWR to improve the pressure control system, feedwater control system, recirculation control system, and automated power regulation control system. As a result, the software failure of these digital control systems can be properly simulated and analyzed. The cognitive frame was simulated by the operator awareness status in the scenarios. Moreover, via an internal characteristics tuning technique, the modified PCTran-ABWR can precisely reflect the characteristics of the power-core flow. Hence, in addition to the transient plots, the analysis results can then be demonstrated on the power-core flow map. A number of postulated I and C system software failure events were derived to achieve the dynamic analyses. The basis for event derivation includes the published classification for software anomalies, the digital I and C design data for ABWR, chapter 15 accident analysis of generic SAR, and the reported NPP I and C software failure events. The case study of this research includes: (1) the software CMF analysis for the major digital control systems; and (2) postulated ABWR digital I and C software failure events derivation from the actual happening of non-ABWR digital I and C software failure events, which were reported to LER of USNRC or IRS of IAEA. These events were analyzed by PCTran-ABWR. Conflicts among plant status, computer status, and human cognitive status are successfully identified. The operator might not easily recognize the abnormal condition, because the computer status seems to progress normally. However, a well

Development and application of a dual RELAP5-3D-based engineering simulator for ABWR

International Nuclear Information System (INIS)

Yang, C.-Y.; Liang, Thomas K.S.; Pei, B.S.; Shih, C.K.; Chiang, S.C.; Wang, L.C.

2009-01-01

For any innovated plant design, the designed paper plant can be converted into a computer as a digital plant with advanced simulation techniques before being constructed into a real plant. A digital plant, namely engineering simulator, can be applied for: (1) verification of system design and system integration, (2) power test simulation, (3) plant transient and accident analyses, (4) plant abnormal and emergency procedure development and verification, (5) design change verification and analysis, etc. An advanced engineering simulator was successfully developed for the LungMen advanced boiling water reactor (ABWR) plant to support various applications before and after commercial operation. This plant specific engineering simulator was developed based on two separate RELAP5-3D modules synchronized on a commercial simulation platform, namely 3-Key Master. On this advanced LungMen plant simulation (ALPS) platform, major plant dynamics were simulated by two separate RELAP5-3D modules, one for reactor system modeling and the other for balance of plant (BOP) system modeling. Moreover, major control systems as well as emergency core cooling system (ECCS) were all simulated in great detail with built-in tasks of this commercial simulation platform. Different from real time calculation on training simulator, precision of engineering calculation is intentionally kept by synchronizing modules based on the most time-consuming one. During synchronization, each module will check its' own converge criteria in each small time advancement. This plant specific advanced ABWR engineering simulator has been successfully applied on: (1) licensing blowdown analysis of feed water line break (FWLB) for containment design; (2) phenomena investigation of low-pressure ECC injection bypass during FWLB; (3) analysis of FW pump performance during power ascending; (4) verification of plant vendor's pre-test calculations of each start-up test.

Feedwater temperature control methods and systems

Science.gov (United States)

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

2014-04-22

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

Aiding operator performance at low power feedwater control

International Nuclear Information System (INIS)

Woods, D.D.

1986-01-01

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

The ultimate response guideline simulation and analysis by using (TRACE) for Lungmen ABWR nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Lin, Hao-Tzu [Atomic Energy Council, Inst. of Nuclear Energy Research, Taoyuan City, Taiwan (China); Yang, Shu-Ming; Chen, Shao-Wen; Wang, Jong-Rong; Shih, Chunkuan [National Tsing Hua Univ., Inst. of Nuclear Engineering and Science, HsinChu, Taiwan (China)

2015-07-15

In this research, the TRACE/SNAP model of Lungmen ABWR nuclear power plant (NPP) has been established for the simulation and analysis of ultimate response guideline (URG). The main actions of URG are depressurization and low pressure water injection of reactor and containment venting. This research focuses to assess the URG utility of Lungmen NPP under Fukushima-like conditions. This study consists of three steps. The first step is the establishment of Lungmen NPP TRACE/SNAP model. In order to evaluate the system response of TRACE/SNAP model, FSAR data (MSIV closure and loss of feedwater flow transient) were used to compare with the results of TRACE. The second step is the URG simulation and analysis under Fukushima-like conditions by using Lungmen NPP TRACE/SNAP model. In this step, the no URG case was also performed in order to evaluate the URG effectiveness of Lungmen NPP. In order to confirm the mechanical property and integrity of fuel rods, the final step is FRAPTRAN analysis. According to TRACE analysis results, the URG can keep the peak cladding temperature (PCT) below the criteria 1088.7 K under Fukushima-like conditions which indicates that Lungmen NPP can be controlled in a safe situation. Nevertheless, if Lungmen NPP does not perform the URG under Fukushima-like conditions, the water level may drop lower than TAF after 1100 s which means a safety issue about the fuel rods may be generated. The analysis results of FRAPTRAN also indicate the integrity of fuel rods cannot be kept under the above conditions.

Experimental study on RCCV of ABWR plant-Part 1

International Nuclear Information System (INIS)

Saita, H.; Kikuchi, R.; Muramatsu, Y.; Hiramoto, M.; Oyamada, O.; Furukawa, H.; Sasagawa, K.; Ohmori, N.; Suzuki, S.; Suoita, M.; Kobayashi, I.; Yamaguchi, I.

1989-01-01

Steel containment vessels have generally been used for BWR (Boiling Water Reactor)-type nuclear power plants in Japan. However, in accordance with the development of the advanced-type BWR (ABWR), the introduction of a reinforced concrete containment vessel (RCCV) has been proposed because it enables greater freedom in flexibility of configuration. The study presented here accumulates basic data on RCCV's and verifies the integrity and ultimate strength of a trial-designed RCCV and the validity of its design method

Secondary coolant circuit operation tests: steam generator feedwater supply

International Nuclear Information System (INIS)

Beroux, M.

1985-01-01

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

Dependence of steam generator vibrations on feedwater pressure

International Nuclear Information System (INIS)

Sadilek, J.

1989-01-01

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

Study of plutonium disposition using the GE Advanced Boiling Water Reactor (ABWR)

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-04-30

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the U.S. to disposition 50 to 100 metric tons of excess of plutonium in parallel with a similar program in Russia. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing long-term diversion resistance to this material. The NAS study {open_quotes}Management and Disposition of Excess Weapons Plutonium{close_quotes} identified light water reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a U.S. disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a 1350 MWe GE Advanced Boiling Water Reactor (ABWR) is utilized to convert the plutonium to spent fuel. The ABWR represents the integration of over 30 years of experience gained worldwide in the design, construction and operation of BWRs. It incorporates advanced features to enhance reliability and safety, minimize waste and reduce worker exposure. For example, the core is never uncovered nor is any operator action required for 72 hours after any design basis accident. Phase 1 of this study was documented in a GE report dated May 13, 1993. DOE`s Phase 1 evaluations cited the ABWR as a proven technical approach for the disposition of plutonium. This Phase 2 study addresses specific areas which the DOE authorized as appropriate for more in-depth evaluations. A separate report addresses the findings relative to the use of existing BWRs to achieve the same goal.

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

Science.gov (United States)

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

2010-02-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

Sanchez B, A.

2003-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Seismic margin analysis for Kashiwazaki Kariwa ABWR plant considering the Niigataken Chuetsu-oki earthquake

International Nuclear Information System (INIS)

Matsuo, Toshihiro; Nagasawa, Kazuyuki; Kawamura, Shinichi; Ueki, Takashi; Higuchi, Tomokazu; Sakaki, Isao

2009-01-01

Seismic Margin Analysis (SMA) study was conducted for Kashiwazaki Kariwa (KK) ABWR representative plant (unit 6). Considering that the installation behaved in a safe manner during and after the Niigataken Chuetsu-oki (NCO) Earthquake which significantly exceeded the level of the seismic input taken into account in the design of the plant, the study to find out how much margin the ABWR plant had toward the same seismic motion was conducted. In this study fragility analyses were conducted for SSCs that were included in the accident sequences and that were considered to have relatively small margin taking EPRI margin analysis method into consideration. In order to calculate plant level seismic margin Min-Max method was adopted. As the result of this study, the plant level High Confidence Low Probability of Failure (HCLPF) acceleration for unit 6 was calculated more than tripled NCO earthquake motion. (author)

Probabilistic based design rules for intersystem LOCAS in ABWR piping

International Nuclear Information System (INIS)

Ware, A.G.; Wesley, D.A.

1993-01-01

A methodology has been developed for probability-based standards for low-pressure piping systems that are attached to the reactor coolant loops of advanced light water reactors (ALWRs) which could experience reactor coolant loop temperatures and pressures because of multiple isolation valve failures. This accident condition is called an intersystem loss-of-coolant accident (ISLOCA). The methodology was applied to various sizes of carbon and stainless steel piping designed to advanced boiling water reactor (ABWR) temperatures and pressures

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Feedwater heater performance evaluation using the heat exchanger workstation

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Water feeding/condensating device and operation method in nuclear power plant

International Nuclear Information System (INIS)

Shibayama, Takashi.

1989-01-01

The present invention overcomes a problem in reactor water level control occurring upon operation of a water feeding/condensating system in a nuclear power plant. That is, the water feed system to a nuclear reactor is constituted with parallel circuit comprising a reactor feedwater pump driven by a steam turbine and a serial circuit composed of a reactor feedwater pump driven by an electrical motor and a pump adjusting valve for controlling the amount of feedwater at the exit of the motor driven feedwater pump. Further, a reactor feedwater control valve having a function of controlling the feedwater to the reactor is disposed to the bypass pipeway for bypassing the parallel circuit of feedwater pumps. In this constitution, water can be fed to the nuclear reactor by way of the reactor feedwater pump bypass control valve upon starting and stopping of a nuclear feedwater pump driven by electric motor upon starting and shutdown of the nuclear reactor. Accordingly, stable water level control can be conducted for the reactor core with no effect of rapid pressure fluctuation due to the starting and the stopping of the reactor feedwater pump driven by electric motor. (I.S.)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Research on instability design method without occurring boiling transition for hyper ABWR plants of extended core power density

International Nuclear Information System (INIS)

Okamoto, T.; Hotta, A.; Ama, T.

2008-01-01

The hyper ABWR (Advanced Boiling Water Reactor) project aims to develop an advanced BWR concept that is competitive in the global market with both highly economic and safety features. Expecting plant construction within the coming ten years, a research program for substantiating the basic design of a high core power density ABWR was conducted. By inheriting the conventional ABWR design, it is possible to reduce construction costs. In order to achieve the rated core power of over 1650MWe which is almost equivalent to that of the EPR (European Pressurized Water Reactor), the core power density of ABWR will be up-rated by at least 25%. Three key subjects linked to this target were recognized. They are, (1) fuel design applicable to the high power density core, (2) improvement of the evaluation method for the coupled neutronic and thermal-hydraulic instability under a wider power-flow operating range, and (3) improvement of the steam separator performance under high quality conditions. In this paper, the second subject has been focused on. In the second subject, the uncertainty approach was introduced in the instability analysis where the best-estimate plant simulator was combined with a direct prediction of boiling transition by the sub-channel code. By employing the CSAU like method, a safety evaluation system that enables to include influences of uncertainties has been developed. Based on the correlation between the time margin for reaching the boiling transition under power oscillations and the decay ratio in the power-flow operation map, an automatic power oscillation suppressing system was designed. The set-point for activating suppression mechanisms (i.e. scram or SRI) could be determined based on this correlation. It was proposed that the present conservative acceptance criterion of the deterministic decay ratio can be replaced with a more rational one of the time margin with including uncertainties. (author)

Manual for investigation and correction of feedwater heater failures

International Nuclear Information System (INIS)

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

1993-01-01

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

Pressurized water-reactor feedwater piping response to water hammer

International Nuclear Information System (INIS)

Arthur, D.

1978-03-01

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

Operating experiences and degradation detection for auxiliary feedwater systems

International Nuclear Information System (INIS)

Casada, D.; Farmer, W.S.

1992-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

ABWR (K-6/7) construction experience (computer-based safety system)

International Nuclear Information System (INIS)

Yokomura, T.

1998-01-01

TEPCO applied a digital safety system to Kashiwazaki-Kariwa Nuclear Power Station Unit Nos. 6 and 7, the world's first ABWR plant. Although this was the first time to apply a digital safety logic system in Japan, we were able to complete construction of K-6/7 very successfully and without any delay. TEPCO took a approach of developing a substantial amount of experience in digital non- safety systems before undertaking the design of the safety protection system. This paper describes the history, techniques and experience behind achieving a highly reliable digital safety system. (author)

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

International Nuclear Information System (INIS)

Fuller, R.; Harrell, J.

1996-01-01

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

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.

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

International Nuclear Information System (INIS)

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

1986-02-01

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

Feedwater connection repair and modification at GKN

Energy Technology Data Exchange (ETDEWEB)

Witteman, C; Klees, J E

1985-03-01

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

Feedwater connection repair and modification at GKN

International Nuclear Information System (INIS)

Witteman, C.; Klees, J.E.

1985-01-01

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

BWR type reactors

International Nuclear Information System (INIS)

Kikuchi, Osamu.

1985-01-01

Purpose: To prevent cavitations in a recycling pump, as well as improve the safety and a reliability of a pressure vessel. Constitution: A feedwater pipeway is connected to the route between the pressure vessel and the recycling pipe and feedwater from the feedwater pipeway is directly introduced to the recycling pump. The temperature of water flowing into the recycling pump is lowered by the feedwater from the feedwater pipeway to prevent the cavitations. (Yoshino, Y.)

Progress of full MOX core design in ABWR

International Nuclear Information System (INIS)

Izutsu, S.; Sasagawa, M.; Aoyama, M.; Maruyama, H.; Suzuki, T.

2000-01-01

Full MOX ABWR core design has been made, based on the MOX design concept of 8x8 bundle configuration with a large central water rod, 40 GWd/t maximum bundle exposure, and the compatibility with 9x9 high-burnup UO 2 bundles. Core performance on shutdown margin and thermal margin of the MOX-loaded core is similar to that of UO 2 cores for the range from full UO 2 core to full MOX core. Safety analyses based on its safety parameters and MOX property have shown its conformity to the design criteria in Japan. In order to confirm the applicability of the nuclear design method to full MOX cores, Tank-type Critical Assembly (TCA) experiment data have been analyzed on criticality, power distribution and β eff /l measurements. (author)

Lungmen ABWR containment analyses during short-term main steam line break LOCA using GOTHIC

International Nuclear Information System (INIS)

Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che

2012-01-01

Highlights: ► The Lungmen ABWR containment responses due to the main steam line break are analyzed. ► In the Lungmen FSAR, the peak drywell temperature is greater than the designed value. ► GOTHIC is used to calculate the containment responses in this study. ► With more realistic conditions, the drywell temperature can be reasonably suppressed. - Abstract: Lungmen Nuclear Power Plant in Taiwan is a GE-designed twin-unit Advanced Boiling Water Reactor (ABWR) plant with rated thermal power of 3926 MWt. Both units are currently under construction. In the Lungmen Final Safety Analysis Report (FSAR) section 6.2, the calculated peak drywell temperature during the short-term Main Steam Line Break (MSLB) event is 176.3 °C, which is greater than the designed temperature of 171.1 °C. It resulted in a controversial issue in the FSAR review process conducted by the Atomic Energy Council in Taiwan. The purpose of this study is to independently investigate the Lungmen ABWR containment pressure and temperature responses to the MSLB using the GOTHIC program. Blowdown conditions are either calculated by using a simplified reactor vessel volume in GOTHIC model, or provided by the RELAP5 transient analysis. The blowdown flow rate from the steam header side is calculated with a more reasonable pressure loss coefficient of the open main steam isolation valves, and the peak drywell temperature is then reduced. By using the RELAP5 blowdown data, the peak drywell temperature can be further reduced because of the initial liquid entrainment in the blowdown flow. The drywell space is either treated as a single volume, or separated into a upper drywell and a lower drywell to reflect the real configuration of the Lungmen containment. It is also found that a single drywell volume may not present the overheating of the upper drywell. With more realistic approaches and assumptions, the drywell temperature can be reasonably below the design value and the Lungmen containment integrity

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

International Nuclear Information System (INIS)

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

2014-01-01

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

A passive decay-heat removal system for an ABWR based on air cooling

Energy Technology Data Exchange (ETDEWEB)

Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

2017-01-15

Highlights: • A passive decay heat removal system for an ABWR is discussed using combined system of the reactor and an air cooler. • Effect of number of pass of the finned heat transfer tubes on heat removal is investigated. • The decay heat can be removed by air coolers with natural convection. • Two types of air cooler are evaluated, i.e., steam condensing and water cooling types. • Measures how to improve the heat removal rate and to make compact air cooler are discussed. - Abstract: This paper describes the capability of an air cooling system (ACS) operated under natural convection conditions to remove decay heat from the core of an Advanced Boiling Water Reactor (ABWR). The motivation of the present research is the Fukushima Severe Accident (SA). The plant suffered damages due to the tsunami and entered a state of Station Blackout (SBO) during which seawater cooling was not available. To prevent this kind of situation, we proposed a passive decay heat removal system (DHRS) in the previous study. The plant behavior during the SBO was calculated using the system code NETFLOW++ assuming an ABWR with the ACS. However, decay heat removal under an air natural convection was difficult. In the present study, a countermeasure to increase heat removal rate is proposed and plant transients with the ACS are calculated under natural convection conditions. The key issue is decreasing pressure drop over the tube banks in order to increase air flow rate. The results of the calculations indicate that the decay heat can be removed by the air natural convection after safety relief valves are actuated many times during a day. Duct height and heat transfer tube arrangement of the AC are discussed in order to design a compact and efficient AC for the natural convection mode. As a result, a 4-pass heat transfer tubes with 2-row staggered arrangement is the candidate of the AC for the DHRS under the air natural convection conditions. The heat removal rate is re-evaluated as

Welding overlay analysis of dissimilar metal weld cracking of feedwater nozzle

International Nuclear Information System (INIS)

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

2010-01-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Containment behavior in MSLB with FIV malfunction

Energy Technology Data Exchange (ETDEWEB)

Choi, Hoon; Song, Dong Soo; Jun, Hwang Yong [Korea Hydro and Nuclear Power Co. Ltd, Daejeon (Korea, Republic of)

2012-10-15

In case of Main Steam Line Break(MSLB) accident, sustained high feedwater flow would cause additional cooldown of primary system. Therefore, in addition to the normal control action that closes the main feedwater valves, a safety injection signal rapidly closes all Feed water Control Valve(FCV)s and Feedwater Isolation Valve(FIV)s, trips the main feedwater pumps, and closes the feedwater pump discharge valves. With a single failure of FCVs, FIVs should act as back up protection measures. However, in a certain plant, the FIVs are not automated. If the FIVs could not be credited, the trip of main feedwater pumps can be act as back up protection measures for the single failure of FVCs. In that case, un isolated feedwater which is contained in the pipe between the main feedwater pump and the upstream of the FCV might be flash and be supplied to the broken steam generator. The containment integrity was studied for this case.

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

International Nuclear Information System (INIS)

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

1983-03-01

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

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

International Nuclear Information System (INIS)

Matal, O.; Sadilek, J.

1989-01-01

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

Nuclear power plant

International Nuclear Information System (INIS)

Igarashi, Yoko; Kato, Naoyoshi.

1982-01-01

Purpose: To decrease the reducing speed of nuclear reactor water level after the water level has reached a turbine trip level to trip the turbine thereby preventing cooling systems or the likes from undesired operation upon separation caused by the reduction of the reactor water level to a low water level before the water level control is switched to the manual control. Constitution: Two feedwater pumps arranged in parallel are operated in usual operation to feedwater to a BWR type reactor. If a trouble should occur in a feedwater controller to increase the feedwater rate and the reactor water level, one of the feedwater pumps is tripped by a signal from a feedwater pump trip device. Then, when the trip level is reached again the remaining pump is tripped. In this way, the sudden decrease in the feedwater rate and the reactor water level can be prevented. (Yoshino, Y.)

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

International Nuclear Information System (INIS)

Anoba, R.C.

1989-01-01

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

Evaluation of examination techniques for ferritic stainless steel feedwater heater tubing

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2002-07-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Aging assessment of auxiliary feedwater systems

International Nuclear Information System (INIS)

Casada, D.A.

1989-01-01

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

Hitachi's experience and achievements in ABWR construction

International Nuclear Information System (INIS)

Inoue, Takashi; Miura, Jun; Murayama, Kouichi

2003-01-01

Hitachi, Ltd. (Hitachi) is taking charge of the turbine island of Chubu Electric Power Company's Hamaoka nuclear power plant Unit No. 5 (Hamaoka 5), and all facilities of Hokuriku Electric Power Company's Shika Nuclear Power Station Unit No. 2 (Shika 2). Hitachi has achieved shorter schedules and lower costs in the two projects with the following construction methods. First, Hitachi applied modularization and the area-by-area method. These methods enable us to undertake building and installation work in parallel. Second, the floor packaging construction method enables us to finish installation in every area earlier than conventional methods. Moreover, manpower can be distributed evenly. This paper reports the application of these advanced construction methods to the Hamaoka 5 and Shika 2 projects. Installations such as RCCV modules and condenser blocks are presented in this paper. In addition, Hitachi is using 3D-CAD engineering and our own construction management system to optimize the planning of construction procedures and the installation of equipment in these projects. Thus Hitachi is always striving to improve the reliability, safety and economy of ABWR construction technologies. (author)

Development of inertia-increased reactor internal pump

International Nuclear Information System (INIS)

Tanaka, Masaaki; Matsumura, Seiichi; Kikushima, Jun; Kawamura, Shinichi; Yamashita, Norimichi; Kurosaki, Toshikazu; Kondo, Takahisa

2000-01-01

The Reactor Internal Pump (RIP) was adopted for the Reactor Recirculation System (RRS) of Advanced Boiling Water Reactor (ABWR) plants, and ten RIPs are located at the bottom of the reactor pressure vessel. In order to simplify the power supply system for the RIPs, a new inertia-increased RIP was developed, which allows to eliminate the Motor-Generator (M-G) sets. The rotating inertia was increased approximately 2.5 times of current RIP inertia by addition of flywheel on its main shaft. A full scale proving test of the inertia-increased RIP under actual plant operating conditions using full scale test loop was performed to evaluate vibration characteristics and coast down characteristics. From the results of this proving test, the validity of the new inertia-increased RIP and its power supply system (without M-G sets) was confirmed. (author)

Defence-in-depth concept for the EU-ABWR

Energy Technology Data Exchange (ETDEWEB)

Yamazaki, Hiroshi; Fuchs, Steffen; Takada, Toshiaki; Kataoka, Kazuyoshi [Toshiba International Limited (Japan)

2013-07-01

The current defence-in-depth (DiD) concept has been established by the Reactor Harmonization Working Group (RHWG) of Western European Nuclear Regulators Association (WENRA). Principally the DiD concept was already part of the very early power reactor designs. However, additional considerations have been done in order to take plant conditions into account which are beyond the original design basis. The most recent advancements have been done based on major lessons learned from the Fukushima Dai-Ichi accident. Especially for new nuclear reactors it has to be demonstrated that DiD aspects have been considered in their design. Currently Toshiba is adapting its Advanced Boiling Water Reactor (ABWR) for the European market, at first in Finland. This presentation aims to describe how the new DiD concept has been applied to achieve the safety goals for a modern reactor type and to ensure a design that can be licensed in Western Europe. (orig.)

Improvement of SSR core design for ABWR-II

International Nuclear Information System (INIS)

Moriwaki, Masanao; Aoyama, Motoo; Okada, Hiroyuki; Kitamura, Hideya; Sakurada, Koichi; Tanabe, Akira

2003-01-01

In order to enhance the spectral shift effect in the ABWR-II reactor, a novel core design to bring out better performance of spectral shift rods (SSRs) is studied. The SSR is a new type of water rod, in which the water level develops naturally during operation and changes according to the coolant flow rate through the channel. By using the SSR, the average moderator density, which is directly related to core reactivity, can be controlled over a wide range by the core flow rate. In the new SSR core design, two types of SSR bundles, in which settings for the SSR water levels are different, are utilized and loaded according to flow distribution in the core. This two-region SSR core design allows wide variation in the average SSR water level, thus improving fuel economy. Enhancement of SSR function in the two-region SSR core increases the uranium saving factor by about 25%, from the 6% of the conventional uniform SSR core to about 8%. (author)

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

Deutschmann, H.

1996-01-01

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

Development of advanced concept for shortening construction period of ABWR plant (part 2)

International Nuclear Information System (INIS)

Fujita, Tomohiro; Satoh, Takashi

2003-01-01

The construction of the first building fully applying SC structure, which is indispensable for shortening construction period of ABWR plants (21.5 months from the first concrete work to fuel loading), has been started since August 2002 in Japan. Before the construction start, a pre-construction test with some actual size SC panels was carried out to confirm the SC modular construction method. The outline of the design and the construction of the first full-SC building and the results of the pre-construction test are reported. (author)

Feedback control of primary circulation pump of PIUS-Type reactor

International Nuclear Information System (INIS)

Fujii, Mikiya; Anoda, Yoshinari; Murata, Hideo; Yonomoto, Taisuke; Kukita, Yutaka; Tasaka, Kanji.

1991-05-01

In operating the PIUS-Type reactor, it is required to keep stationary density interfaces between the primary loop hot water and the poison tank cold, borated water by maintaining pressure balance between the primary-loop and the poison-tank. The authors have developed a primary circulation pump speed control system and tested it in small-scale experiments. This control system regulates the pump speed based on measurements of the density lock differential pressure which is proportional to the elevation of the interface in the density lock. This pump speed control facilitated the normal plant operation which included core power changes. However, the elevation of the density interface indicated oscillatory behavior when the pump speed was regulated as a linear function of the density lock differential pressure. The mechanism responsible for such oscillatory behavior was found to be manometric oscillations that could be eliminated by adding a damping term to compensate for the mechanical delay of the primary pump speed. The passive shutdown function of the reactor was retained by setting an upper limit to the pump speed. This was confirmed in a loss-of-feedwater abnormal transient test. (author)

Excessive heat removal due to feedwater system malfunction

International Nuclear Information System (INIS)

Beader, D.; Peterlin, G.

1986-01-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

Zhao Xiaoyu

1996-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Inferential smart sensing for feedwater flowrate in PWRs

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

Cox, D.F.

1991-01-01

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

Fuzzy logic control of water level in advanced boiling water reactor

International Nuclear Information System (INIS)

Lin, Chaung; Lee, Chi-Szu; Raghavan, R.; Fahrner, D.M.

1995-01-01

The feedwater control system in the Advanced Boiling Water Reactor (ABWR) is more challenging to design compared to other control systems in the plant, due to the possible change in level from void collapses and swells during transient events. A basic fuzzy logic controller is developed using a simplified ABWR mathematical model to demonstrate and compare the performance of this controller with a simplified conventional controller. To reduce the design effort, methods are developed to automatically tune the scaling factors and control rules. As a first step in developing the fuzzy controller, a fuzzy controller with a limited number of rules is developed to respond to normal plant transients such as setpoint changes of plant parameters and load demand changes. Various simulations for setpoint and load demand changes of plant performances were conducted to evaluate the modeled fuzzy logic design against the simplified ABWR model control system. The simulation results show that the performance of the fuzzy logic controller is comparable to that of the Proportional-Integral (PI) controller, However, the fuzzy logic controller produced shorter settling time for step setpoint changes compared to the simplified conventional controller

MELCOR 1.8.2 calculations of selected sequences for the ABWR

International Nuclear Information System (INIS)

Kmetyk, L.N.

1994-07-01

This report summarizes the results from MELCOR calculations of severe accident sequences in the ABWR and presents comparisons with MAAP calculations for the same sequences. MELCOR was run for two low-pressure and three high-pressure sequences to identify the materials which enter containment and are available for release to the environment (source terms), to study the potential effects of core-concrete interaction, and to obtain event timings during each sequence; the source terms include fission products and other materials such as those generated by core-concrete interactions. Sensitivity studies were done on the impact of assuming limestone rather than basaltic concrete and on the effect of quenching core debris in the cavity compared to having hot, unquenched debris present

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

Science.gov (United States)

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

2015-04-01

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

Boiler feedwater treatment using reverse osmosis at Suncor OSG

International Nuclear Information System (INIS)

Brown, T.

1997-01-01

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

Automatic control device for the reduction of reactor power

International Nuclear Information System (INIS)

Sumida, Susumu; Mizuno, Hiroshi.

1982-01-01

Purpose: To early detect troubles in condensate pipeways and feedwater pipeways of BWR-type reactor. Constitution: Detectors are provided to a condensate pipe, a condensator, a low pressure condensate pump, a condensate desalting device and a high pressure condensate pump for constituting condensate pipeways, as well as to a feedwater heater, a feedwater pipe and a feedwater pump for constituting feedwater pipeways. Each of the detectors is connected by way of a lead wire to an abnormal detection and processing device. The abnormal detection and processing device, which are connected to a recycling control device, monitor the input from the detector and sends a control signal to the recycling control system upon calculation of a trouble signal from the detector. (Sekiya, K.)

Fuzzy Logic Approach to Diagnosis of Feedwater Heater Performance Degradation

International Nuclear Information System (INIS)

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

2014-01-01

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

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

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

International Nuclear Information System (INIS)

Jordan, E.L.

1992-01-01

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

Digital feedwater and recirculation flow control for GPUN Oyster Creek

International Nuclear Information System (INIS)

Burjorjee, D.; Gan, B.

1992-01-01

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

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)

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

International Nuclear Information System (INIS)

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

2014-10-01

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

Development of Advanced Concept for Shortening Construction Period of ABWR Plant

International Nuclear Information System (INIS)

Hiroshi Ijichi; Toshio Yamashita; Masahiro Tsutagawa; Hiroya Mori; Nobuaki Ooshima; Jun Miura; Minoru Kanechika; Nobuaki Miura

2002-01-01

Construction of a nuclear power plant (NPP) requires a very long period because of large amount of construction materials and many issues for negotiation among multiple sections. Shortening the construction period advances the date of return on an investment, and can also result in reduced construction cost. Therefore, the study of this subject has a very high priority for utilities. We achieved a construction period of 37 months from the first concrete work to fuel loading (F/L) (51.5 months from the inspection of the foundation (I/F) to the start of commercial operation (C/O)) at the Kashiwazaki-Kariwa NPPs No. 6 and 7 (KK-6/7), which are the first ABWR plants in the world. At TEPCO's next plant, we think that a construction period of less than 36 months (45 months from I/F to C/O) can be realized based on conventional methods such as early start of equipment installation and blocking of equipment to be brought in advance. Furthermore, we are studying the feasibility of a 21.5-month construction period (30 months from I/F to C/O) with advanced ideas and methods. The important concepts for a 21.5-month construction period are adoption of a new building structure that is the steel plate reinforced concrete (SC) structure and promotion of extensive modularization of equipment and building structure. With introducing these new concepts, we are planning the master schedule (M/S) and finding solutions to conflicts in the schedule of area release from building construction work to equipment installation work (schedule-conflicts.) In this report, we present the shortest construction period and an effective method to put it into practice for the conventional general arrangement (GA) of ABWR. In the future, we will continue the study on the improvement of building configuration and arrangements, and make clear of the concept for large composite modules of building structures and equipment. (authors)

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

International Nuclear Information System (INIS)

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

1988-01-01

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

Participation in the ABWR Man-Machine interface design. Applicability to the Spanish Electrical Sector

International Nuclear Information System (INIS)

Rodriguez, C.; Manrique Martin, A.; Nunez, J.

1997-01-01

Project coordinated by DTN within the advanced reactor programme. Participation in the design activities for the Advanced Boiling Water Reactor (ABWR) man-machine interface was divided into two phases: Phase I: Preparation of drawings for designing, developing and assessing the advanced control room Phase II: Application of these drawings in design activities Participation in this programme has led to the following possible future applications to the electrical sector: 1. Design and implementation of man-machine interfaces 2. Human factor criteria 3. Assessment of man-machine interfaces 4. Functional specification, computerised operating procedures 5. Computerised alarm prototypes. (Author)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Operational challenges to feedwater/steam generator water level control

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

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.

Analisis Termal High Pressure Feedwater Heater di PLTU PT. XYZ

Directory of Open Access Journals (Sweden)

Maria Ulfa Damayanti

2017-01-01

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

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

International Nuclear Information System (INIS)

Casada, D.A.

1988-01-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

Saiu, G.

1996-01-01

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

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.

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

International Nuclear Information System (INIS)

Nurkkala, P.; Hoikkanen, J.

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

1988-12-01

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

Study on high performance MOX fuel and core design in full MOX ABWR(1) by GNF-J

International Nuclear Information System (INIS)

Izutsu, Sadayuki; Goto, Daisuke; Saeki, Jun; Kokubun, Takehiro; Yokoya, Jun

2003-01-01

The concepts of high-performance MOX fuel using 10x10 lattices suitable for full-MOX ABWR are shown in this paper, in which average discharge exposure is extended up to 45 GWd/t with heavy-metal inventory increased over current MOX, reducing the number of refueling bundles, resulting in fuel cycle cost reduction and core performance satisfaction. Also, the increase of Pu inventory is taken into account from the viewpoint to extend the flexibility of MOX fuel utilization. (author)

Boiling water reactor

International Nuclear Information System (INIS)

Matsumoto, Tomoyuki; Inoue, Kotaro; Ishida, Masayoshi.

1975-01-01

Object: To connect a feedwater pipe to a recycling pipe line, the recycling pipe line being made smaller in diameter, thereby minimizing loss of coolant resulting from rupture of the pipe and improving safety against trouble of coolant loss. Structure: A feedwater pipe is directly connected to a recycling pipe line before a booster pump, and a mixture of recycling water and feedwater is increased in pressure by the booster pump, after which it is introduced into a jet pump in the form of water for driving the jet pump to suck surrounding water causing it to be flown into the core. In accordance with the abovementioned structure, since the flow of feedwater can be used as a part of water for driving the jet pump, the flow within the recycling pipe line may be decreased so that the recycling pipe line can be made smaller in diameter to reduce the flow of coolant in the reactor, which flows out when the pipe is ruptured. (Furukawa, Y.)

Auxiliary feedwater system aging study

International Nuclear Information System (INIS)

Kueck, J.D.

1993-07-01

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

Identification of BWR feedwater control system using autoregressive integrated model

International Nuclear Information System (INIS)

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

1983-01-01

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

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)

Design features which mitigate severe accident challenges in the GE ABWR and SBWR

International Nuclear Information System (INIS)

Buchholz, Carol E.

2004-01-01

A reduction of the requirements for the emergency planning zone (EPZ) is a goal of advanced light water reactors. The technical basis for reducing the EPZ requirements is based on a very low frequency of a severe accident and high confidence that the offsite dose would be low even if a severe accident was to occur. Design features have been included in both the ABWR and SBWR to ensure that both of these goals are achieved. Probabilistic Risk Assessments (PRAs) have been performed for both plants. The PRAs indicate a core damage frequency on the order of IE-7 for both plants. The PRAs also show that the containments will not fail even if a severe accident should occur. The potential offsite is extremely low. (author)

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

International Nuclear Information System (INIS)

Cox, D.F.

1991-01-01

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

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

International Nuclear Information System (INIS)

Pappx, L.

1994-01-01

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

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.

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.).

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

International Nuclear Information System (INIS)

Papp, L.

1995-01-01

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

Multi-unit shutdown due to boiler feedwater chemical excursion

International Nuclear Information System (INIS)

Diebel, M.E.

1991-01-01

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

Cavitation preventing device in a nuclear power plant

International Nuclear Information System (INIS)

Hirose, Masao.

1983-01-01

Purpose: To prevent the generation of cavitation upon loss of feedwater flow rate in BWR nuclear power plant by reliably and rapidly tripping a recycling pump. Constitution: Two phase streams from a nuclear reactor are divided into main steams and saturated water in a steam drum. The deviation between the corresponding flow rate of the main steams and the feedwater flow rate of the feedwater pump sending condensates to the steam drum, as well as the continuing period of the deviation are monitored. Then, if it is detected that both of the deviation and the continuing period thereof exceed specified levels, the recycling pump feeding the saturated water to the reactor is tripped. In this way, the recycling pump can be tripped rapidly and reliably upon loss of feedwater flow rate, whereby the generation of the cavitation can be prevented and the normal operation of the nuclear power plant can be insured. (Moriyama, K.)

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Feedwater heater tube-to-tubesheet connections

International Nuclear Information System (INIS)

Yokell, S.

1993-01-01

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

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

International Nuclear Information System (INIS)

Racz, A.

1990-12-01

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

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

International Nuclear Information System (INIS)

Svenson, Ola

1998-01-01

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

Ultrasonic pattern recognition study of feedwater nozzle inner radius indication

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Optimization algorithms intended for self-tuning feedwater heater model

International Nuclear Information System (INIS)

Czop, P; Barszcz, T; Bednarz, J

2013-01-01

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

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

International Nuclear Information System (INIS)

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

2010-10-01

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

Initiatives to reduce the occupational radiation exposure of ABWR plants

International Nuclear Information System (INIS)

Hirasawa, Hajime; Urata, Hidehiro; Ueda, Taku; Yamamoto, Seiji; Yaita, Yumi

2014-01-01

Toshiba has carried out radiation exposure reduction by radiation level reduction, as reduction of reactor water activated corrosion products concentration, reduction of activated corrosion products deposition and radiation shielding, and exposure time reduction, as remote control and improvement of maintenance work procedures. Water chemistry has been mainly carried out reduction of reactor water activated corrosion products concentration and reduction of activated corrosion products deposition in radiation level reduction. The reduction measures of reactor water activated corrosion products concentration are mainly reduction of iron crud concentration and reduction of cobalt ion concentration. The activated corrosion products deposition are reduced by the means of water quality control and the surface treatment. Water quality control for reduction of activated corrosion products deposition moves to ultra low iron high nickel control from Ni/Fe ratio control. The surface treatments are adopted to the stainless steel piping and carbon steel piping. As a measure further to radiation exposure reduction for ABWR (Advanced Boiling Water Reactors), we report on the effect of generation amount reduction by the adoption of alternate material and the effect of deposition reduction by material change of piping and the adoption of advanced water quality control, etc. (author)

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

International Nuclear Information System (INIS)

1985-07-01

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

Way to LWR nuclear power stations taking root

International Nuclear Information System (INIS)

Toyota, Masatoshi

1994-01-01

In this report, as for BWR nuclear power stations, the countermeasures to initial troubles, the circumstances of making bold start for the improvement and standardization after the remarkable lowering of the rate of operation, the development of A-BWR aiming at the most excellent BWR plant and the effort to reduce the construction cost are looked back, and the opinion on advancing nuclear energy development hereafter is mentioned. As the initial troubles, the thermal fatigue of reactor feedwater nozzles, the cracking of return water nozzles in control rod drive, the damage of fuel elements and others occurred. The circumstances till starting the project of improvement and standardization, and the investigation by the ad hoc committee and the results are reported. The feasibility study on A-BWR was carried out from July, 1978 to October, 1979. The basic design and the research and development for verification and the optimization design were carried out till December, 1985, as the result, the improvement of the plant performance was obtained. The effort for reducing the construction cost was successful. (K.I.)

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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.).

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1984-07-01

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

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

International Nuclear Information System (INIS)

1984-01-01

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

Simulation of the fault transitory of the feedwater controller in a Boiling water reactor with the Ramona-3B code; Simulacion del transitorio de falla del controlador de agua de alimentacion en un reactor de agua en ebullicion con el codigo RAMONA-3B

Energy Technology Data Exchange (ETDEWEB)

Hernandez M, J.L.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2005-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

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

International Nuclear Information System (INIS)

Gammage, D.; Idvorian, N.

2012-01-01

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

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

International Nuclear Information System (INIS)

Anon.

1981-01-01

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

Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

International Nuclear Information System (INIS)

Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

2008-01-01

In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

Science.gov (United States)

Shirvan, Koroush

Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2012-09-01

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

Basic philosophy of the safety design of the Toshiba boiling water reactor

International Nuclear Information System (INIS)

Sato, T.

1992-01-01

This paper discusses the safety design of the Toshiba Boiling Water Reactor (TOSBWR) which was created ∼8 years ago. The design concept is intermediate between conventional boiling water reactors (BWRs) and the advanced BWR (ABWR). It utilizes internal pumps and fine motion control rod drive, but the emergency core cooling system (ECCS) configuration is different from both conventional BWRs and the ABWR. The plant output is 1350 MW (electric). The design is based on two important philosophies: the positive cost reduction philosophy and the constant risk philosophy

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

Science.gov (United States)

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

2018-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

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

International Nuclear Information System (INIS)

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

1997-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-02-15

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

Expert system for nuclear power plant feedwater system diagnosis

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Start of the construction of initial advanced BWRs

International Nuclear Information System (INIS)

Yamada, Yasuo; Furukawa, Hideyasu; Yoshida, Tomiharu

1992-01-01

The construction of first advanced BWRs (ABWR), No. 6 and No. 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc., was started, and the commercial operation is scheduled to begin in 1996 and 1997. The ABWR with the unit capacity of 1356 MW which is the largest class in the world has been developed aiming at the further improvement of safety, operation performance, economical efficiency and so on. Hitachi Limited carried out the detailed review on new parts design and expected the perfection so as to make the first machine into the plant maintaining high reliability. The course of development of the ABWR, the outline of No. 6 and No. 7 plants in Kashiwazaki Kariwa Nuclear Power Station, the review of the design of new parts, the establishment of the technology of manufacturing RCCVs, the verification of the maintenance and checkup of internal pumps and improved type control rod driving mechanism and so on are reported. The ABWR is Japanese third improved type standard plant, but also it is expected to be adopted worldwide. (K.I.)

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.

Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

International Nuclear Information System (INIS)

1994-07-01

This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

International Nuclear Information System (INIS)

1994-07-01

This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

Energy Technology Data Exchange (ETDEWEB)

1994-07-01

This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

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

International Nuclear Information System (INIS)

Bennert, J.; Becher, L.

1977-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Ferromagnetic material inspection for feedwater heater and condenser tubes

International Nuclear Information System (INIS)

Anon.

1987-01-01

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

Recent technology for BWR nuclear steam turbine unit

International Nuclear Information System (INIS)

Moriya, Shin-ichi; Masuda, Toyohiko; Kashiwabara, Katsuto; Oshima, Yoshikuni

1990-01-01

As to the ABWR plants which is the third improvement standard boiling water reactor type plants, already the construction of a plant of 1356 MWe class for 50 Hz is planned. Hitachi Ltd. has accumulated the technology for the home manufacture of a whole ABWR plant including a turbine. As the results, the application of a butterfly type combination intermediate valve to No.5 plant in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc., which began the commercial operation recently and later plants, the application of a moisture separating heater to No.4 plant in Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., which is manufactured at present and later plants and so on were carried out. As to the steam turbine facilities for nuclear power generation manufactured by Hitachi Ltd., three turbines of 1100 MWe class for 50 Hz and one turbine for 60 Hz are in operation. As the new technologies for the steam turbines, the development of 52 in long last stage blades, the new design techniques for the rotor system, the moisture separating heater, the butterfly type combination intermediate valve, cross-around pipes and condensate and feedwater system are reported. (K.I.)

Feedback control of primary pump using midplane temperature of lower density lock for a PIUS-type reactor

International Nuclear Information System (INIS)

Tasaka, Kanji; Haga, Katsuhiro; Tamaki, Masayoshi

1993-01-01

A new automatic pump speed control system, using a measurement of the temperature distribution in the lower density lock, is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the lower density lock in order to prevent the poison water from penetrating into the core during normal operation. In a startup test, the effectiveness of this control system to bring the system quickly to the stable state from a very small initial temperature difference between top and bottom of the lower density lock has been confirmed. The effectiveness of the primary pump trip at the limit speed in the control system to shutdown the core power safely in an accident such as a loss-of-feedwater accident with and without the primary loop isolation has also been proved

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

Science.gov (United States)

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

2017-09-01

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

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

International Nuclear Information System (INIS)

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

1978-01-01

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

'Better feedwater quality through heat exchange equipment renovation'

International Nuclear Information System (INIS)

Pouzenc, C.

2002-01-01

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

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

International Nuclear Information System (INIS)

Basdekas, D.L.

1988-09-01

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

Simulation of a passive auxiliary feedwater system with TRACE5

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

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

International Nuclear Information System (INIS)

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

1979-01-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Analysis of water hammer phenomena - Application to deaerator-feedwater pump node

International Nuclear Information System (INIS)

Bigu, Melania; Tenescu, Mircea; Nita, Iulian Pavel

2008-01-01

The hydraulic hammer adverse effects are extensively presented in the literature available to those who operate and design installations in which this phenomenon occurs. There are specialized computational programs which evaluate diverse technical aspects which occur in this phenomenon. One must be noticed that not all the technical characteristics and not all effective operating modes which are treated in this paper are covered by existing computational programs. Moreover, even specialized developers of such programs recommend insistently that computational results offered by specialized programs to be verified by specialized technologists with experience in alternative theoretical computations in order to avoid any misinterpretation of results obtained by computational codes. After selective exposures of theoretical fundamentals of the problem there are presented a computational calculation obtained using the specialized calculation code PIPENET (Sunrise System Limited, Cambridge, Great Britain). The PIPENET calculation is compared with a standard computational calculation using theoretical correlations. An evaluation of the differences between those two computational methods is made in order to reveal the capabilities of the computational codes in solving the hydraulic hammering problems. In the first stage we obtained the elastic characteristics of the pipe where the phenomena of hydraulic hammering takes place. There are derivative descriptions of differential equations which describe the physical phenomena. In the second part we carried out a complete system analysis of water hammer effect due to a faulty closing of the four level control valves in steam generators. We compared the highest attended pressure with design pressure of the system. We observed that the design pressure is not overpassed. The analysis concluded that pumps' head protection is a very important parameter against overpressure in the feed water system. (authors)

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

International Nuclear Information System (INIS)

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

1990-05-01

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

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.

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

KAUST Repository

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

2015-01-01

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

Heat pump augmentation of nuclear process heat

International Nuclear Information System (INIS)

Koutz, S.L.

1986-01-01

A system is described for increasing the temperature of a working fluid heated by a nuclear reactor. The system consists of: a high temperature gas cooled nuclear reactor having a core and a primary cooling loop through which a coolant is circulated so as to undergo an increase in temperature, a closed secondary loop having a working fluid therein, the cooling and secondary loops having cooperative association with an intermediate heat exchanger adapted to effect transfer of heat from the coolant to the working fluid as the working fluid passes through the intermediate heat exchanger, a heat pump connected in the secondary loop and including a turbine and a compressor through which the working fluid passes so that the working fluid undergoes an increase in temperature as it passes through the compressor, a process loop including a process chamber adapted to receive a process fluid therein, the process chamber being connected in circuit with the secondary loop so as to receive the working fluid from the compressor and transfer heat from the working fluid to the process fluid, a heat exchanger for heating the working fluid connected to the process loop for receiving heat therefrom and for transferring heat to the secondary loop prior to the working fluid passing through the compressor, the secondary loop being operative to pass the working fluid from the process chamber to the turbine so as to effect driving relation thereof, a steam generator operatively associated with the secondary loop so as to receive the working fluid from the turbine, and a steam loop having a feedwater supply and connected in circuit with the steam generator so that feedwater passing through the steam loop is heated by the steam generator, the steam loop being connected in circuit with the process chamber and adapted to pass steam to the process chamber with the process fluid

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Future directions in boiling water reactor design

International Nuclear Information System (INIS)

Wilkins, D.R.; Hucik, S.A.; Duncan, J.D.; Sweeney, J.I.

1987-01-01

The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuver-ability; and reduced occupational exposure and radwaste. The ABWR incorporates the best proven features from BWR designs in Europe, Japan and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electrohydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling netwoek; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced trubine/generator with 52'' last stage buckets; and advanced radwaste technology. The ABWR is ready for lead plant application in Japan, where it is being developed as the next generation Japan standard BWR under the guidance and leadership of The Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. In the United States it is being adapted to the needs of US utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the US Nuclear Regulatory Commission for certification as a preapproved US standard BWR under the US Department of Energy's ALWR Design Verification Program. These cooperative Japanese and US programs are expected to establish the ABWR as a world class BWR for the 1990's...... (author)

ABWR, an option for the electric generation in Mexico; ABWR, una opcion para la generacion electrica en Mexico

Energy Technology Data Exchange (ETDEWEB)

Gomez T, A.M.; Ramirez S, J.R.; Xolocostli M, J.V. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: amgt@nuclear.inin.mx

2005-07-01

The A BWR reactor (Advanced Boiling Water Reactor) it was developed in a project group among the Company TEPCO, (Tokyo Electric Power Company), Hitachi, Toshiba and General Electric. The A BWR is the first nuclear reactor of the type BWR third generation that entered in commercial operation in the 90 Th decade. One of those main characteristics of the A BWR are that the system of external recirculation has been eliminated, that is to say, the pumps and external recirculation pipes have been replaced by 10 internal recirculation pumps mounted in the inferior part of the pressure vessel, for that external recirculation systems neither the use of jet pumps are not needed. Another important characteristic of the A BWR is the simplification of the activation of the safety systems. The simplifications in the design of the A BWR and the use of new technologies have reduced the quantity of equipment and the time of construction compared with the previous designs of BWRs. The construction project for the A BWR consists of a period of construction from 48 to 54 months, measured since that the first concrete structure is placed until that it enters in commercial operation, in accordance with the documents liberated by G E. (Author)

Atmospheric-pressure small-scale thermal-hydraulic experiment of a PIUS-type reactor

International Nuclear Information System (INIS)

Tasaka, Kanji; Tamaki, Masayoshi; Imai, Satoshi; Kohketsu, Hideto; Anoda, Yoshinari; Murata, Hideo; Kukita, Yutaka.

1992-01-01

An experimental small-scale low-pressure setup of a PIUS (Process Inherent Ultimate Safety)-type reactor was used for the examination of the stability during normal operation such as startup and load following operation and of the safety during accidents such as loss-of-feedwater and pump runaway. Automatic feedback pump control system based on differential pressure at lower honeycomb density lock was quite effective to maintain the stratified interface between primary and pool water in the honeycomb density lock during normal operation. The process inherent ultimate safety characteristics of the PIUS-type reactor was confirmed with pump-trip scram at the pump speed limit for the various simulated accidents such as a loss-of-feedwater and pump runaway. (author)

ABWR2. Innovative passive containment cooling system for the innovative ABWR

International Nuclear Information System (INIS)

Sato, Takashi; Matsumoto, Keiji

2015-01-01

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. The iB1350 uses innovative passive containment cooling system (iPCCS). The iPCCS is a part of the in-containment filtered venting system (IFVS). The vent pipe is submerged in the IFVS tank in the outer well (OW) of the Mark W containment. The conventional PCCS has a suction pipe only from the dry well (DW). On the contrary, the iPCCS has two suction pipes. One is normally opened to the wet well (WW) and another normally closed to the DW. The suction pipe in the conventional design cannot be connected to the WW because the PCCS vent pipe is connected to the WW. A PCCS functions using differential pressure between two nodes to discharge noncondensable gases in a PCCS heat exchanger (Hx). A suction pipe and a vent pipe must be connected to different nodes to use differential pressure. Therefore, the conventional PCCS never can cool the S/P. Although the S/P is the in-containment heat sink, heat up of the S/P is the most unfavorable for the conventional PCCS. In order to use the PCCS the conventional design must discharge steam directly into the DW instead of the S/P. Therefore, the conventional PCCS must open depressurization valves (DPV) at a SBO if the isolation condenser (IC) fails. On the contrary, the iPCCS can cool the S/P directly using the suction pipe connected to the WW and without DPV. Instead of DPV the iB1350 has modulating valves (MV) of which discharge lines are submerged in the S/P. Even if the IC fails at a SBO, the iB1350 can cool the core using the severe accident feedwater system (SAFWS), the SRV or the MV, and the iPCCS. The SAFWS makes up the core. The decay heat is carried by steam to the S/P through the SRV or the MV. The S/P works as in-containment heat sink. Once the S/P starts boiling the iPCCS automatically initiates cooling of the steam from the S/P. In the case of a core melt accident, a certain amount of FP is released into the S/P and heats up the S/P. Once the S

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Development of advanced boiling water reactor for medium capacity

International Nuclear Information System (INIS)

Kazuo Hisajima; Yutaka Asanuma

2005-01-01

This paper describes a result of development of an Advanced Boiling Water Reactor for medium capacity. 1000 MWe was selected as the reference. The features of the current Advanced Boiling Water Reactors, such as a Reactor Internal Pump, a Fine Motion Control Rod Drive, a Reinforced Concrete Containment Vessel, and three-divisionalized Emergency Core Cooling System are maintained. In addition, optimization for 1000 MWe has been investigated. Reduction in thermal power and application of the latest fuel reduced the number of fuel assemblies, Control Rods and Control Rod Drives, Reactor Internal Pumps, and Safety Relief Valves. The number of Main Steam lines was reduced from four to two. As for the engineered safety features, the Flammability Control System was removed. Special efforts were made to realize a compact Turbine Building, such as application of an in line Moisture Separator, reduction in the number of pumps in the Condensate and Feedwater System, and change from a Turbine-Driven Reactor Feedwater Pump to a Motor-Driven Reactor Feedwater Pump. 31% reduction in the volume of the Turbine Building is expected in comparison with the current Advanced Boiling Water Reactors. (authors)

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

Science.gov (United States)

2012-03-16

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

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

International Nuclear Information System (INIS)

Miron, A.

1998-01-01

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

Operating experience feedback report: Reliability of safety-related steam turbine-driven standby pumps. Commercial power reactors, Volume 10

International Nuclear Information System (INIS)

Boardman, J.R.

1994-10-01

This report documents a detailed analysis of failure initiators, causes and design features for steam turbine assemblies (turbines with their related components, such as governors and valves) which are used as drivers for standby pumps in the auxiliary feedwater systems of US commercial pressurized water reactor plants, and in the high pressure coolant injection and reactor core isolation cooling systems of US commercial boiling water reactor plants. These standby pumps provide a redundant source of water to remove reactor core heat as specified in individual plant safety analysis reports. The period of review for this report was from January 1974 through December 1990 for licensee event reports (LERS) and January 1985 through December 1990 for Nuclear Plant Reliability Data System (NPRDS) failure data. This study confirmed the continuing validity of conclusions of earlier studies by the US Nuclear Regulatory Commission and by the US nuclear industry that the most significant factors in failures of turbine-driven standby pumps have been the failures of the turbine-drivers and their controls. Inadequate maintenance and the use of inappropriate vendor technical information were identified as significant factors which caused recurring failures

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

International Nuclear Information System (INIS)

Finicle, D.P.

1977-01-01

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

Advanced boiling water reactor

International Nuclear Information System (INIS)

Nishimura, N.; Nakai, H.; Ross, M.A.

1999-01-01

In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

NRC Information Notice No. 92-67: Deficiency in design modifications to address failures of Hiller actuators upon a gradual loss of air pressure

International Nuclear Information System (INIS)

Rossi, C.E.

1993-01-01

On January 7, 1992, Carolina Power and Light Company (Shearon Harris Plant) components associated with the air supply to the actuators of the three main feedwater preheater bypass isolation valves were not qualified for a Q class application. Specifically, the failure of the air pump in the non-Q Class, non-seismic instrument air supply to the valve actuator accumulator could prevent pressure switches upstream of the air pump from detecting slow leakage in the Q Class, seismic portion of the actuator air lines. The pressure switches were installed to ensure valve closure by sending an automatic close signal if the instrument air system pressure (upstream of the actuator air pump) dropped to 66 psig as discussed in IN 82-25. The main feedwater preheater bypass isolation valves function as containment isolation valves upon receipt of a feedwater isolation signal. The function of the air pump is to raise the normal instrument air supply pressure from 70 to 100 psig to approximately 150 psig. If accumulator pressure drops from 150 psig to 122 psig, the main feedwater preheater bypass isolation valve may not close within 10 seconds. If pressure drops to a value as low as 20 psig, it may not be sufficient to close the main feedwater preheater bypass isolation valve and keep it closed against the maximum differential pressure across the valve seat. Upon discovery of this condition, Shearon Harris established a surveillance interval for verifying that the actuators' components were functioning properly and that the accumulators were fully pressurized. On January 12, 1992, non-Q components were replaced with suitable components and testing was completed satisfactorily

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

International Nuclear Information System (INIS)

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

2014-01-01

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

BWR 200 % recirculation pump suction line break LOCA tests, RUNs 942 and 943 at ROSA-III without HPCS

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Tasaka, Kanji; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Murata, Hideo; Koizumi, Yasuo

1986-03-01

This report presents the experimental results of RUNs 942 and 943 in ROSA-III program, which are 200 % recirculation pump suction line break LOCA tests with assumption of HPCS failure. The ROSA-III test facility simulates a BWR system with volume scale of 1/424 and has four half-length electrically heated fuel bundles, two active recirculation loops, ECCS's, and steam and feedwater systems. Effects of initial core void distribution and other fluid conditions on overall LOCA phenomena with special interest on transient core cooling phenomena were investigated by comparing the present test results with those of RUN 926, a 200 % suction line break test with standard initial fluid conditions. The initial core outlet quality was changed between 5 % and 43 %. As conclusions, (1) the initial lower core flow and higher void fraction affected significantly the core cooling conditions and resulted in earlier and higher PCT. (2) The lower plenum flashing temporarily contributed to cool down the core. (3) Flashing of remained hot water in the feedwater line affected slightly the pressure response and delayed the actuation of LPCI by 11 seconds. (4) The whole core was completely cooled down within 104 seconds after the LPCI actuation in these large break tests. (author)

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Analysis of KNU1 loss of normal feedwater

International Nuclear Information System (INIS)

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

1986-01-01

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

Water chemistry and radiation buildup at the Commonwealth Edison Company LaSalle-1 BWR. Final report

International Nuclear Information System (INIS)

Earls, C.E.; Blok, J.

1986-09-01

This report presents the results of a comprehensive study of the water quality and radiation buildup at the LaSalle County Unit 2 boiling warer reactor (BWR). The purpose of the study was to determine the effect of corrosion product inputs from the forward pumped heater drains on overall water quality. Since the drains are pumped into the feedwater line without filtration or demineralization, corrosion products in these streams will directly add to the impurity levels of the final feedwater. At LaSalle, the forward pumped heater drains contributed less to the feedwater impurities, on average, than the effluent of the condensate demineralizer. The feedwater quality at LaSalle was generally in the ''acceptable'' range. Nevertheless, significant water chemistry improvements, especially in reducing the corrosion product spikes associated with power or flow transients, is highly desirable for this plant. Such improvements should begin with a more consistent quality of demineralizer operation. Quantitative gamma scans of the primary system piping at LaSalle 2 were carried out in the course of the water chemistry study. Although the cumulative operational exposure of the plant was relatively limited at the time this study was carried out, the radiation buildup rate did appear to be rapid (in fact, among the most rapid) compared to other similar BWRs

Feedback control of a primary pump for safe and stable operation of a PIUS-type reactor

International Nuclear Information System (INIS)

Tasaka, K.; Imai, S.; Masaoka, H.; Tamaki, M.; Kukita, Y.

1993-01-01

A new automatic pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the density lock in order to prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments such as start-up and power ramping tests for the stable normal operation and a loss-of-feedwater test for the safe shutdown in an accident condition, using a small scale atmospheric pressure test loop which simulated the PIUS principle. (orig.)

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

International Nuclear Information System (INIS)

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

1993-12-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

International Nuclear Information System (INIS)

Hungerbuehler, T.; Langenstein, M.

2007-01-01

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

San Onofre/Zion auxiliary feedwater system seismic fault tree modeling

International Nuclear Information System (INIS)

Najafi, B.; Eide, S.

1982-02-01

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

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.

Protection method and protection device for liquid supply channel for nuclear reactor

International Nuclear Information System (INIS)

Sato, Masanori; Fujimoto, Sachiko

1998-01-01

In the present invention, thermal stresses exerted on feedwater pipelines and a feedwater nozzle portion in a LBR type reactor are reduced to improve integrity of a reactor, suppress addition of facilities and reliably reduce thermal stresses by a simple structure. A connection pipe channel is formed between the upper side of a horizontal portion of a pipeline for feeding water to a reactor pressure vessel and a vertical portion of the feedwater pipeline. A transferring pump is disposed in the midway of the connection pipe channel for sucking supernatant water in the horizontal portion and rendering it to join with water in the vertical portion. When supply of water is stagnated, high temperature water in the horizontal portion is transferred by the action of the transfer pump to the low temperature water in the vertical portion to join them. With such procedures, the water supplied to the horizontal portion in the feedwater pipeline is flown to suppress occurrence of heat-stratification phenomenon thereby enabling to reduce the temperature difference. (T.M.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-05-15

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

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

International Nuclear Information System (INIS)

Takamori, Kazuhide; Minato, Akihiko

1993-01-01

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

Kuosheng BWR/6 recirculation pump trip transient analysis with the RETRAN02/MOD5 code

International Nuclear Information System (INIS)

Wang, J.R.; Shih, C.

1992-01-01

A recirculation pump trip (RPT) event results in a reduction in recirculation flow, which reduces the core coolant flow rate. A reduction in core flow results in an increase in core void fraction and hence a decrease in core power due to negative void reactivity feedback. Although this category of events is less severe than others and generally considered as nonlimiting, core instability still may occur such as that at LaSalle on March 9, 1988. This paper focuses on the RPT transient analysis of Kuosheng Nuclear Power Plant (KNPP), which has two units of General Electric-designed boiling water reactor (BWR)/6 with rated core thermal power of 2894 MW and rated core flow of 10645 kg/s (23472 lb m /s). The approach to investigating the RPT transient of KNPP consists of two steps. The first step is to develop a plant-specific model using the RETRAN02/MOD5 code. In this step, various plant-specific information, including design documentation, drawings, safety analysis reports, and other information supplied by vendors were collected for model development. The RPT startup test at 68% power was used for system model benchmarking to ensure the adequacy of this model and identify several sensitive parameters. The second step is to assess whether similar power oscillation phenomena may occur at KNPP because of an RPT with isolated feedwater heater event. Two transient analyses (with or without reactor scram) of the KNPP RPT with isolated feedwater heater were investigated

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

International Nuclear Information System (INIS)

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

2014-01-01

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

A study on transient heat transfer of the EU-ABWR external core catcher using the phase-change effective convectivity model

International Nuclear Information System (INIS)

Tran Chi Thanh; Nguyen Viet Hung; Tahara, Mika; Kojima, Yoshihiro; Hamazaki, Ryoichi; Kudinov, Pavel

2015-01-01

In advanced designs of Nuclear Power Plants (NPPs), for mitigation of severe accident consequences, on the one hand, the In-Vessel Retention (IVR) concept has been implemented. On the other hand in other new NPP designs (Generation III and III+) with large power reactors, the External Core Catcher (ECC) has been widely adopted. Assessment of ECC design robustness is largely based on analysis of heat transfer of a melt pool formed in the ECC. Transient heat transfer analysis of an ECC is challenging due to (i) uncertainty in the in-vessel accident progression and subsequent vessel failure modes; (ii) long transient, (iii) high Rayleigh number and complex flows involving phase change of the melt pool formed in an ECC. The present paper is concerned with analysis of transient melt pool heat transfer in the ECC of new Advanced Boiling Water Reactor (ABWR) designed by Toshiba Corporation (Japan). According to the ABWR severe accident management strategy, the ECC is initially dry. In order to prevent steam explosion flooding is initiated after termination of melt relocation from the vessel. The ECC full of melt is cooled from the top directly by water and from the bottom through the ECC walls. In order to assess sustainability of the ECC, heat transfer simulation of a stratified melt pool formed in the ECC is carried out. The problem addressed in this work is heat flux distribution at ECC boundaries when cooling is applied (i) from the bottom, (ii) from the top and from the bottom. To perform melt pool heat transfer simulation, we employ Phase-change Effective Convectivity Model (PECM) which was originally developed as a computationally efficient, sufficiently accurate, 2D/3D accident analysis tools for simulation of transient melt pool heat transfer in the reactor lower plenum. Thermal loads from the melt pool to ECC boundaries are determined for selected ex-vessel accident scenarios. Performance of the ECC, efficiency of severe accident management (SAM) measures and

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

Control of feedwater composition of BWR power plant

International Nuclear Information System (INIS)

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

1983-01-01

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

Method of controlling the operation of a feedwater system

International Nuclear Information System (INIS)

Takemaru, Koichi; Omori, Takashi.

1975-01-01

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

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

International Nuclear Information System (INIS)

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

1981-09-01

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

Development of a dynamic model of a BWR nuclear power plant

International Nuclear Information System (INIS)

Nonboel, E.

1975-12-01

A description is given of a one-dimensional steady-state model of a high-pressure steam turbine, a low-pressure steam turbine, a moisture separator, a reheater, a condenser, feedwater heaters and feedwater pump for a nuclear power plant. The model is contained in the program ''TURBPLANT''. The dynamic part of this model is presented in part II of this report. (author)

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Steady state flow evaluations for passive auxiliary feedwater system of APR

International Nuclear Information System (INIS)

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

2012-01-01

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

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

Model extension and improvement for simulator-based software safety analysis

Energy Technology Data Exchange (ETDEWEB)

Huang, H.-W. [Department of Engineering and System Science, National Tsing Hua University (NTHU), 101 Section 2 Kuang Fu Road, Hsinchu, Taiwan (China) and Institute of Nuclear Energy Research (INER), No. 1000 Wenhua Road, Chiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)]. E-mail: hwhwang@iner.gov.tw; Shih Chunkuan [Department of Engineering and System Science, National Tsing Hua University (NTHU), 101 Section 2 Kuang Fu Road, Hsinchu, Taiwan (China); Yih Swu [Department of Computer Science and Information Engineering, Ching Yun University, 229 Chien-Hsin Road, Jung-Li, Taoyuan County 320, Taiwan (China); Chen, M.-H. [Institute of Nuclear Energy Research (INER), No. 1000Wenhua Road, Chiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Lin, J.-M. [Taiwan Power Company (TPC), 242 Roosevelt Road, Section 3, Taipei 100, Taiwan (China)

2007-05-15

One of the major concerns when employing digital I and C system in nuclear power plant is digital system may introduce new failure mode, which differs with previous analog I and C system. Various techniques are under developing to analyze the hazard originated from software faults in digital systems. Preliminary hazard analysis, failure modes and effects analysis, and fault tree analysis are the most extensive used techniques. However, these techniques are static analysis methods, cannot perform dynamic analysis and the interactions among systems. This research utilizes 'simulator/plant model testing' technique classified in (IEEE Std 7-4.3.2-2003, 2003. IEEE Standard for Digital Computers in Safety Systems of Nuclear Power Generating Stations) to identify hazards which might be induced by nuclear I and C software defects. The recirculation flow system, control rod system, feedwater system, steam line model, dynamic power-core flow map, and related control systems of PCTran-ABWR model were successfully extended and improved. The benchmark against ABWR SAR proves this modified model is capable to accomplish dynamic system level software safety analysis and better than the static methods. This improved plant simulation can then further be applied to hazard analysis for operator/digital I and C interface interaction failure study, and the hardware-in-the-loop fault injection study.

Analysis Of Feedwater Line Break Of APR1400 By MARS Code

International Nuclear Information System (INIS)

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

2011-01-01

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

Cleaning the feed-water pipeline internal surfaces

International Nuclear Information System (INIS)

Podkopaev, V.A.

1984-01-01

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

Cleaning the feed-water pipeline internal surfaces

Energy Technology Data Exchange (ETDEWEB)

Podkopaev, V.A.

1984-12-01

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

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

International Nuclear Information System (INIS)

Faure, Celine; Eade, Kevin; Fontan, Guillaume

2012-09-01

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

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

International Nuclear Information System (INIS)

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

1985-05-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Radioactive waste slurry dehydrating and drum filling device

International Nuclear Information System (INIS)

Ichihashi, Toshio; Abe, Kazuaki; Hasegawa, Akira

1981-01-01

Purpose: To obtain a device for simultaneously filling and dehydrating radioactive waste in a waste can without the necessity of a special device for dehydration. Constitution: This device includes a radioactive waste storage tank, a pump for supplying the waste from the tank to a can, a drain tube having a filter at the lower end and installed displaceable in the axial direction of the can, and a drain pump. The slurry stored in the radioactive waste storage tank is supplied by the pump to the can, and the feedwater in the slurry is removed by another pump through a drain pipe having a filter which does not pass solid content from the can. Accordingly, as the slurry is filled in the can, the feedwater contained therein is removed. Consequently, it can simultaneously dehydrate and fill the dehydrated waste in the can. (Yoshihara, H.)

Iron concentration controller in feedwater in nuclear plant

International Nuclear Information System (INIS)

Aizawa, Motohiro; Isaka, Yoshitaka

1990-01-01

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

ABWR, an option for the electric generation in Mexico

International Nuclear Information System (INIS)

Gomez T, A.M.; Ramirez S, J.R.; Xolocostli M, J.V.

2005-01-01

The A BWR reactor (Advanced Boiling Water Reactor) it was developed in a project group among the Company TEPCO, (Tokyo Electric Power Company), Hitachi, Toshiba and General Electric. The A BWR is the first nuclear reactor of the type BWR third generation that entered in commercial operation in the 90 Th decade. One of those main characteristics of the A BWR are that the system of external recirculation has been eliminated, that is to say, the pumps and external recirculation pipes have been replaced by 10 internal recirculation pumps mounted in the inferior part of the pressure vessel, for that external recirculation systems neither the use of jet pumps are not needed. Another important characteristic of the A BWR is the simplification of the activation of the safety systems. The simplifications in the design of the A BWR and the use of new technologies have reduced the quantity of equipment and the time of construction compared with the previous designs of BWRs. The construction project for the A BWR consists of a period of construction from 48 to 54 months, measured since that the first concrete structure is placed until that it enters in commercial operation, in accordance with the documents liberated by G E. (Author)

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.)

Critical heat flux in a vertical annulus under low upward flow and near atmospheric pressure

International Nuclear Information System (INIS)

Schoesse, T.; Aritomi, Masanori; Lee, Sang-Ryoul; Kataoka, Yoshiaki; Yoshioka, Yuzuru; Chung, Moon-Ki.

1997-01-01

As future boiling water reactors (BWR), concepts of evolutional ABWR (ABWR-IER) and natural circulation BWR (JSBWR) have been investigated in order to reduce their construction cost and simplify their maintenance and inspection procedures. One of the promised features of the design of the evolutional ABWR is to reduce the number of internal pumps and to remove the Motor Generation (MG) sets. These design changes may induce boiling transition in the fuel rods of reactor core during a pump trip transient due to the more rapid flow coastdown characteristics than these of the present design. In addition, the understanding of critical heat flux (CHF) is one important subject to grasp safety margin during the start-up for the natural circulation BWR and to establish the rational start-up procedure in which thermo-hydraulic instabilities can be suppressed. The present study is to clarify CHF characteristics under low velocity conditions. CHF measurements were conducted in a vertical upward annulus channel composed of an inner heated rod and an outer tube made of glass. CHF data were obtained repeatedly under the condition of stable inlet flow to examine statistically their reproducibility. The flow regime was investigated from flow observation and measurement of differential pressure fluctuation. The CHF data are correlated with the flow regime transition. It was clear from the obtained flow pattern and the CHF data that the CHF behavior could be classified into specified regions by the mass flux and inlet subcooling conditions. A CHF correlation was developed and agreed with other researchers' data within acceptable error. (author)

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

International Nuclear Information System (INIS)

Meier, J.K.

1983-01-01

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

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

International Nuclear Information System (INIS)

Hattori, Tsunekazu.

1984-01-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

International Nuclear Information System (INIS)

Dordonat, M.; Huet, M.

1994-01-01

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

Sodium pumping: pump problems

International Nuclear Information System (INIS)

Guer, M.; Guiton, P.

Information on sodium pumps for LMFBR type reactors is presented concerning ring pump design, pool reactor pump design, secondary pumps, sodium bearings, swivel joints of the oscillating annulus, and thermal shock loads

Level controlling system in BWR type reactors

International Nuclear Information System (INIS)

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

1981-01-01

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

Application of flow network models of SINDA/FLUINT{sup TM} to a nuclear power plant system thermal hydraulic code

Energy Technology Data Exchange (ETDEWEB)

Chung, Ji Bum [Institute for Advanced Engineering, Yongin (Korea, Republic of); Park, Jong Woon [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

1998-12-31

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUINT{sup TM} has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA). 5 refs., 10 figs. (Author)

Seismic qualification of PWR plant auxiliary feedwater systems

International Nuclear Information System (INIS)

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

1983-08-01

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

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

International Nuclear Information System (INIS)

Long, R.

2012-01-01

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

Pumps and pump facilities. 2. ed.

International Nuclear Information System (INIS)

Bohl, W.; Bauerfeind, H.; Gutmann, G.; Leuschner, G.; Matthias, H.B.; Mengele, R.; Neumaier, R.; Vetter, G.; Wagner, W.

1981-01-01

This book deals with the common fundamental aspects of liquid pumps and gives an exemplary choice of the most important kinds of pumps. The scientific matter is dealt with by means of practical mathematical examples among other ways of presenting the matter. Survey of contents: Division on main operational data of pumps - pipe characteristics - pump characteristics - suction behaviour of the pumps - projecting and operation of rotary pumps - boiler feed pumps - reactor feed pumps - oscillating positive-displacement pumps - eccentric spiral pumps. (orig./GL) [de

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

Science.gov (United States)

Narcisi, V.; Giannetti, F.; Del Nevo, A.; Tarantino, M.; Caruso, G.

2017-11-01

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

Evaluation of load case ''switch-off of the high pressure pump of the emergency core cooling system'', measures of verification and in situ-test

International Nuclear Information System (INIS)

Trobitz, M.; Mattheis, A.; Kerkhof, K.; Hippelein, K.; Hofstoetter, P.

1998-01-01

Within the framework of periodic safety inspection of the Gundremmingen power station (RWE-Bayernwerk - KRB II), the load collectives used for the design of safety-relevant systems and components were checked for their consistency with latest updates of the design basis. It was found that there was no analytical information or study available describing a particular process and its effects, namely switch-off of the high-pressure feedwater pump of the emergency core cooling system. The paper reports the work performed for closing the gap, including preparatory analyses, accompanying measures such as vibration measurements during plant shut-down, as well as the preparation and performance of the in-situ test. The experimental results and the comparative evaluation of calculated and experimental data are presented. (orig./CB) [de

Leak Detection of High Pressure Feedwater Heater Using Empirical Models

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

Tijerina S, F.

2008-01-01

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

Supercritical Water Reactor Cycle for Medium Power Applications

International Nuclear Information System (INIS)

BD Middleton; J Buongiorno

2007-01-01

Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency (ge)20%; Steam turbine outlet quality (ge)90%; and Pumping power (le)2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

Nuclear plant power up-rate study: feedwater heater evaluations

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Pumping behavior of sputter ion pumps

International Nuclear Information System (INIS)

Chou, T.S.; McCafferty, D.

The ultrahigh vacuum requirements of ISABELLE is obtained by distributed pumping stations. Each pumping station consists of 1000 l/s titanium sublimation pump for active gases (N 2 , H 2 , O 2 , CO, etc.), and a 20 l/s sputter ion pump for inert gases (methane, noble gases like He, etc.). The combination of the alarming production rate of methane from titanium sublimation pumps (TSP) and the decreasing pumping speed of sputter ion pumps (SIP) in the ultrahigh vacuum region (UHV) leads us to investigate this problem. In this paper, we first describe the essential physics and chemistry of the SIP in a very clean condition, followed by a discussion of our measuring techniques. Finally measured methane, argon and helium pumping speeds are presented for three different ion pumps in the range of 10 -6 to 10 -11 Torr. The virtues of the best pump are also discussed

Control rod drive mechanism

International Nuclear Information System (INIS)

Mizuno, Katsuyuki.

1976-01-01

Object: To restrict the reduction in performance due to stress corrosion cracks by making use of condensate produced in a turbine steam condenser. Structure: Water produced in a turbine steam condenser is forced into a condensed water desalting unit by low pressure condensate pump. The condensate is purified and then forced by a high pressure condensate pump into a feedwater heater for heating before it is returned to the reactor by a feedwater pump. Part of the condensate issuing from the condensate desalting unit is branched from the remaining portion at a point upstream the pump and is withdrawn into a control rod drive water pump after passing through a motordriven bypass valve, an orifice and a condenser water level control valve, is pressurized in the control rod drive water desalting unit and supplied to a control rod drive water pressure system. The control rod is vertically moved by the valve operation of the water pressure system. Since water of high oxygen concentration does not enter during normal operation, it is possible to prevent the stress cracking of the stainless steel apparatus. (Nakamura, S.)

Aging of turbine drives for safety-related pumps in nuclear power plants

International Nuclear Information System (INIS)

Cox, D.F.

1995-06-01

This study was performed to examine the relationship between time-dependent degradation and current industry practices in the areas of maintenance, surveillance, and operation of steam turbine drives for safety-related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized-water reactor plants and in the Reactor Core Isolation Cooling and High-Pressure Coolant Injection systems for boiling-water reactor plants. This research has been conducted by examination of failure data in the Nuclear Plant Reliability Data System, review of Licensee Event Reports, discussion of problems with operating plant personnel, and personal observation. The reported failure data were reviewed to determine the cause of the event and the method of discovery. Based on the research results, attempts have been made to determine the predictability of failures and possible preventive measures that may be implemented. Findings in a recent study of AFW systems indicate that the turbine drive is the single largest contributor to AFW system degradation. However, examination of the data shows that the turbine itself is a reliable piece of equipment with a good service record. Most of the problems documented are the result of problems with the turbine controls and the mechanical overspeed trip mechanism; these apparently stem from three major causes which are discussed in the text. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design, have led to improved performance resulting in a reliable safety-related component. However, these improvements have not been universally implemented

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

International Nuclear Information System (INIS)

1980-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Analysis of limit cycling on a boiler feedwater control system

International Nuclear Information System (INIS)

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

1986-01-01

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

Impeller radial force evolution in a large double-suction centrifugal pump during startup at the shut-off condition

International Nuclear Information System (INIS)

Zou, Zhichao; Wang, Fujun; Yao, Zhifeng; Tao, Ran; Xiao, Ruofu; Li, Huaicheng

2016-01-01

Highlights: • Conclude the characteristics of transient radial force in the startup process for a large double-suction centrifugal pump. • The overall direction of the radial force during startup process is also confirmed. • A formula used to calculate the transient radial force during startup process is proposed. • A relationship between radial force variation and axial vortex development in blade channel during the startup process is established. The mechanism of the radial force evolution is revealed. - Abstract: Double-suction centrifugal pumps play an important role in the main feedwater systems of nuclear power plant. The impeller radial force in a centrifugal pump varies dramatically during startup at the shut-off condition. In this study, the startup process of a large double-suction centrifugal pump is investigated using CFD. During testing, the impeller speed is accelerated from zero to its rated speed in 1.0 s (marked as t_0) and is then maintained at the rated speed. The results show that the radial force increase lags behind the impeller speed increase. At 0–0.4t_0, the radial force is small (approaching zero). At 0.4–1.4t_0, the radial force increases rapidly. After 1.4t_0, the average radial force stabilizes and reaches its maximum value of 55,619 N. The observed maximum radial force value during startup is approximately nine times as high as the radial force under rated condition. During startup, the overall radial force direction is proximate to the radial line located 25° from the volute tongue along circumferential direction. A transient radial force formula is proposed to evaluate the changes in radial force during startup. The streamline distribution in impeller passages and the impeller outlet pressure profile varying over time are produced. The relationship between radial force evolution and the varying axial-to-spiral vortex structure is analyzed. The radial force change mechanism is revealed. This research provides a scientific

Impeller radial force evolution in a large double-suction centrifugal pump during startup at the shut-off condition

Energy Technology Data Exchange (ETDEWEB)

Zou, Zhichao [College of Water Resources and Civil Engineering, China Agricultural University, Beijing 100083 (China); Wang, Fujun, E-mail: wangfj@cau.edu.cn [College of Water Resources and Civil Engineering, China Agricultural University, Beijing 100083 (China); Beijing Engineering Research Center of Safety and Energy Saving Technology for Water Supply Network System, China Agricultural University, Beijing 100083 (China); Yao, Zhifeng [College of Water Resources and Civil Engineering, China Agricultural University, Beijing 100083 (China); Beijing Engineering Research Center of Safety and Energy Saving Technology for Water Supply Network System, China Agricultural University, Beijing 100083 (China); Tao, Ran [College of Water Resources and Civil Engineering, China Agricultural University, Beijing 100083 (China); Xiao, Ruofu [College of Water Resources and Civil Engineering, China Agricultural University, Beijing 100083 (China); Beijing Engineering Research Center of Safety and Energy Saving Technology for Water Supply Network System, China Agricultural University, Beijing 100083 (China); Li, Huaicheng [Shanghai Liancheng (Group) Co., Ltd., Shanghai 201812 (China)

2016-12-15

Highlights: • Conclude the characteristics of transient radial force in the startup process for a large double-suction centrifugal pump. • The overall direction of the radial force during startup process is also confirmed. • A formula used to calculate the transient radial force during startup process is proposed. • A relationship between radial force variation and axial vortex development in blade channel during the startup process is established. The mechanism of the radial force evolution is revealed. - Abstract: Double-suction centrifugal pumps play an important role in the main feedwater systems of nuclear power plant. The impeller radial force in a centrifugal pump varies dramatically during startup at the shut-off condition. In this study, the startup process of a large double-suction centrifugal pump is investigated using CFD. During testing, the impeller speed is accelerated from zero to its rated speed in 1.0 s (marked as t{sub 0}) and is then maintained at the rated speed. The results show that the radial force increase lags behind the impeller speed increase. At 0–0.4t{sub 0}, the radial force is small (approaching zero). At 0.4–1.4t{sub 0}, the radial force increases rapidly. After 1.4t{sub 0}, the average radial force stabilizes and reaches its maximum value of 55,619 N. The observed maximum radial force value during startup is approximately nine times as high as the radial force under rated condition. During startup, the overall radial force direction is proximate to the radial line located 25° from the volute tongue along circumferential direction. A transient radial force formula is proposed to evaluate the changes in radial force during startup. The streamline distribution in impeller passages and the impeller outlet pressure profile varying over time are produced. The relationship between radial force evolution and the varying axial-to-spiral vortex structure is analyzed. The radial force change mechanism is revealed. This research

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

International Nuclear Information System (INIS)

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

1984-01-01

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

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

International Nuclear Information System (INIS)

Araujo, Jefferson Borges

1998-01-01

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

Method for keeping equipment and pipeline of nuclear power plant

International Nuclear Information System (INIS)

Okubo, Osamu.

1990-01-01

The present invention intends to suppress corrosion of equipments and pipelines in condensate, feedwater and feedwater heater drain systems during operation of a nuclear power plant. That is, condensate, feedwater and drain remained in equipments and pipelines just after the stopping of operation are passed through pipelines comprising only conduits, or they are introduced to a condensator passing through the pipelines and condensate pipes. Further, the remaining droplets on the inner surfaces are evaporated by the remaining heat of the equipments and the pipelines themselves. Then, the equipments and pipelines are isolated from other regions and kept. In view of the above, since condensate, feedwater and water feeder drains are introduced directly to the condensator passing through the conduits in which other equipments such as tanks and pumps are not present and are isolated and kept, corrosion of the equipments and the pipelines is suppressed and radioactive contamination is suppressed from prevailing by way of cruds. (I.S.)

Evaluation of cracking in steam generator feedwater piping in pressurized water reactor plants

International Nuclear Information System (INIS)

Goldberg, A.; Streit, R.D.

1981-05-01

Cracking in feedwater piping was detected near the inlet to steam generators in 15 pressurized water reactor plants. Sections with cracks from nine plants are examined with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Using transmission electron microscopy, fatigue striations are observed on replicas of cleaned crack surfaces. Calculations based on the observed striation spacings gave a cyclic stress value of 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses and it is concluded that the overriding factor in the cracking problem was the presence of such undocumented cyclic loads

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Adsorption pump for helium pumping out

International Nuclear Information System (INIS)

Donde, A.L.; Semenenko, Yu.E.

1981-01-01

Adsorption pump with adsorbent cooling by liquid helium is described. Shuttered shield protecting adsorbent against radiation is cooled with evaporating helium passing along the coil positioned on the shield. The pump is also equipped with primed cylindrical shield, cooled with liquid nitrogen. The nitrogen shield has in the lower part the shuttered shield, on the pump casing there is a valve used for pump pre-burning, and valves for connection to recipient as well. Pumping- out rates are presented at different pressures and temperatures of adsorbent. The pumping-out rate according to air at absorbent cooling with liquid nitrogen constituted 5x10 -4 Pa-3000 l/s, at 2x10 -2 Pa-630 l/s. During the absorbent cooling with liquid hydrogen the pumping-out rate according to air was at 4x10 -4 Pa-580 l/s, at 2x10 -3 Pa-680 l/s, according to hydrogen - at 8x10 -5 Pa-2500 l/s, at 5x10 -3 Pa-4200 l/s. During adsorbent cooling with liquid helium the rate of pumping-out according to hydrogen at 3x10 5 Pa-2400% l/s, at 6x10 3 Pa-1200 l/s, and according to helium at 3.5x10 -5 Pa-2800 l/s, at 4x10 -3 Pa-1150 l/s. The limit vacuum is equal to 1x10 -7 Pa. The volume of the vessel with liquid helium is equal to 3.5 l. Helium consumption is 80 cm 3 /h. Consumption of liquid nitrogen from the shield is 400 cm 3 /h. The limit pressure in the pump is obtained after forevacuum pumping-out (adsorbent regeneration) at 300 K temperature. The pump is made of copper. The pump height together with primed tubes is 800 mm diameter-380 mm [ru

Sodium hydroxide injection passivation work for the reactor water clean-up system in a new ABWR plant

International Nuclear Information System (INIS)

Wen, Tung-Jen; Lu, Ju-Huang

2012-09-01

Several studies have identified that Co-58 and Co-60 as the primary source of radiation build up on out-of-core components in new BWR plants. The deposition rate of Co on stainless steel and carbon steel is shown to be controlled mainly by the thickness of oxide films and its morphology formed through pretreatment. The passivation treatment was implemented accordingly at Lungmen unit 1 in an ABWR plant in September 2010. It is determined that the passivation conditions should be maintained at the temperature of 180∼230 deg. C, pH of 8.0∼8.5 and dissolved oxygen content over 400 ppb. The films would provide effective protection against radioactive deposition. The application of the pre-filming process on piping before the pre-operation is done during the flow induced vibration test (FIV) period. The protectiveness of stable magnetite can be increased by the pH control under the specific condition. The pre-filming control process and evaluation of passivation effectiveness is discussed in detail based on the surface analysis of the passivated specimens. Many efforts have been devoted to sodium hydroxide injection method for pH control of the system through the filter demineralizer under smooth operation. A comparison of test specimens on the properties of oxide film formed between laboratory and in-plant tests through alkaline treatment are also shown in this report. (authors)

Control room unfiltered in-leakage limit analysis of design-basis LOCA for Lungmen ABWR plant

International Nuclear Information System (INIS)

Tsai Chihming; Chang Chinjang; Yuann Yngruey

2014-01-01

In USNRC's Generic Letter 2003-01, 'Control Room Habitability,' it requests utilities provide information to demonstrate that the control room at each of their respective facilities complies with the current licensing and design bases, and applicable regulatory requirements, and that suitable design, maintenance and testing control measures are in place for maintaining this compliance. In particular, each utility is required to perform the control room in-leakage test to demonstrate that the unfiltered in-leakage rate is within that assumed in the licensing analyses. It must be ensured that the control room envelope habitability, in terms of radiation dose, is maintained during normal operations as well as design basis accidents. In view of this, a dose analysis has been performed to establish the control room unfiltered in-leakage limit which can be used as an acceptance criterion for the in-leakage test. The analysis in this study is for Lungmen ABWR plant. The plant has twin units, with each unit having its own control room. The TID-4844 source terms and associated methodology are used. The USNRC RADTRAD v3.03 code is employed for the transport calculation of radioactive materials in different paths, including control room in-leakage path. The radiological criterion on protection of the operators specified in 10 CFR 50, Appendix A, General Design Criterion 19 is followed. It's demonstrated that the performance of Lungmen control room with 500 cfm unfiltered in-leakage air could meet the radiological habitability acceptance criteria in case of radiation hazards. (author)

Recent technology for nuclear steam turbine-generator units

International Nuclear Information System (INIS)

Moriya, Shin-ichi; Kuwashima, Hidesumi; Ueno, Takeshi; Ooi, Masao

1988-01-01

As the next nuclear power plants subsequent to the present 1,100 MWe plants, the technical development of ABWRs was completed, and the plan for constructing the actual plants is advanced. As for the steam turbine and generator facilities of 1,350 MWe output applied to these plants, the TC6F-52 type steam turbines using 52 in long blades, moisture separation heaters, butterfly type intermediate valves, feed heater drain pumping-up system and other new technologies for increasing the capacity and improving the thermal efficiency were adopted. In this paper, the outline of the main technologies of those and the state of examination when those are applied to the actual plants are described. As to the technical fields of the steam turbine system for ABWRs, the improvement of the total technologies of the plants was promoted, aiming at the good economical efficiency, reliability and thermal efficiency of the whole facilities, not only the main turbines. The basic specification of the steam turbine facilities for 50 Hz ABWR plants and the main new technologies applied to the turbines are shown. The development of 52 in long last stage blades, the development of the analysis program for the coupled vibration of the large rotor system, the development of moisture separation heaters, the turbine control system, condensate and feed water system, and the generators are described. (Kako, I.)

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

International Nuclear Information System (INIS)

Ishiwatari, Yuki; Oka, Yoshiaki; Koshizuka, Seiichi

2003-01-01

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

Pumping mechanisms in sputter-ion pumps low pressure operation

International Nuclear Information System (INIS)

Welch, K.M.

1991-01-01

It is shown that significant H 2 pumping occurs in the walls of triode pumps. Also, H 2 is pumped in the anode cells of sputter-ion pumps. This pumping occurs in a manner similar to that by which the inert gases are pumped. That is, H 2 is pumped in the walls of the anode cells by high energy neutral burial. Hydrogen in the pump walls and anodes limits the base pressure of the pump

Pumping mechanisms in sputter-ion pumps low pressure operation

International Nuclear Information System (INIS)

Welch, K.M.

1991-01-01

It is shown that significant H 2 pumping occurs in the walls of triode pumps. Also, H 2 is pumped in the anode cells of sputter-ion pumps. This pumping occurs in a manner similar to that by which the inert gases are pumped. That is, H 2 pumped in the walls of the anode cells by high energy neutral burial. Hydrogen in the pump walls and anodes limits the base pressure of the pump. 13 refs., 5 figs., 1 tab

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.

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Centrifugal pumps

CERN Document Server

Anderson, HH

1981-01-01

Centrifugal Pumps describes the whole range of the centrifugal pump (mixed flow and axial flow pumps are dealt with more briefly), with emphasis on the development of the boiler feed pump. Organized into 46 chapters, this book discusses the general hydrodynamic principles, performance, dimensions, type number, flow, and efficiency of centrifugal pumps. This text also explains the pumps performance; entry conditions and cavitation; speed and dimensions for a given duty; and losses. Some chapters further describe centrifugal pump mechanical design, installation, monitoring, and maintenance. The

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-09-15

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

LMFBR with booster pump in pumping loop

International Nuclear Information System (INIS)

Rubinstein, H.J.

1975-01-01

A loop coolant circulation system is described for a liquid metal fast breeder reactor (LMFBR) utilizing a low head, high specific speed booster pump in the hot leg of the coolant loop with the main pump located in the cold leg of the loop, thereby providing the advantages of operating the main pump in the hot leg with the reliability of cold leg pump operation

Considerations on the question of applying ion exchange or reverse osmosis methods in boiler feedwater processing

International Nuclear Information System (INIS)

Marquardt, K.; Dengler, H.

1976-01-01

This consideration is to show that the method of reverse osmosis presents in many cases an interesting and economical alternative to part and total desolination plants using ion exchangers. The essential advantages of the reverse osmosis are a higher degree of automization, no additional salting of the removed waste water, small constructional volume of the plant as well as favourable operational costs with increasing salt content of the crude water to be processed. As there is a relatively high salt breakthrough compared to the ion exchange method, the future tendency in boiler feedwater processing will be more towards a combination of methods of reverse osmosis and post-purification through continuous ion exchange methods. (orig./LH) [de

Summary of commissioning of Hamaoka Nuclear Power Station Unit No.5

International Nuclear Information System (INIS)

Wakunaga, T.; Sekine, Y.; Yamada, K.; Nakamura, Y.; Kawahara, M.

2006-01-01

The Hamaoka Nuclear Power Station Unit No.5 was put into commercial operation in January 2005, which is the 1380 MWe advanced boiling water reactor (ABWR) incorporating design improvements and latest technologies of safer operation, reliability and maintenance. For example, S-FMCRD (Sealless Fine-Motion Control Rod Drive) was equipped to eliminate the use of seal housing by adopting a magnetic coupling and also ASD (Adjustable Speed Drive- the multiple drive power supply to reactor internal pumps) that can drive two or three Recirculation Internal Pumps with a large-capacity inverter. The reactor start-up tests were performed about for eleven months from February 2004 to confirm the plant's required performance including design change points. (T. Tanaka)

Domestic and overseas development of advanced boiling water reactors

International Nuclear Information System (INIS)

Hatazawa, Mamoru; Fuchino, Satoshi; Nakada, Kotaro

2010-01-01

Since Toshiba delivered the world's first advanced boiling water reactor (ABWR) to The Tokyo Electric Power Company, Inc. in 1996, we have been devoting continuous efforts to the construction and operational support of ABWR systems as major products. We are now promoting the construction of domestic and overseas ABWR systems along with the standardization of ABWRs. We are also engaged in the research and development of core technologies to support further promotion of ABWRs as a concurrent solution to the issues of global warming and energy security for individual countries. (author)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Simulations of the recent LaSalle-2 incident with the BNL plant analyzer

International Nuclear Information System (INIS)

Cheng, H.S.; Mallen, A.N.; Wulff, W.

1989-01-01

This paper presents the results of simulations of the recent power oscillation incident at the LaSalle-2 nuclear power plant using the BNL plant analyzer. The causes of the oscillation were investigated and the sensitivity of the oscillation to key parameters was studied. It is concluded that the observed power oscillation was caused by boiling instability (i.e., density wave oscillation) reinforced by the reactivity feedback in neutron kinetics, and that the density wave oscillation resulted from flow reduction due to recirculation pump trip and feedwater temperature reduction due to partial loss of feedwater heating capability as well as power peaking

The pumping of hydrogen and helium by sputter-ion pumps

International Nuclear Information System (INIS)

Welch, K.M.; Pate, D.J.; Todd, R.J.

1992-01-01

The pumping of hydrogen in diode and triode sputter-ion pumps is discussed. The type of cathode material used in these pumps is shown to have a significant impact on the effectiveness with which hydrogen is pumped. Examples of this include data for pumps with aluminum and titanium-alloy cathodes. Diode pumps with aluminum cathodes are shown to be no more effective in the pumping of hydrogen than in the pumping of helium. The use of titanium or titanium alloy anodes is also shown to measurably impact on the speed of these pumps at.very low pressures. This stems from the fact that hydrogen is x10 6 more soluble in titanium than in stainless steel. Hydrogen becomes resident in the anodes because of fast neutral burial. Lastly, quantitative data are given for the He speeds and capacities of both noble and conventional diode and triode pumps. The effectiveness of various pump regeneration procedures, subsequent to the pumping of He, is reported.These included bakeout and N 2 glow discharge cleaning. The comparative desorption of He with the subsequent pumping of N 2 is reported on. The N 2 speed of these pumps was used as the benchmark for defining the size of the pumps vs. their respective He speeds

Liquid metals pumping

International Nuclear Information System (INIS)

Le Frere, J.P.

1984-01-01

Pumps used to pump liquid metals depend on the liquid metal and on the type of application concerned. One deals more particularly with electromagnetic pumps, the main pumps used with mechanical pumps. To pump sodium in the nuclear field, these two types of pumps are used; the pumps of different circuits of Super Phenix are presented and described [fr

Continuously pumping and reactivating gas pump

International Nuclear Information System (INIS)

Batzer, T.H.; Call, W.R.

1984-01-01

Apparatus for continuous pumping using cycling cyropumping panels. A plurality of liquid helium cooled panels are surrounded by movable nitrogen cooled panels the alternatively expose or shield the helium cooled panels from the space being pumped. Gases condense on exposed helium cooled panels until the nitrogen cooled panels are positioned to isolate the helium cooled panels. The helium cooled panels are incrementally warmed, causing captured gases to accumulate at the base of the panels, where an independent pump removes the gases. After the helium cooled panels are substantially cleaned of condensate, the nitrogen cooled panels are positioned to expose the helium cooled panels to the space being pumped

Continuously pumping and reactivating gas pump

Science.gov (United States)

Batzer, Thomas H.; Call, Wayne R.

1984-01-01

Apparatus for continuous pumping using cycling cyropumping panels. A plurality of liquid helium cooled panels are surrounded by movable nitrogen cooled panels the alternatively expose or shield the helium cooled panels from the space being pumped. Gases condense on exposed helium cooled panels until the nitrogen cooled panels are positioned to isolate the helium cooled panels. The helium cooled panels are incrementally warmed, causing captured gases to accumulate at the base of the panels, where an independent pump removes the gases. After the helium cooled panels are substantially cleaned of condensate, the nitrogen cooled panels are positioned to expose the helium cooled panels to the space being pumped.

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.

Effect of pump limiter throat on pumping efficiency

International Nuclear Information System (INIS)

Ghendrih, P.; Grosman, A.; Samain, A.; Capes, H.; Morera, J.P.

1988-01-01

The necessary control of plasma edge density has led to the development of pump limiters to achieve this task. On Tore Supra, where a large part of the program is devoted to plasma edge studies, two types of such density control apparatus have been implemented, a set of pump limiters and the pumps associated to the ergodic divertor (magnetically assisted pump limiters). Generally two different kinds of pump limiters can be used, those with a throat which drives the plasma from the open edge plasma (SOL) to the neutralizer plate, and those without or with a very short throat. We are interested here in this aspect of the pump limiter concept, i.e. on the throat effect on neutral density build-up in the vicinity of the pumping plates (and hence on pumping efficieny). The underlying idea of this throat effect can be readily understood; indeed while the neutral capture in pump limiters without throats is only a ballistic effect, on expects the plasma to improve the efficiency of pump-limiters via plasma-neutral-sidewall interactions in the throat. This problem has been studied both numerically and analytically. The paper is divided as follows. In section 2, we describe the basic features of pump-limiters which are modelized by the numerical code Cezanne. Section 3 is devoted to the throat length effect considering in particular the neutral density profile in the throat and the neutral density buil-up as a function of the throat lenght. In section 4, we show that the plugging effect occurs for reasonnable values of throat lengths. An analytical value of the plugging length is discussed and compared to the values obtained numerically

Study of Hydrogen Pumping through Condensed Argon in Cryogenic pump

International Nuclear Information System (INIS)

Jadeja, K A; Bhatt, S B

2012-01-01

In ultra high vacuum (UHV) range, hydrogen is a dominant residual gas in vacuum chamber. Hydrogen, being light gas, pumping of hydrogen in this vacuum range is limited with widely used UHV pumps, viz. turbo molecular pump and cryogenic pump. Pre condensed argon layers in cryogenic pump create porous structure on the surface of the pump, which traps hydrogen gas at a temperature less than 20° K. Additional argon gas injection in the cryogenic pump, at lowest temperature, generates multiple layers of condensed argon as a porous frost with 10 to 100 A° diameters pores, which increase the pumping capacity of hydrogen gas. This pumping mechanism of hydrogen is more effective, to pump more hydrogen gas in UHV range applicable in accelerator, space simulation etc. and where hydrogen is used as fuel gas like tokamak. For this experiment, the cryogenic pump with a closed loop refrigerator using helium gas is used to produce the minimum cryogenic temperature as ∼ 14° K. In this paper, effect of cryosorption of hydrogen is presented with different levels of argon gas and hydrogen gas in cryogenic pump chamber.

Dry vacuum pumps

International Nuclear Information System (INIS)

Sibuet, R

2008-01-01

For decades and for ultimate pressure below 1 mbar, oil-sealed Rotary Vane Pumps have been the most popular solution for a wide range of vacuum applications. In the late 80ies, Semiconductor Industry has initiated the development of the first dry roughing pumps. Today SC applications are only using dry pumps and dry pumping packages. Since that time, pumps manufacturers have developed dry vacuum pumps technologies in order to make them attractive for other applications. The trend to replace lubricated pumps by dry pumps is now spreading over many other market segments. For the Semiconductor Industry, it has been quite easy to understand the benefits of dry pumps, in terms of Cost of Ownership, process contamination and up-time. In this paper, Technology of Dry pumps, its application in R and D/industries, merits over conventional pumps and future growth scope will be discussed

Off-Pump Versus On-Pump Coronary Artery Bypass Grafting

DEFF Research Database (Denmark)

Møller, Christian H; Steinbrüchel, Daniel A

2014-01-01

Coronary artery bypass grafting (CABG) remains the preferred treatment in patients with complex coronary artery disease. However, whether the procedure should be performed with or without the use of cardiopulmonary bypass, referred to as off-pump and on-pump CABG, is still up for debate....... Intuitively, avoidance of cardiopulmonary bypass seems beneficial as the systemic inflammatory response from extracorporeal circulation is omitted, but no single randomized trial has been able to prove off-pump CABG superior to on-pump CABG as regards the hard outcomes death, stroke or myocardial infarction....... In contrast, off-pump CABG is technically more challenging and may be associated with increased risk of incomplete revascularization. The purpose of the review is to summarize the current literature comparing outcomes of off-pump versus on-pump coronary artery bypass surgery....

Heat pump technology

CERN Document Server

Von Cube, Hans Ludwig; Goodall, E G A

2013-01-01

Heat Pump Technology discusses the history, underlying concepts, usage, and advancements in the use of heat pumps. The book covers topics such as the applications and types of heat pumps; thermodynamic principles involved in heat pumps such as internal energy, enthalpy, and exergy; and natural heat sources and energy storage. Also discussed are topics such as the importance of the heat pump in the energy industry; heat pump designs and systems; the development of heat pumps over time; and examples of practical everyday uses of heat pumps. The text is recommended for those who would like to kno

Liquid metal pump

Science.gov (United States)

Pennell, William E.

1982-01-01

The liquid metal pump comprises floating seal rings and attachment of the pump diffuser to the pump bowl for isolating structural deflections from the pump shaft bearings. The seal rings also eliminate precision machining on large assemblies by eliminating the need for a close tolerance fit between the mounting surfaces of the pump and the seals. The liquid metal pump also comprises a shaft support structure that is isolated from the pump housing for better preservation of alignment of shaft bearings. The shaft support structure also allows for complete removal of pump internals for inspection and repair.

Liquid metal pump

International Nuclear Information System (INIS)

Pennell, W.E.

1982-01-01

The liquid metal pump comprises floating seal rings and attachment of the pump diffuser to the pump bowl for isolating structural deflections from the pump shaft bearings. The seal rings also eliminate precision machining on large assemblies by eliminating the need for a close tolerance fit between the mounting surfaces of the pump and the seals. The liquid metal pump also comprises a shaft support structure that is isolated from the pump housing for better preservation of alignment of shaft bearings. The shaft support structure also allows for complete removal of pump internals for inspection and repair

Gas cooled reactors

International Nuclear Information System (INIS)

Kojima, Masayuki.

1985-01-01

Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

Heat pumps

CERN Document Server

Brodowicz, Kazimierz; Wyszynski, M L; Wyszynski

2013-01-01

Heat pumps and related technology are in widespread use in industrial processes and installations. This book presents a unified, comprehensive and systematic treatment of the design and operation of both compression and sorption heat pumps. Heat pump thermodynamics, the choice of working fluid and the characteristics of low temperature heat sources and their application to heat pumps are covered in detail.Economic aspects are discussed and the extensive use of the exergy concept in evaluating performance of heat pumps is a unique feature of the book. The thermodynamic and chemical properties o

Experimental Study on Series Operation of Sliding Vane Pump and Centrifugal Pump

OpenAIRE

Li, Tao; Zhang, Weiming; Jiang, Ming; Li, Zhengyang

2013-01-01

A platform for sliding vane pump and centrifugal pump tests is installed to study the series operation of them under different characteristics of pipeline. Firstly, the sliding vane pump and the centrifugal pump work independently, and the performance is recorded. Then, the two types of pumps are combined together, with the sliding vane pump acting as the feeding pump. Comparison is made between the performance of the independently working pump and the performance of series operation pump. Re...

Centrifugal pump handbook

CERN Document Server

Pumps, Sulzer

2010-01-01

This long-awaited new edition is the complete reference for engineers and designers working on pump design and development or using centrifugal pumps in the field. This authoritative guide has been developed with access to the technical expertise of the leading centrifugal pump developer, Sulzer Pumps. In addition to providing the most comprehensive centrifugal pump theory and design reference with detailed material on cavitation, erosion, selection of materials, rotor vibration behavior and forces acting on pumps, the handbook also covers key pumping applications topics and operational

Thermodynamic analysis of a supercritical water reactor

International Nuclear Information System (INIS)

Edwards, M.

2007-01-01

A thermodynamic model has been developed for a hypothetical design of a Supercritical Water Reactor, with emphasis on Canadian design criteria. The model solves for cycle efficiency, mass flows and physical conditions throughout the plant based on input parameters of operating pressures and efficiencies of components. The model includes eight feedwater heaters, three feedwater pumps, a deaerator, a condenser, the core, three turbines and two reheaters. To perform the calculations, Microsoft Excel was used in conjunction with FLUIDCAL-IAPWS95 and VBA code. The calculations show that a thermal efficiency of 47.5% can be achieved with a core outlet temperature of 625 o C. (author)

Penis Pump

Science.gov (United States)

... your appointment might be less involved. Choosing a penis pump Some penis pumps are available without a ... it doesn't get caught in the ring. Penis pumps for penis enlargement Many advertisements in magazines ...

Pumping characteristics of sputter ion pump (SIP) and titanium sublimation pump (TSP) combination

International Nuclear Information System (INIS)

Ratnakala, K.C.; Patel, R.J.; Bhavsar, S.T.; Pandiyar, M.L.; Ramamurthi, S.S.

1995-01-01

For achieving hydrocarbon free, clean ultra high vacuum, SIP-TSP combination is one of the ideal choice for pumping. For the SRS facility in Centre for Advanced Technology (CAT), we are utilising this combination, enmass. For this purpose, two modules of these combination set-ups are assembled, one with the TSP as an integral part of SIP and the other, with TSP as a separate pump mounted on the top of SIP. The pump bodies were vacuum degassed at 700 degC at 10 -5 mbar for 3 hrs. An ultimate vacuum of 3 x 10 -11 mbar was achieved, after a bake-out at 250 degC for 4 hrs, followed by continuous SIP pumping for 48 hrs, with two TSP flashing at approximately 10 hrs interval. The pump-down patterns as well as the pressure-rise patterns are studied. (author). 2 refs., 5 figs

Smart Sensing of the Aux. Feed-water Pump Performance in NPP Severe Accidents Using Advanced GMDH Method

Energy Technology Data Exchange (ETDEWEB)

No, Young Gyu; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

2016-05-15

In order to develop and verify the models, a number of data obtained by simulating station black out (SBO) scenario for the optimized power reactor 1000 (OPR1000) using MARS code were used. Most of monitoring systems for component have been suggested by using the directly measured data. However, it is very difficult to acquire data related to safety-critical component' status. Therefore, it is necessary to develop the new method that combines the data-based equipped with learning system and data miming technique. Many data-based modeling methods have been applied successfully to nuclear engineering area, such as signal validation, plant diagnostics and event identification. Also, the data miming is the process of analyzing data from different perspectives and summarizing it into useful information. In this study, the smart sensing technique was developed using advanced group method of data handing (GMDH) model. The original GMDH is an inductive self organizing algebraic model. The advanced GMDH model is equipped with a fuzzy concept. The proposed advanced GMDH model enhances the original GMDH model by reducing the effect of outliers and noise. The advanced GMDH uses different weightings according to their importance which is specified by the fuzzy membership grade. The developed model was verified using SBO accident simulation data for the OPR1000 nuclear power plant acquired with MARS code. Also, the advanced GMDH model was trained using the simulated development data and verified with simulated test data. The development and test data sets were independent. The simulation results show that the performance of the developed advanced GMDH model was very satisfactory, as shown in Table 1. Therefore, if the developed model can be optimized using diverse and specific data, it will be possible to predict the performance of Aux. feed water pump accurately.

Trojan Nuclear Plant. 1976 Annual operating report

International Nuclear Information System (INIS)

1977-01-01

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

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

International Nuclear Information System (INIS)

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

1992-10-01

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

Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

International Nuclear Information System (INIS)

1997-05-01

This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff's review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff's review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE's application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design

Final safety evaluation report related to the certification of the Advanced Boiling Water Reactor design. Supplement 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-05-01

This report supplements the final safety evaluation report (FSER) for the US Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the US Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staff`s review of the US ABWR design. The US ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staff`s review of the changes to the US ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluations, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE`s application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the US ABWR design.

Lessons learned from the SONGS Unit 1 water hammer event

International Nuclear Information System (INIS)

Chiu, C.

1987-01-01

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

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

International Nuclear Information System (INIS)

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

1991-09-01

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

Centrifugal pumps

CERN Document Server

Gülich, Johann Friedrich

2014-01-01

This book gives an unparalleled, up-to-date, in-depth treatment of all kinds of flow phenomena encountered in centrifugal pumps including the complex interactions of fluid flow with vibrations and wear of materials. The scope includes all aspects of hydraulic design, 3D-flow phenomena and partload operation, cavitation, numerical flow calculations, hydraulic forces, pressure pulsations, noise, pump vibrations (notably bearing housing vibration diagnostics and remedies), pipe vibrations, pump characteristics and pump operation, design of intake structures, the effects of highly viscous flows, pumping of gas-liquid mixtures, hydraulic transport of solids, fatigue damage to impellers or diffusers, material selection under the aspects of fatigue, corrosion, erosion-corrosion or hydro-abrasive wear, pump selection, and hydraulic quality criteria. As a novelty, the 3rd ed. brings a fully analytical design method for radial impellers, which eliminates the arbitrary choices inherent to former design procedures. The d...

Development of advanced BWR fuel bundle with spectral shift rod (3) -transient analysis of ABWR core with SSR

International Nuclear Information System (INIS)

Ikegawa, T.; Chaki, M.; Ohga, Y.; Abe, M.

2010-01-01

The spectral shift rod (SSR) is a new type of water rod, utilized instead of the conventional water rod, in which a water level develops during core operation. The water level can be changed according to the fuel channel flow rate. In this study, ABWR plant performance with SSR fuel bundles under transient conditions has been evaluated using the TRACG code. The TRACG code, which can treat three-dimensional hydrodynamic calculations in a reactor pressure vessel, is well suited for evaluating the reactor transient performance with the SSR fuel bundles because it can calculate the water levels in the SSR at each channel grouping and therefore evaluate the core reactivity according to the water level changes in the SSR. 'Generator load rejection with total turbine bypass failure' and 'Recirculation flow control failure with increasing flow' were selected as cases which may increase the reactivity with the increasing water level in the SSR. It was found that the absolute value of the void reactivity coefficient in the SSR core was larger than that in the conventional water rod core because the core averaged void fraction in the SSR core, which has the vapor region above the water level in the SSR, was larger than that in the conventional water rod core. Therefore, AMCPR for the SSR core was a little larger than that for the conventional water rod core; however, the difference was smaller than 0.02 because the inlet of the SSR ascending path was designed to be small enough to prevent the rapid water level increase in the SSR. (authors)

Evaluation of total loss of feedwater accident/recovery phase and investigation of the associated EOP

International Nuclear Information System (INIS)

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

1993-01-01

To evaluate the sequence of event and the thermohydraulic behavior during total loss of feedwater accident and recovery procedure, a RELAP5/MOD3 calculation is performed and compared with the LOFT L9-1/L3-3 experiment. Also, the predictability of the code for the major thermohydraulic phenomena following the accident is assessed. As a result, it is found that a pressure control using the spray until the time the water level reaches the top of the pressurizer, an overpressure protection by pressurizer PORV, a recovery of the secondary heat removal capability by refilling steam generator, and an effective cooldown by the continued natural circulation can be perfomed without core uncovery. It is also found that the plantspecific evaluation is necessary to confirm the effectiveness of the current symptom-oriented emergency operating procedure, especially in an overpressure protection performance and steam generator recovery performance. (Author)

Pumping characteristics of roots blower pumps for light element gases

International Nuclear Information System (INIS)

Hiroki, Seiji; Abe, Tetsuya; Tanzawa, Sadamitsu; Nakamura, Jun-ichi; Ohbayashi, Tetsuro

2002-07-01

The pumping speed and compression ratio of the two-stage roots blower pumping system were measured for light element gases (H 2 , D 2 and He) and for N 2 , in order to assess validity of the ITER torus roughing system as an ITER R and D task (T234). The pumping system of an Edwards EH1200 (nominal pumping speed of 1200 m 3 /s), two EH250s (ibid. 250 m 3 /s) and a backing pump (ibid. 100 m 3 /s) in series connection was tested under PNEUROP standards. The maximum pumping speeds of the two-stage system for D 2 and N 2 were 1200 and 1300 m 3 /h, respectively at 60 Hz, which satisfied the nominal pumping speed. These experimental data support the design validity of the ITER torus roughing system. (author)

PUMPS

Science.gov (United States)

Thornton, J.D.

1959-03-24

A pump is described for conveving liquids, particure it is not advisable he apparatus. The to be submerged in the liquid to be pumped, a conduit extending from the high-velocity nozzle of the injector,and means for applying a pulsating prcesure to the surface of the liquid in the conduit, whereby the surface oscillates between positions in the conduit. During the positive half- cycle of an applied pulse liquid is forced through the high velocity nozzle or jet of the injector and operates in the manner of the well known water injector and pumps liquid from the main intake to the outlet of the injector. During the negative half-cycle of the pulse liquid flows in reverse through the jet but no reverse pumping action takes place.

Pump characteristics and applications

CERN Document Server

Volk, Michael

2013-01-01

Providing a wealth of information on pumps and pump systems, Pump Characteristics and Applications, Third Edition details how pump equipment is selected, sized, operated, maintained, and repaired. The book identifies the key components of pumps and pump accessories, introduces the basics of pump and system hydraulics as well as more advanced hydraulic topics, and details various pump types, as well as special materials on seals, motors, variable frequency drives, and other pump-related subjects. It uses example problems throughout the text, reinforcing the practical application of the formulae

Pump element for a tube pump

DEFF Research Database (Denmark)

2011-01-01

The invention relates to a tube pump comprising a tube and a pump element inserted in the tube, where the pump element comprises a rod element and a first and a second non-return valve member positioned a distance apart on the rod element. The valve members are oriented in the same direction...... relative to the rod element so as to allow for a fluid flow in the tube through the first valve member, along the rod element, and through the second valve member. The tube comprises an at least partly flexible tube portion between the valve members such that a repeated deformation of the flexible tube...... portion acts to alternately close and open the valve members thereby generating a fluid flow through the tube. The invention further relates to a pump element comprising at least two non-return valve members connected by a rod element, and for insertion in an at least partly flexible tube in such tube...

Pumping behavior of sputtering ion pump

Energy Technology Data Exchange (ETDEWEB)

Chou, T.S.; Bittner, J.; Schuchman, J.

1991-12-31

To optimize the design of a distributed ion pump (DIP) for the Superconducting X-Ray Lithography Source (SXLS) the stability of the rotating electron cloud at very high magnetic field beyond transition, must be re-examined. In this work the pumping speed and frequency spectrum of a DIP at various voltages (1 to 10 KV) and various magnetic fields (0.1 to 4 Tesla) are measured. Three cell diameters 10 mm, 5 mm and 2.5 mm, each 8 mm long, and with 3 or 4 mm gaps between anode and cathode are investigated. In this study both the titanium cathodes and the stainless steel anode plates are perforated with holes comparable in size to the anode cell diameters. Only the partially saturated pumping behavior is under investigation. The ultimate pressure and conditioning of the pump will be investigated at a later date when the stability criterion for the electron cloud is better understood.

Pumping behavior of sputtering ion pump

Energy Technology Data Exchange (ETDEWEB)

Chou, T.S.; Bittner, J.; Schuchman, J.

1991-01-01

To optimize the design of a distributed ion pump (DIP) for the Superconducting X-Ray Lithography Source (SXLS) the stability of the rotating electron cloud at very high magnetic field beyond transition, must be re-examined. In this work the pumping speed and frequency spectrum of a DIP at various voltages (1 to 10 KV) and various magnetic fields (0.1 to 4 Tesla) are measured. Three cell diameters 10 mm, 5 mm and 2.5 mm, each 8 mm long, and with 3 or 4 mm gaps between anode and cathode are investigated. In this study both the titanium cathodes and the stainless steel anode plates are perforated with holes comparable in size to the anode cell diameters. Only the partially saturated pumping behavior is under investigation. The ultimate pressure and conditioning of the pump will be investigated at a later date when the stability criterion for the electron cloud is better understood.

Multiple pump housing

Science.gov (United States)

Donoho, II, Michael R.; Elliott; Christopher M.

2010-03-23

A fluid delivery system includes a first pump having a first drive assembly, a second pump having a second drive assembly, and a pump housing. At least a portion of each of the first and second pumps are located in the housing.

Assessing the energy efficiency of pumps and pump units background and methodology

CERN Document Server

Bernd Stoffel, em Dr-Ing

2015-01-01

Assessing the Energy Efficiency of Pumps and Pump Units, developed in cooperation with Europump, is the first book available providing the background, methodology, and assessment tools for understanding and calculating energy efficiency for pumps and extended products (pumps+motors+drives). Responding to new EU requirements for pump efficiency, and US DOE exploratory work in setting pump energy efficiency guidelines, this book provides explanation, derivation, and illustration of PA and EPA methods for assessing energy efficiency. It surveys legislation related to pump energy eff

Effect of Non-linear Velocity Loss Changes in Pumping Stage of Hydraulic Ram Pumps on Pumping Discharge Rate

Directory of Open Access Journals (Sweden)

Reza Fatahialkouhi

2018-03-01

Full Text Available The ram pump is a device which pumps a portion of input discharge to the pumping system in a significant height by using renewable energy of water hammer. The complexities of flow hydraulic on one hand and on the other hand the use of simplifying assumptions in ram pumps have caused errors in submitted analytical models for analyzing running cycle of these pumps. In this study it has been tried to modify the governing analytical model on hydraulic performance of these pumps in pumping stage. In this study by creating a logical division, the cycle of the ram pump was divided into three stages of acceleration, pumping and recoil and the governing equations on each stage of cycling are presented by using method of characteristics. Since the closing of impulse valve is nonlinear, velocity loss in pumping stage is considered nonlinearly. Also the governing equations in pumping stage were modified by considering disc elasticity of impulse valve and changing volume of the pump body when the water hammer phenomenon is occurred. In order to evaluate results and determine empirical factors of the proposed analytical model, a physical model of the ram pump is made with internal diameter of 51 mm. Results of this study are divided into several parts. In the first part, loss coefficients of the impulse valve were measured experimentally and empirical equations of drag coefficient and friction coefficient of the impulse valve were submitted by using nonlinear regression. In the second part, results were evaluated by using experimental data taken from this study. Evaluation of statistical error functions showed that the proposed model has good accuracy for predicting experimental observations. In the third part, in order to validate the results in pumping stage, the analytical models of Lansford and Dugan (1941 and Tacke (1988 were used and the error functions resulted from prediction of experimental observations were investigated through analytical models of

Performance Analysis Of Single-Pumped And Dual-Pumped Parametric Optical Amplifier

Directory of Open Access Journals (Sweden)

Sandar Myint

2015-06-01

Full Text Available Abstract In this study we present a performance analysis of single-pumped and dual- pumped parametric optical amplifier and present the analysis of gain flatness in dual- pumped Fiber Optical Parametric Amplifier FOPA based on four-wave mixing FWM. Result shows that changing the signal power and pump power give the various gains in FOPA. It is also found out that the parametric gain increase with increase in pump power and decrease in signal power. .Moreover in this paper the phase matching condition in FWM plays a vital role in predicting the gain profile of the FOPAbecause the parametric gain is maximum when the total phase mismatch is zero.In this paper single-pumped parametric amplification over a 50nm gain bandwidth is demonstrated using 500 nm highly nonlinear fiber HNLF and signal achieves about 31dB gain. For dual-pumped parametric amplification signal achieves 26.5dB gains over a 50nm gain bandwidth. Therefore dual-pumped parametric amplifier can provide relatively flat gain over a much wider bandwidth than the single-pumped FOPA.

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

International Nuclear Information System (INIS)

Hlasova, M.; Jary, A.

1997-01-01

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

76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

Science.gov (United States)

2011-01-20

... accepted the application as a docketed application for amendment to the U.S. ABWR design certification... perform the design work associated with the amended portion of the U.S. ABWR design represented by STPNOC's application and to supply the amended portion of the U.S. ABWR design. STPNOC's amendment to the U...

Residual heat removal pump and low pressure safety injection pump retrofit program

International Nuclear Information System (INIS)

Dudiak, J.G.; McKenna, J.M.

1992-01-01

Residual Heat Removal (RHR) and low pressure safety injection (LPSI) pumps installed in pressurized water-to-reactor power plants are used to provide low-head safety injection in the event of loss of coolant in the reactor coolant system. Because these pumps are subjected to rather severe temperature and pressure transients, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. Typically the pump impeller is mounted on an extended motor shaft (close-coupled configuration) and a mechanical seal is employed at the pump end of the shaft. Traditionally RHR and LPSI pumps have been a significant maintenance item for many utilities. Periodic mechanical seal of motor bearing replacement often is considered routine maintenance. The closed-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. This paper introduces a design modification developed to convert the close-coupled RHR and LPSI pumps to a coupled configuration

Pump Application as Hydraulic Turbine – Pump as Turbine (PaT)

OpenAIRE

Rusovs, D

2009-01-01

The paper considers pump operation as hydraulic turbine with purpose to produce mechanical power from liquid flow. The Francis hydraulic turbine was selected for comparison with centrifugal pump in reverse operation. Turbine and centrifugal pump velocity triangles were considered with purpose to evaluate PaT efficiency. Shape of impeller blades for turbine and pumps was analysed. Specific speed calculation is carried out with purpose to obtain similarity in pump and turbine description. For ...

Fundamentals of boiling water reactor systems

International Nuclear Information System (INIS)

Mattern, J.

1976-01-01

The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator, dryer assemblies, feedwater spargers, internal recirculation pumps and control rod drive housings. Connected to the steam lines are the pressure relief valves which protect the pressure boundary from damage due to overpressure. (orig./TK) [de

Heat pumps

CERN Document Server

Macmichael, DBA

1988-01-01

A fully revised and extended account of the design, manufacture and use of heat pumps in both industrial and domestic applications. Topics covered include a detailed description of the various heat pump cycles, the components of a heat pump system - drive, compressor, heat exchangers etc., and the more practical considerations to be taken into account in their selection.

Reactor coolant purification system circulation pumps (CUW pumps)

International Nuclear Information System (INIS)

Tsutsui, Toshiaki

1979-01-01

Coolant purification equipments for BWRs have been improved, and the high pressure purifying system has become the main type. The quantity of purifying treatment also changed to 2% of the flow rate of reactor feed water. As for the circulation pumps, canned motor pumps are adopted recently, and the improvements of reliability and safety are attempted. The impurities carried in by reactor feed water and the corrosion products generated in reactors and auxiliary equipments are activated by neutron irradiation or affect heat transfer adversely, adhering to fuel claddings are core structures. Therefore, a part of reactor coolant is led to the purification equipments, and returned to reactors after the impurities are eliminated perfectly. At the time of starting and stopping reactors, excess reactor water and the contaminated water from reactors are transferred to main condenser hot wells or waste treatment systems. Thus the prescribed water quality is maintained. The operational modes of and the requirements for the CUW pumps, the construction and the features of the canned motor type CUW pumps are explained. Recently, a pump operated for 11 months without any maintenance has been disassembled and inspected, but the wear of bearings has not been observed, and the high reliability of the pump has been proved. (Kako, I.)

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

Lapa, Celso Marcelo Franklin.

1996-05-01

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

Work plan, AP-102 mixer pump removal and pump replacement

International Nuclear Information System (INIS)

Jimenez, R.F.

1994-01-01

The objective of this work plan is to plan the steps and estimate the costs required to remove the failed AP-102 mixer pump, and to plan and estimate the cost of the necessary design and specification work required to order a new, but modified, mixer pump including the pump and pump pit energy absorbing design. The main hardware required for the removal of the mixer is as follows: a flexible receiver and blast shield; a metal container for the pulled mixer pump; and a trailer and strongback to haul and manipulate the container. Additionally: a gamma scanning device will be needed to detect the radioactivity emanating from the mixer as it is pulled from the tank; a water spray system will be required to remove tank waste from the surface of the mixer as it is pulled from the AP-102 tank; and a lifting yoke to lift the mixer from the pump pit (the SY-101 Mixer Lifting Yoke will be used). A ''green house'' will have to be erected over the AP-102 pump pit and an experienced Hoisting and Rigging crew must be assembled and trained in mixer pump removal methods before the actual removal is undertaken

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

Directory of Open Access Journals (Sweden)

Nyers Jozsef

2017-01-01

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

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

International Nuclear Information System (INIS)

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

1993-02-01

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

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

International Nuclear Information System (INIS)

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

1994-05-01

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

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

International Nuclear Information System (INIS)

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

1992-10-01

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

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

International Nuclear Information System (INIS)

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

1993-02-01

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

Electromagnetic pump technology

International Nuclear Information System (INIS)

Prabhakar, R.

1994-01-01

Fast Breeder Reactors have an important role to play in our nuclear power programme. Liquid metal sodium is used as the coolant for removing fission heat generated in fast reactors and a distinctive physical property of sodium is its high electrical conductivity. This enables application of electromagnetic (EM) pumps, working on same principle as electric motors, for pumping liquid sodium. Due to its lower efficiency as compared to centrifugal pumps, use of EM pumps has been restricted to reactor auxiliary circuits and experimental facilities. As part of our efforts to manufacture fast reactor components indigenously, work on the development of two types of EM pumps namely flat linear induction pump (FLIP) and annular linear induction pump (ALIP) has been undertaken. Design procedures based on an equivalent circuit approach have been established and results from testing a 5.6 x 10E-3 Cum/s (20 Cum/h) FLIP in a sodium loop were used to validate the procedure. (author). 7 refs., 6 figs

Safety design/analysis and scenario for prevention of CDA with ECCS in lead-bismuth-cooled fast reactor

International Nuclear Information System (INIS)

Minoru, Takahashi; Vaclav, Dostal; Abu Khalid, Rivai; Novitrian; Yumi, Yamada

2007-01-01

Safety design has been developed to show safety feature of Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR). The core is designed to have negative void reactivity even if the entire core and upper plenum are voided by steam intrusion from above. In-vessel type control rod driving mechanisms are used to prevent control rods from accidental ejection due to high pressure in the reactor vessel. In cases of coolant leakage from reactor vessel and feed water pipes, Pb-Bi coolant level in the reactor vessel is kept at the required level for decay heat removal by means of closed type guard vessel. Dual pipes are adopted to avoid leak of water in the feedwater system. Pump trip in feedwater systems initiates loss of coolant flow (LOF) event, although there is no concern of loss of flow accident due to primary pump trip. Injection of high pressure water slows down the flow-coast-down of feedwater at the LOF event. It has been evaluated that the fuel temperature is kept lower than safety limits at the unprotected loss of flow and heat sink (ATWS). A scenario for prevention of the core disruptive accident (CDA) with the emergency core cooling system (ECCS) is examined. The reactor becomes super-critical when the reactor vessel is filled with water. It is necessary to use water with boric acid for the ECC system, and additional backup rods for sub-critical core in water injection. (authors)

ECOSIM - Applied to a study on the thermo-hydraulic behaviour of feedwater heaters

International Nuclear Information System (INIS)

Huelamo Martinez, E.; Casado Flores, E.; Bosch Aparicio, F.

1998-01-01

In order to carry out a behaviour study on the secondary circuit of a nuclear power plant operating at a load level higher than originally planned, it is essential to know if the cycle heaters are valid from the thermo-dynamic point of view. This paper describes the models which were used for the study of certain heaters; these models were validated by checking that they faithfully reproduced the behaviour of the equipment (TTD and DCA) in areas where data from the manufacturer was available. The behaviour of said equipment was later obtained in the foreseen operating range. The calculations necessary for these studies were carried out by building ECOSIM models, taking into account that the behaviour of the feedwater heaters depends both on the entry conditions of the extraction steam and also on the remaining mass and energy inputs. For this reason the actual plant layout was taken into consideration, as it was different from the original design. This paper describes the starting hypothesis, the correlations used, the results obtained, an analysis of said results, and a comparison with the manufacturer's data where available. (Author)

Pumping life

DEFF Research Database (Denmark)

Sitsel, Oleg; Dach, Ingrid; Hoffmann, Robert Daniel

2012-01-01

The name PUMPKIN may suggest a research centre focused on American Halloween traditions or the investigation of the growth of vegetables – however this would be misleading. Researchers at PUMPKIN, short for Centre for Membrane Pumps in Cells and Disease, are in fact interested in a large family o......’. Here we illustrate that the pumping of ions means nothing less than the pumping of life....

Champagne Heat Pump

Science.gov (United States)

Jones, Jack A.

2004-01-01

The term champagne heat pump denotes a developmental heat pump that exploits a cycle of absorption and desorption of carbon dioxide in an alcohol or other organic liquid. Whereas most heat pumps in common use in the United States are energized by mechanical compression, the champagne heat pump is energized by heating. The concept of heat pumps based on other absorption cycles energized by heat has been understood for years, but some of these heat pumps are outlawed in many areas because of the potential hazards posed by leakage of working fluids. For example, in the case of the water/ammonia cycle, there are potential hazards of toxicity and flammability. The organic-liquid/carbon dioxide absorption/desorption cycle of the champagne heat pump is similar to the water/ammonia cycle, but carbon dioxide is nontoxic and environmentally benign, and one can choose an alcohol or other organic liquid that is also relatively nontoxic and environmentally benign. Two candidate nonalcohol organic liquids are isobutyl acetate and amyl acetate. Although alcohols and many other organic liquids are flammable, they present little or no flammability hazard in the champagne heat pump because only the nonflammable carbon dioxide component of the refrigerant mixture is circulated to the evaporator and condenser heat exchangers, which are the only components of the heat pump in direct contact with air in habitable spaces.

Pumping of methane by an ionization assisted Zr/Al getter pump

International Nuclear Information System (INIS)

Shen, G.L.

1987-01-01

The pumping of methane by an ionization assisted Zr/Al getter pump has been investigated. This pump consists of 12 pieces of ring getters. A spiral shape W filament is located within the ring getters. A bias voltage is applied across the filament and the getter itself. The experiments have shown that (1) when the bias voltage is turned off, the pumping speed of the getter pump for methane increases exponentially with the filament temperature; (2) when the filament temperature is held constant, its pumping speed varies directly with the ionization electron current; (3) when the filament temperature is 2063 0 C and the electron current is 57 mA, the pumping speed of the Zr/Al getter pump is 475 ml/s, and the specific speed is 16.8 ml/s cm 2 ; and (4) an activation energy and critical temperature measured for methane molecules decomposition are, respectively, 47.4 kcal/mol and about 1700 0 C. Analysis of the results indicates that methane is pumped by an ionization assisted Zr/Al getter pump not because of the adsorption and the diffusion of methane molecules directly, but because methane molecules are decomposed as C and H 2 through a catalysis of the hot W filament, carbon is adsorbed on the surface of the W filament, and is diffused into the interior of the W lattice. H 2 is immediately absorbed by the Zr/Al getters. Besides, electron impact with CH 4 would result in the additional decomposition and ionization, then the effect of electron bombardment enhances methane pumping by the Zr/Al getters

Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

Energy Technology Data Exchange (ETDEWEB)

1993-05-13

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

Study of Pu consumption in Advanced Light Water Reactors

International Nuclear Information System (INIS)

1993-01-01

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE's 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology

Breastfeeding FAQs: Pumping

Science.gov (United States)

... of pump is best? You can buy or rent a breast pump from lactation consultants, hospitals, retail ... place to do it. Many companies offer their employees pumping and nursing areas. If yours doesn't, ...

Recent development of advanced BWR technology for plant application

International Nuclear Information System (INIS)

Horiuchi, Tetsuo; Sakurai, Mikio; Mase, Noriaki; Oyamada, Osamu; Nakadaira, Shiro.

1988-01-01

The development of advanced BWRs (ABWR) was completed in 1985. Through the authorization as the third improved and standardized plants of LWRs by the Ministry of International Trade and Industry, the detailed design phase of the actual project has begun, and the improved technology to be applied to the plants has been steadily verified and put in practical use. In the ABWRs, the operational capability, safety and economical efficiency as the general characteristics of the plants were further heightened by simplifying, heightening the performance and compactifying. Particularly internal pumps brought about the improvement of the operational capability and safety of the plants together with improved control rod driving system, besides promoting the simplification and compactification of the reactor system. Also the reinforced concrete containment vessels constructed into one body with the buildings have the compact structure and the solid property sufficiently withstanding hypothetical accidents, and contribute to the improvement of the economical efficiency and safety. These key improvement technologies completed their tests for practical use, and it was shown that the expected objectives are realized with the characteristics of high level, thus the steady steps toward the construction of the actual plants are promoted. (Kako, I.)

Large nuclear steam turbine plants

International Nuclear Information System (INIS)

Urushidani, Haruo; Moriya, Shin-ichi; Tsuji, Kunio; Fujita, Isao; Ebata, Sakae; Nagai, Yoji.

1986-01-01

The technical development of the large capacity steam turbines for ABWR plants was partially completed, and that in progress is expected to be completed soon. In this report, the outline of those new technologies is described. As the technologies for increasing the capacity and heightening the efficiency, 52 in long blades and moisture separating heaters are explained. Besides, in the large bore butterfly valves developed for making the layout compact, the effect of thermal efficiency rise due to the reduction of pressure loss can be expected. As the new technology on the system side, the simplification of the turbine system and the effect of heightening the thermal efficiency by high pressure and low pressure drain pumping-up method based on the recent improvement of feed water quality are discussed. As for nuclear steam turbines, the actual records of performance of 1100 MW class, the largest output at present, have been obtained, and as a next large capacity machine, the development of a steam turbine of 1300 MWe class for an ABWR plant is in progress. It can be expected that by the introduction of those new technologies, the plants having high economical efficiency are realized. (Kako, I.)

Development and evaluation of cryosorption pump and cryotrapping pump for CTR applications

International Nuclear Information System (INIS)

Kuribayashi, S.; Ota, H.; Sato, H.

1986-01-01

In order to obtain the engineering data to design compound cryopump for CTR, the authors tested the cryosorption pump and cryotrapping pump. The cryosorption panel was consisted of coconut charcoal metallically bonded to 4.2K cryopanel by brazing. The initial pumping speed of helium of cryosorption pump was found to be ≅2.2 iota/scm/sup 2/. The speed dropped off with loading (about 8 Torr iota/cm/sup 2/) to 1.5 iota/scm/sup 2/. The initial helium pumping speed of the 4.2K cryotrapping pump by argon spray was found to be ≅6 iota/scm/sup 2/. The speed, however, dropped off with loading (≅0.3 Torr iota/cm/sup 2/) to less than 5%. These results indicate that the cryosorption pump by coconut charcoal is superior to the cryotrapping pump, because the capacity of the former is larger than the latter

Efficiency and threshold pump intensity of CW solar-pumped solid-state lasers

Science.gov (United States)

Hwang, In H.; Lee, Ja H.

1991-01-01

The authors consider the relation between the threshold pumping intensity, the material properties, the resonator parameters, and the ultimate slope efficiencies of various solid-state laser materials for solar pumping. They clarify the relation between the threshold pump intensity and the material parameters and the relation between the ultimate slope efficiency and the laser resonator parameters such that a design criterion for the solar-pumped solid-state laser can be established. Among the laser materials evaluated, alexandrite has the highest slope efficiency of about 12.6 percent; however, it does not seem to be practical for a solar-pumped laser application because of its high threshold pump intensity. Cr:Nd:GSGG is the most promising for solar-pumped lasing. Its threshold pump intensity is about 100 air-mass-zero (AM0) solar constants and its slope efficiency is about 12 percent when thermal deformation is completely prevented.

Main feedwater valve diagnostics at Waterford 3 nuclear generating station

International Nuclear Information System (INIS)

Fitzgerald, W.V.

1991-01-01

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

Classification of Feedwater Heater Performance Degradation Using Residual Sign Matrix

International Nuclear Information System (INIS)

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

2016-01-01

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

Classification of Feedwater Heater Performance Degradation Using Residual Sign Matrix

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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)

Pumping station design for a pumped-storage wind-hydro power plant

International Nuclear Information System (INIS)

Anagnostopoulos, John S.; Papantonis, Dimitris E.

2007-01-01

This work presents a numerical study of the optimum sizing and design of a pumping station unit in a hybrid wind-hydro plant. The standard design that consists of a number of identical pumps operating in parallel is examined in comparison with two other configurations, using one variable-speed pump or an additional set of smaller jockey pumps. The aim is to reduce the amount of the wind generated energy that cannot be transformed to hydraulic energy due to power operation limits of the pumps and the resulting step-wise operation of the pumping station. The plant operation for a period of one year is simulated by a comprehensive evaluation algorithm, which also performs a detailed economic analysis of the plant using dynamic evaluation methods. A preliminary study of the entire plant sizing is carried out at first using an optimization tool based on evolutionary algorithms. The performance of the three examined pumping station units is then computed and analyzed in a comparative study. The results reveal that the use of a variable-speed pump constitutes the most effective and profitable solution, and its superiority is more pronounced for less dispersed wind power potential

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

International Nuclear Information System (INIS)

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

1986-01-01

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

BWR series pump recirculation system

International Nuclear Information System (INIS)

Dillmann, C.W.

1992-01-01

This patent describes a recirculation system for driving reactor coolant water contained in an annular downcomer defined between a boiling water reactor vessel and a reactor core spaced radially inwardly therefrom. It comprises a plurality of circumferentially spaced second pumps disposed in the downcomer, each including an inlet for receiving from the downcomer a portion of the coolant water as pump inlet flow, and an outlet for discharging the pump inlet flow pressurized in the second pump as pump outlet flow; and means for increasing pressure of the pump inlet flow at the pump inlet including a first pump disposed in series flow with the second pump for first receiving the pump inlet flow from the downcomer and discharging to the second pump inlet flow pressurized in the first pump

Early outcome of off-pump versus on-pump coronary revascularization

African Journals Online (AJOL)

Introduction: The use of coronary artery bypass surgery (CABG) with cardiopulmonary bypass (CPB) or without CPB technique (off-pump) can be associated with different mortality and morbidity and their outcomes remain uncertain. The goal of this study was to evaluate the early outcome of on-pump versus off-pump CABG.

Pumps in wearable ultrafiltration devices: pumps in wuf devices.

Science.gov (United States)

Armignacco, Paolo; Garzotto, Francesco; Bellini, Corrado; Neri, Mauro; Lorenzin, Anna; Sartori, Marco; Ronco, Claudio

2015-01-01

The wearable artificial kidney (WAK) is a device that is supposed to operate like a real kidney, which permits prolonged, frequent, and continuous dialysis treatments for patients with end-stage renal disease (ESRD). Its functioning is mainly related to its pumping system, as well as to its dialysate-generating and alarm/shutoff ones. A pump is defined as a device that moves fluids by mechanical action. In such a context, blood pumps pull blood from the access side of the dialysis catheter and return the blood at the same rate of flow. The main aim of this paper is to review the current literature on blood pumps, describing the way they have been functioning thus far and how they are being engineered, giving details about the most important parameters that define their quality, thus allowing the production of a radar comparative graph, and listing ideal pumps' features. © 2015 S. Karger AG, Basel.

TFTR ultrahigh-vacuum pumping system incorporating mercury diffusion pumps

International Nuclear Information System (INIS)

Sink, D.A.; Sniderman, M.

1976-06-01

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

World class BWRs for the 1990s

International Nuclear Information System (INIS)

Wilkins, D.; Lowton, R.

1989-01-01

The first step in Advanced BWR development took place in 1978. It was the formation in San Jose, California, of an Advanced Engineering Team, consisting of representatives of worldwide BWR suppliers and leading architect engineer firms. The Advanced Engineering Team effort was completed in 1979 and established the basic ABWR concept. During the period 1980-85, ABWR development was focused in Japan. Under the sponsorship of the Tokyo Electric Power Company and Japanese BWR utilities, GE, Hitachi and Toshiba performed design development studies to confirm the technical and economic attractiveness of the ABWR, and also conducted a comprehensive test programme encompassing the new features. In 1987, The Tokyo Electric Power Company announced its decision to proceed with two ABWR units at its Kashiwazaki-Kariwa nuclear power station with commercial operation of the first unit in 1996 and the second in 1998. The company submitted the application for the establishment permit for the two ABWR units in May 1988. The units will be supplied by a joint venture of the three companies who worked on the development studies. This article goes on to describe the ABWR design resulting form this international development effort. (3 figures, 1 table). (Author)

An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

International Nuclear Information System (INIS)

Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

2012-09-01

To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE - ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

Nuclear power for the new millennium

International Nuclear Information System (INIS)

Hucik, S.A.; Redding, J.R.

1998-01-01

Advanced nuclear technology is being commercially deployed. Two ABWR's have been constructed in Japan and are reliably generating large amounts of low cost electricity. Taiwan is now in the process of licensing and constructing two more ABWR's, which will enter commercial operation in 2004 and 2005. Other countries have similar strategies to deploy advanced nuclear plants and the successful deployment of ABWR's in Japan and Taiwan, coupled with international agreements to limit CO 2 emissions, will only reinforce these plans. The ABWR will play an important role in meeting the conflicting needs of developed and developing economies for more massive amounts of electricity and the need worldwide to limit CO 2 emissions. Successful ABWR projects in Japan arid Taiwan, coupled with licensing approval in the United States, represent the new approach to the design, licensing, construction and operation of nuclear power in the new millennium. (author)

Photovoltaic pump systems

Science.gov (United States)

Klockgether, J.; Kiessling, K. P.

1983-09-01

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

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

CERN Document Server

Merkle, Thomas

2014-01-01

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

Electrokinetic pump

Science.gov (United States)

Patel, Kamlesh D.

2007-11-20

A method for altering the surface properties of a particle bed. In application, the method pertains particularly to an electrokinetic pump configuration where nanoparticles are bonded to the surface of the stationary phase to alter the surface properties of the stationary phase including the surface area and/or the zeta potential and thus improve the efficiency and operating range of these pumps. By functionalizing the nanoparticles to change the zeta potential the electrokinetic pump is rendered capable of operating with working fluids having pH values that can range from 2-10 generally and acidic working fluids in particular. For applications in which the pump is intended to handle highly acidic solutions latex nanoparticles that are quaternary amine functionalized can be used.

Pumping behavior of ion pump elements at high and misaligned magnetic fields

Energy Technology Data Exchange (ETDEWEB)

Hseuh, H. C.; Jiang, W. S.; Mapes, M.

1994-11-01

The pumping speeds of several diode type ion pump elements with cell radii of 5 mm, 5.5 mm, 6 mm, 9 mm and 12 mm were measured while being subjected to a magnetic field B, ranging from 500 Gauss up to 15 KG, and misalignment angles (angles between the direction of B and the anode axis) from 0 to 13 degrees. The pumping speeds of elements with the 9 mm and 12 mm cells peaked around 1--2 KG, then dropped off rapidly with an increasing magnetic field. The pumping speeds of the smaller cell elements remained constant with an increasing magnetic field. The pumping speeds of all the elements decreased with increasing misalignment. The measured pumping speeds from this study are 3--4 times lower than the calculated pumping, speeds using previously reported empirical formulae.

Pumping behavior of ion pump elements at high and misaligned magnetic fields

International Nuclear Information System (INIS)

Hseuh, H.C.; Jiang, W.S.; Mapes, M.

1994-01-01

The pumping speeds of several diode type ion pump elements with cell radii of 5 mm, 5.5 mm, 6 mm, 9 mm and 12 mm were measured while being subjected to a magnetic field B, ranging from 500 Gauss up to 15 KG, and misalignment angles (angles between the direction of B and the anode axis) from 0 to 13 degrees. The pumping speeds of elements with the 9 mm and 12 mm cells peaked around 1--2 KG, then dropped off rapidly with an increasing magnetic field. The pumping speeds of the smaller cell elements remained constant with an increasing magnetic field. The pumping speeds of all the elements decreased with increasing misalignment. The measured pumping speeds from this study are 3--4 times lower than the calculated pumping, speeds using previously reported empirical formulae

Liquid metal pump

International Nuclear Information System (INIS)

Pennell, W.E.

1981-01-01

A liquid metal pump comprising a shaft support structure which is isolated from the pump housing for better preservation of alignment of shaft bearings. The shaft carries an impeller and the support structure carries an impeller cage which is slidably disposed in a diffuser so as to allow complete removal of pump internals for inspection and repair. The diffuser is concentrically supported in the pump housing which also takes up all reaction forces generated by the discharge of the liquid metal from the diffuser, with floating seals arranged between impeller cage and the diffuser. The space between the diffuser and the pump housing permits the incoming liquid to essentially surround the diffuser. (author)

Rotary piston blood pumps: past developments and future potential of a unique pump type.

Science.gov (United States)

Wappenschmidt, Johannes; Autschbach, Rüdiger; Steinseifer, Ulrich; Schmitz-Rode, Thomas; Margreiter, Raimund; Klima, Günter; Goetzenich, Andreas

2016-08-01

The design of implantable blood pumps is either based on displacement pumps with membranes or rotary pumps. Both pump types have limitations to meet the clinical requirements. Rotary piston blood pumps have the potential to overcome these limitations and to merge the benefits. Compared to membrane pumps, they are smaller and with no need for wear-affected membranes and valves. Compared to rotary pumps, the blood flow is pulsatile instead of a non-physiological continuous flow. Furthermore, the risk of flow-induced blood damage and platelet activation may be reduced due to low shear stress to the blood. The past developments of rotary piston blood pumps are summarized and the main problem for long-term application is identified: insufficient seals. A new approach with seal-less drives is proposed and current research on a simplified rotary piston design is presented. Expert commentary: The development of blood pumps focuses mainly on the improvement of rotary pumps. However, medical complications indicate that inherent limitations of this pump type remain and restrict the next substantial step forward in the therapy of heart failure patients. Thus, research on different pump types is reasonable. If the development of reliable drives and bearings succeeds, rotary piston blood pumps become a promising alternative.

High-vacuum plasma pump

International Nuclear Information System (INIS)

Dorodnov, A.M.; Minajchev, V.E.; Miroshkin, S.I.

1980-01-01

The action of an electric-arc high-vacuum pump intended for evacuating the volumes in which the operation processes are followed by a high gas evolution is considered. The operation of the pump is based on the principle of controlling the getter feed according to the gas load and effect of plasma sorbtion pumping. The pump performances are given. The starting pressure is about 5 Pa, the limiting residual pressure is about 5x10 -6 Pa, the pumping out rate of nitrogen in the pressure range 5x10 -5 -5x10 -3 Pa accounts for about 4000 l/s, the power consumption comes to 6 kW. Analyzing the results of the test operation of the pump, it has been concluded that its principal advantages are the high starting pressure, controlled getter feed rate and possibility of pumping out the gases which are usually pumped out with difficulty. The operation reliability of the pump is defined mainly by reliable operation of the ignition system of the vacuum arc [ru

Pumping machinery theory and practice

CERN Document Server

Badr, Hassan M

2014-01-01

Pumping Machinery Theory and Practice comprehensively covers the theoretical foundation and applications of pumping machinery. Key features: Covers characteristics of centrifugal pumps, axial flow pumps and displacement pumpsConsiders pumping machinery performance and operational-type problemsCovers advanced topics in pumping machinery including multiphase flow principles, and two and three-phase flow pumping systemsCovers different methods of flow rate control and relevance to machine efficiency and energy consumptionCovers different methods of flow rate control and relevance to machine effi

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

International Nuclear Information System (INIS)

Holmes, B.J.

1991-03-01

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

Water pressure control device for control rod drive

International Nuclear Information System (INIS)

Sato, Hideyuki.

1981-01-01

Purpose: To minimize the fluctuations in the reactor water level upon occurrence of abnormality by inputting the level signal of the reactor to an arithmetic unit for controlling the pressure of control rod drive water to thereby enable effective reactor level control. Constitution: Signal from a flow rate transmitter is inputted into an arithmetic unit to perform constant flow rate control upon normal operation. While on the other hand, if abnormality occurs such as feedwater pump trips, the arithmetic unit is switched from the constant flow rate control to the reactor water level control. Reactor water level signal is inputted into the arithmetic unit and the control valve is most suitably controlled, whereby water is fed from CST to the reactor by way of control rod drive water system to secure the reactor water level if feedwater to the reactor is interrupted by loss of coolants on the feedwater system. Since this enables to minimize the fluctuations in the reactor water level upon abnormality, the reactor water level can be controlled most suitably by the reactor water level signal. (Moriyama, K.)

Hydraulic optimization of 'S' characteristics of the pump-turbine for Xianju pumped storage plant

International Nuclear Information System (INIS)

Liu, W C; Zheng, J S; Cheng, J; Shi, Q H

2012-01-01

The pump-turbine with a rated power capacity of 375MW each at Xianju pumped storage plant is the most powerful one under construction in China. In order to avoid the instability near no-load conditions, the hydraulic design of the pump-turbine has been optimized to improving the 'S' characteristic in the development of the model pump-turbine. This paper presents the cause of 'S' characteristic of a pump-turbine by CFD simulation of the internal flow. Based on the CFD analysis, the hydraulic design optimization of the pump-turbine was carried out to eliminate the 'S' characteristics of the machine at Xianju pumped storage plant and a big step for removing the 'S' characteristic of a pump-turbine has been obtained. The model test results demonstrate that the pump-turbine designed for Xianju pumped storage plant can smoothly operate near no-load conditions without an addition of misaligned guide vanes.

Device for preventing cooling water from flowing out of reactor

International Nuclear Information System (INIS)

Chinen, Masanori; Kotani, Koichi; Murase, Michio.

1976-01-01

Object: To provide emergency cooling system, which can prevent cooling water bearing radioactivity from flowing to the outside of the reactor at the time of breakage of feedwater pipe, thus eliminating the possibility of exposure of the fuel rod to provide high reliability and also reducing the possibility of causing radioactive pollution. Structure: The device for preventing cooling water from flowing out from the reactor features a jet nozzle inserted in a feedwater pipe adjacent to the inlet or outlet thereof immediately before the reactor container. The nozzle outlet is provided in the vicinity of the reactor wall and in a direction opposite to the direction of out-flow, and water supplied from a high pressure pump is jetted from it. (Nakamura, S.)

ATWS sensitivity studies to support PSA success criteria

International Nuclear Information System (INIS)

Zheng Yaoyao; Xu Zhen; Ke Xiao

2010-01-01

The limiting anticipated transient without scram (ATWS) event is the heatup transient caused by a reduction of heat removal capability by the secondary side of the plant. In order to evaluate the AP1000 plant behavior following an ATWS, loss of normal feedwater ATWS event has been analyzed using the LOFTRAN code. Several sensitivity studies are also performed to address some key issues, such as steam dump capacity, core makeup tank (CMT) characteristic and boron coefficient, reactor coolant pumps (RCPs) availability, startup feedwater system (STS) availability and steam generator (SG) heat flux. The results of the analysis show that in order to mitigate the consequence of such an accident, the steam dump should be isolated and RCP should trip on CMT actuation signal. (authors)

Pump integrated closed loop control. The intelligent actuator pump - state of the art and perspectives; Pumpenintegrierte Prozessregelung. Der intelligente Aktor Pumpe - Stand und Moeglichkeiten

Energy Technology Data Exchange (ETDEWEB)

Paulus, Thomas; Schullerer, Joachim; Oesterle, Manfred [KSB AG, Frankenthal (Germany)

2009-07-01

In many areas of industrial automation, centrifugal pumps and systems of centrifugal pumps are important actuators of a process und therefore a fundamental part of the entire plant. In contrary to the controlled valves, located behind the pump in the pipework system, the functions of the intelligent actuator pump or system of pumps are rarely used and noticed. With regard to the reduction of life cycle costs of the asset pump, pump integrated closed loop control of the fluid transport task has advantages over central closed loop control e.g. in a process control system. This article contains a survey of the intelligent actuator pump, its structure and functions with regard to the solution of the fluid transport task and according pump integrated closed loop control. (orig.)

High-vacuum pumping out of hydrogen isotopes by compressed and electrophysical pumps

International Nuclear Information System (INIS)

Bychkova, A.D.; Ershova, Z.V.; Saksaganskij, G.L.; Serebrennikov, D.V.

1982-01-01

To explain the selection of parameters of vacuum systems of projected thermonuclear devices, experiments are performed on the pumping-out of deuterium and tritium by high-vacuum pumps of different types. The values of the fast response of turbomolecular, diffusion vapour-mercury, magneto-discharge and titanium getter pumps in the operation pressure range are determined. The rate of sorption of hydrogen isotopes by non-spraying gas absorber of cial alloy depending on the amount of the gas absorbed and temperature, is measured. Gas current is determined by the pressure drop on the diagram of the known conductivity. Individual calibration of manometric converters for different gases using a mercury burette is performed preliminarily. The means of high-vacuum pumping-out that have been studied have the following values of fast response for tritium (relatively to protium): turbomolecular pump-0.95; evaporation getter pump-0.25; magneto-discharge pumps-0.65-0.9; cial alloy-0.1...0.5

Measure Guideline: Replacing Single-Speed Pool Pumps with Variable Speed Pumps for Energy Savings

Energy Technology Data Exchange (ETDEWEB)

Hunt, A.; Easley, S.

2012-05-01

The report evaluates potential energy savings by replacing traditional single-speed pool pumps with variable speed pool pumps, and provide a basic cost comparison between continued uses of traditional pumps verses new pumps. A simple step-by-step process for inspecting the pool area and installing a new pool pump follows.

Wet motor geroter fuel pump

Energy Technology Data Exchange (ETDEWEB)

Wiernicki, M.V.

1987-05-05

This patent describes a wet motor gerotor fuel pump for pumping fuel from a fuel source to an internal combustion which consists of: gerotor pump means comprising an inner pump gear, an outer pump gear, and second tang means located on one of the inner and outer pump gears. The second tang means further extends in a second radial direction radially offset from the first radial direction and forms a driving connection with the first tang means such that the fuel pump pumps fuel from the fuel source into the narrow conduit inlet chamber, through the gerotor pump means past the electric motor means into the outlet housing means substantially along the flow axis to the internal combustion engine.

Pump failure leads to alternative vertical pump condition monitoring technique

International Nuclear Information System (INIS)

DeVilliers, Adriaan; Glandon, Kevin

2011-01-01

Condition monitoring and detecting early signs of potential failure mechanisms present particular problems in vertical pumps. Most often, the majority of the pump assembly is not readily accessible for visual or audible inspection or conventional vibration monitoring techniques using accelerometers and/or proximity sensors. The root cause failure analysis of a 2-stage vertical centrifugal service-water pump at a nuclear power generating facility in the USA is presented, highlighting this long standing challenge in condition monitoring of vertical pumps. This paper will summarize the major findings of the root cause analysis (RCA), highlight the limitations of traditional monitoring techniques, and present an expanded application of motor current monitoring as a means to gain insight into the mechanical performance and condition of a pump. The 'real-world' example of failure, monitoring and correlation of the monitoring technique to a detailed pump disassembly inspection is also presented. This paper will explain some of the reasons behind well known design principles requiring natural frequency separation from known forcing frequencies, as well as explore an unexpected submerged brittle fracture failure mechanism, and how such issues may be avoided. (author)

Diode-pumped laser with improved pumping system

Science.gov (United States)

Chang, Jim J.

2004-03-09

A laser wherein pump radiation from laser diodes is delivered to a pump chamber and into the lasing medium by quasi-three-dimensional compound parabolic concentrator light channels. The light channels have reflective side walls with a curved surface and reflective end walls with a curved surface. A flow tube between the lasing medium and the light channel has a roughened surface.

Optically pumped atoms

CERN Document Server

Happer, William; Walker, Thad

2010-01-01

Covering the most important knowledge on optical pumping of atoms, this ready reference is backed by numerous examples of modelling computation for optical pumped systems. The authors show for the first time that modern scientific computing software makes it practical to analyze the full, multilevel system of optically pumped atoms. To make the discussion less abstract, the authors have illustrated key points with sections of MATLAB codes. To make most effective use of contemporary mathematical software, it is especially useful to analyze optical pumping situations in the Liouville spa

Outcome of Cardiac Rehabilitation Following Off-Pump Versus On-Pump Coronary Bypass Surgery

Directory of Open Access Journals (Sweden)

Reza Arefizadeh

2017-05-01

CONCLUSIONS: Regarding QOL and psychological status, there were no differences in the CR outcome between those who underwent off-pump bypass surgery and those who underwent on-pump surgery; nevertheless, the off-pump technique was superior to the on-pump method on METs improvement following CR.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Measure Guideline. Replacing Single-Speed Pool Pumps with Variable Speed Pumps for Energy Savings

Energy Technology Data Exchange (ETDEWEB)

Hunt, A. [Building Media and the Building America Retrofit Alliance (BARA), Wilmington, DE (United States); Easley, S. [Building Media and the Building America Retrofit Alliance (BARA), Wilmington, DE (United States)

2012-05-01

This measure guideline evaluates potential energy savings by replacing traditional single-speed pool pumps with variable speed pool pumps, and provides a basic cost comparison between continued uses of traditional pumps verses new pumps. A simple step-by-step process for inspecting the pool area and installing a new pool pump follows.

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1993-12-01

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

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

International Nuclear Information System (INIS)

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

1991-07-01

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

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

International Nuclear Information System (INIS)

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

1993-08-01

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

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

International Nuclear Information System (INIS)

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

1990-10-01

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

Pumps for nuclear facilities

International Nuclear Information System (INIS)

1999-01-01

The guide describes how the Finnish Radiation and Nuclear Safety Authority (STUK) controls pumps and their motors at nuclear power plants and other nuclear facilities. The scope of the control is determined by the Safety Class of the pump in question. The various phases of the control are: (1) review of construction plan, (2) control of manufacturing, and construction inspection, (3) commissioning inspection, and (4) control during operation. STUK controls Safety Class 1, 2 and 3 pumps at nuclear facilities as described in this guide. STUK inspects Class EYT (non-nuclear) pumps separately or in connection with the commissioning inspections of the systems. This guide gives the control procedure and related requirements primarily for centrifugal pumps. However, it is also applied to the control of piston pumps and other pump types not mentioned in this guide

Detection of pump degradation

International Nuclear Information System (INIS)

Greene, R.H.; Casada, D.A.; Ayers, C.W.

1995-08-01

This Phase II Nuclear Plant Aging Research study examines the methods of detecting pump degradation that are currently employed in domestic and overseas nuclear facilities. This report evaluates the criteria mandated by required pump testing at U.S. nuclear power plants and compares them to those features characteristic of state-of-the-art diagnostic programs and practices currently implemented by other major industries. Since the working condition of the pump driver is crucial to pump operability, a brief review of new applications of motor diagnostics is provided that highlights recent developments in this technology. The routine collection and analysis of spectral data is superior to all other technologies in its ability to accurately detect numerous types and causes of pump degradation. Existing ASME Code testing criteria do not require the evaluation of pump vibration spectra but instead overall vibration amplitude. The mechanical information discernible from vibration amplitude analysis is limited, and several cases of pump failure were not detected in their early stages by vibration monitoring. Since spectral analysis can provide a wealth of pertinent information concerning the mechanical condition of rotating machinery, its incorporation into ASME testing criteria could merit a relaxation in the monthly-to-quarterly testing schedules that seek to verify and assure pump operability. Pump drivers are not included in the current battery of testing. Operational problems thought to be caused by pump degradation were found to be the result of motor degradation. Recent advances in nonintrusive monitoring techniques have made motor diagnostics a viable technology for assessing motor operability. Motor current/power analysis can detect rotor bar degradation and ascertain ranges of hydraulically unstable operation for a particular pump and motor set. The concept of using motor current or power fluctuations as an indicator of pump hydraulic load stability is presented

Detection of pump degradation

Energy Technology Data Exchange (ETDEWEB)

Greene, R.H.; Casada, D.A.; Ayers, C.W. [and others

1995-08-01

This Phase II Nuclear Plant Aging Research study examines the methods of detecting pump degradation that are currently employed in domestic and overseas nuclear facilities. This report evaluates the criteria mandated by required pump testing at U.S. nuclear power plants and compares them to those features characteristic of state-of-the-art diagnostic programs and practices currently implemented by other major industries. Since the working condition of the pump driver is crucial to pump operability, a brief review of new applications of motor diagnostics is provided that highlights recent developments in this technology. The routine collection and analysis of spectral data is superior to all other technologies in its ability to accurately detect numerous types and causes of pump degradation. Existing ASME Code testing criteria do not require the evaluation of pump vibration spectra but instead overall vibration amplitude. The mechanical information discernible from vibration amplitude analysis is limited, and several cases of pump failure were not detected in their early stages by vibration monitoring. Since spectral analysis can provide a wealth of pertinent information concerning the mechanical condition of rotating machinery, its incorporation into ASME testing criteria could merit a relaxation in the monthly-to-quarterly testing schedules that seek to verify and assure pump operability. Pump drivers are not included in the current battery of testing. Operational problems thought to be caused by pump degradation were found to be the result of motor degradation. Recent advances in nonintrusive monitoring techniques have made motor diagnostics a viable technology for assessing motor operability. Motor current/power analysis can detect rotor bar degradation and ascertain ranges of hydraulically unstable operation for a particular pump and motor set. The concept of using motor current or power fluctuations as an indicator of pump hydraulic load stability is presented.

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

International Nuclear Information System (INIS)

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

1988-01-01

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

Survey of pumps for tritium gas

International Nuclear Information System (INIS)

Dowell, T.M.

1983-05-01

This report considers many different types of pumps for their possible use in pumping tritium gas in the low, intermediate and high vacuum ranges. No one type of pump is suitable for use over the wide range of pumping pressure required in a typical pumping system. The favoured components for such a system are: bellows pump (low vacuum); orbiting scroll pump (intermediate vacuum); magnetically suspended turbomolecular pump (high vacuum); cryopump (high vacuum). Other pumps which should be considered for possible future development are: mound modified vane pump; SRTI wobble pump; roots pump with canned motor. It is proposed that a study be made of a future tritium pumping system in a Canadian tritium facility, e.g. a tritium laboratory

TRACE/PARCS modelling of rips trip transients for Lungmen ABWR

Energy Technology Data Exchange (ETDEWEB)

Chang, C. Y. [Inst. of Nuclear Engineering and Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Lin, H. T.; Wang, J. R. [Inst. of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Shih, C. [Inst. of Nuclear Engineering and Science, Dept. of Engineering and System Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China)

2012-07-01

The objectives of this study are to examine the performances of the steady-state results calculated by the Lungmen TRACE/PARCS model compared to SIMULATE-3 code, as well as to use the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen TRACE/PARCS model. In this study, three power generation methods in TRACE were utilized to analyze the three reactor internal pumps (RIPs) trip transient for the purpose of validating the TRACE/PARCS model. In general, the comparisons show that the transient responses of key system parameters agree well with the FSAR results, including core power, core inlet flow, reactivity, etc. Further studies will be performed in the future using Lungmen TRACE/PARCS model. After the commercial operation of Lungmen nuclear power plant, TRACE/PARCS model will be verified. (authors)

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

International Nuclear Information System (INIS)

Takeda, Takeshi

2016-07-01

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

Vertical pump assembly

International Nuclear Information System (INIS)

Dohnal, M.; Rosel, J.; Skarka, V.

1988-01-01

The mounting is described of the drive assembly of a vertical pump for nuclear power plants in areas with seismic risk. The assembly is attached to the building floor using flexible and damping elements. The design allows producing seismically resistant pumps without major design changes in the existing types of vertical pumps. (E.S.). 1 fig

Normetex Pump Alternatives Study

International Nuclear Information System (INIS)

Clark, Elliot A.

2013-01-01

A mainstay pump for tritium systems, the Normetex scroll pump, is currently unavailable because the Normetex company went out of business. This pump was an all-metal scroll pump that served tritium processing facilities very well. Current tritium system operators are evaluating replacement pumps for the Normetex pump and for general used in tritium service. An all-metal equivalent alternative to the Normetex pump has not yet been identified. 1. The ideal replacement tritium pump would be hermetically sealed and contain no polymer components or oils. Polymers and oils degrade over time when they contact ionizing radiation. 2. Halogenated polymers (containing fluorine, chlorine, or both) and oils are commonly found in pumps. These materials have many properties that surpass those of hydrocarbon-based polymers and oils, including thermal stability (higher operating temperature) and better chemical resistance. Unfortunately, they are less resistant to degradation from ionizing radiation than hydrocarbon-based materials (in general). 3. Polymers and oils can form gaseous, condensable (HF, TF), liquid, and solid species when exposed to ionizing radiation. For example, halogenated polymers form HF and HCl, which are extremely corrosive upon reaction with water. If a pump containing polymers or oils must be used in a tritium system, the system must be designed to be able to process the unwanted by-products. Design features to mitigate degradation products include filters and chemical or physical traps (eg. cold traps, oil traps). 4. Polymer components can work in tritium systems, but must be replaced regularly. Polymer components performance should be monitored or be regularly tested, and regular replacement of components should be viewed as an expected normal event. A radioactive waste stream must be established to dispose of used polymer components and oil with an approved disposal plan developed based on the facility location and its regulators. Polymers have varying

Performance of Wind Pump Prototype

African Journals Online (AJOL)

Mulu

Mekelle University, Mekelle, Ethiopia (*mul_at@yahoo.com). ABSTRACT. A wind ... balanced rotor power and reciprocating pump, hence did not consider the effect of pump size. ... Keywords: Wind pump, Windmill, Performance testing, Pump efficiency, Pump discharge, ... Unfortunately, in rural places, where the houses are.

Development of model pump for establishing hydraulic design of primary sodium pumps in PFBR

International Nuclear Information System (INIS)

Chougule, R.J.; Sahasrabudhe, H.G.; Rao, A.S.L.K.; Balchander, K.; Kale, R.D.

1994-01-01

Indira Gandhi Centre for Atomic Research, Kalpakkam indicated requirement of indigenous development of primary sodium pump, handling liquid sodium as coolant in Fast Breeder Reactor. The primary sodium pump concept selected in its preliminary design is a vertical, single stage, with single suction impeller, suction facing downwards. The pump is having diffuser, discharge casing and discharge collector. The 1/3 rd size model pump is developed to establish the hydraulic performance of the prototype primary sodium pump. The main objectives were to verify the hydraulic design to operate on low net positive suction head available (NPSHA), no evidence of visible cavitation at available NPSHA, the pump should be designed with a diffuser etc. The model pump PSP 250/40 was designed and successfully developed by Research and Development Division of M/s Kirloskar Brothers Ltd., Kirloskarvadi. The performance testing using model pump was successfully carried out on a closed circuit test rig. The performance of a model pump at three different speeds 1900 rpm, 1456 rpm and 975 rpm was established. The values of hydraulic axial thrust with and without balancing holes on impeller at 1900 rpm was measured. Visual cavitation study at 1900 rpm was carried out to establish the NPSH at bubble free operation of the pump. The tested performance of the model pump is converted to the full scale prototype pump. The predicted performance of prototype pump at 700 rpm was found to be meeting fully with the expected duties. (author). 6 figs., 3 tabs

Solar Pumping : The Basics

OpenAIRE

World Bank Group

2018-01-01

Solar photovoltaic water pumping (SWP) uses energy from solar photovoltaic (PV) panels to power an electric water pump. The entire process, from sunlight to stored energy, is elegant and simple. Over last seven years, the technology and price of solar pumping have evolved dramatically and hence the opportunities it presents. Solar pumping is most competitive in regions with high solar inso...

Electrokinetic pumps and actuators

International Nuclear Information System (INIS)

Phillip M. Paul

2000-01-01

Flow and ionic transport in porous media are central to electrokinetic pumping as well as to a host of other microfluidic devices. Electrokinetic pumping provides the ability to create high pressures (to over 10,000 psi) and high flow rates (over 1 mL/min) with a device having no moving parts and all liquid seals. The electrokinetic pump (EKP) is ideally suited for applications ranging from a high pressure integrated pump for chip-scale HPLC to a high flow rate integrated pump for forced liquid convection cooling of high-power electronics. Relations for flow rate and current fluxes in porous media are derived that provide a basis for analysis of complex microfluidic systems as well as for optimization of electrokinetic pumps

Electrokinetic pumps and actuators

Energy Technology Data Exchange (ETDEWEB)

Phillip M. Paul

2000-03-01

Flow and ionic transport in porous media are central to electrokinetic pumping as well as to a host of other microfluidic devices. Electrokinetic pumping provides the ability to create high pressures (to over 10,000 psi) and high flow rates (over 1 mL/min) with a device having no moving parts and all liquid seals. The electrokinetic pump (EKP) is ideally suited for applications ranging from a high pressure integrated pump for chip-scale HPLC to a high flow rate integrated pump for forced liquid convection cooling of high-power electronics. Relations for flow rate and current fluxes in porous media are derived that provide a basis for analysis of complex microfluidic systems as well as for optimization of electrokinetic pumps.

Off-pump versus on-pump coronary artery bypass grafting for ischaemic heart disease

DEFF Research Database (Denmark)

Møller, Christian H; Penninga, Luit; Wetterslev, Jørn

2012-01-01

Coronary artery bypass grafting (CABG) is performed both without and with cardiopulmonary bypass, referred to as off-pump and on-pump CABG respectively. However, the preferable technique is unclear.......Coronary artery bypass grafting (CABG) is performed both without and with cardiopulmonary bypass, referred to as off-pump and on-pump CABG respectively. However, the preferable technique is unclear....

Internal pump monitoring device

International Nuclear Information System (INIS)

Kurosaki, Toshikazu.

1996-01-01

In the present invention, a thermometer is disposed at the upper end of an internal pump casing of a coolant recycling system in a BWR type reactor to detect leakage of reactor water thereby ensuring the improvement of reliability of the internal pump. Namely, a thermometer is disposed, which can detect temperature elevation occurred when water in the internal pump leaked from a reactor pressure vessel passes through the gap between a stretch tube and an upper end of the pump casing. Signals from the thermometer are transmitted to a signal processing device by an instrumentation cable. The signal processing device generates an alarm when the temperature signal exceeds a predetermined value and announces that leakage of reactor water occurs in the internal pump. Since the present invention can detect the leakage of the reactor water in the pump casing in an early stage, it can contribute to the improvement of the safety and reliability of the internal pump. (I.S.)

Improvement of centrifugal pump performance through addition of splitter blades on impeller pump

Science.gov (United States)

Kurniawan, Krisna Eka; Santoso, Budi; Tjahjana, Dominicus Danardono Dwi Prija

2018-02-01

The workable way to improve pump performance is to redesign or modify the impellers of centrifugal pump. The purpose of impeller pump modification is to improve pump efficiency, reduce cross flow, reduce secondary incidence flows, and decrease backflow areas at impeller outlets. Number blades and splitter blades in the impeller are three. The outlet blade angle is 20°, and the rotating speed of impeller is 2400 rpm. The added splitter blades variations are 0.25, 0.375, and 0.5 of the original blade length. The splitter blade placements are on the outer side of the impeller. The addition of splitter blades on the outer side of the impeller with 0.5L increases the pump head until 22% and the pump has 38.66% hydraulic efficiency. The best efficiency point of water flow rate pump (Qbep) was 3.02 × 10-3 m3/s.