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

International Nuclear Information System (INIS)

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

1990-01-01

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

Trace analysis of loss of feedwater flow event in Lungmen ABWR

International Nuclear Information System (INIS)

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

2009-01-01

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

Analysis of steam generator loss-of-feedwater experiments with APROS and RELAP5/MOD3.1 computer codes

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

Three experiments were conducted to study the behavior of the new horizontal steam generator construction of the PACTEL test facility. In the experiments the secondary side coolant level was reduced stepwise. The experiments were calculated with two computer codes RELAP5/MOD3.1 and APROS version 2.11. A similar nodalization scheme was used for both codes to that the results may be compared. Only the steam generator was modelled and the rest of the facility was given as a boundary condition. The results show that both codes calculate well the behaviour of the primary side of the steam generator. On the secondary side both codes calculate lower steam temperatures in the upper part of the heat exchange tube bundle than was measured in the experiments.

Analysis of steam generator loss-of-feedwater experiments with APROS and RELAP5/MOD3.1 computer codes

International Nuclear Information System (INIS)

Virtanen, E.; Haapalehto, T.; Kouhia, J.

1997-01-01

Three experiments were conducted to study the behaviour of the new horizontal steam generator construction of the PACTEL test facility. In the experiments the secondary side coolant level was reduced stepwise. The experiments were calculated with two computer codes RELAP5/MOD3.1 and APROS version 2.11. A similar nodalization scheme was used for both codes so that the results may be compared. Only the steam generator was modeled and the rest of the facility was given as a boundary condition. The results show that both codes calculate well the behaviour of the primary side of the steam generator. On the secondary side both codes calculate lower steam temperatures in the upper part of the heat exchange tube bundle than was measured in the experiments. (orig.)

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

ISP 33. OECD/NEA/CSNI International Standard Problem n. 33. Pactel natural circulation stepwise coolant inventory reduction experiment. Comparison report. Volume 1 + 2

International Nuclear Information System (INIS)

Purhonen, H.; Kouhia, J.; Holmstrom, H.

1994-12-01

This is the comparison report of the CSNI ISP n.33, which is based on a natural circulation experiment with various coolant inventories conducted in Pactel facility (Finland), a 1/305 volumetrically scaled, full-height simulator of a Russian type VVER-440 pressurized water reactor. It presents all submitted blind calculational results from different countries, using various codes (Athlet, Cathare2, etc.) and compares them with the experimental data. The Pactel facility and the ISP 33 experiment are described, and the summaries of the participants, the computer codes and the nodalizations used for the blind calculations are presented

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.

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

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

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

PACTEL ISP-33. RELAP5 assessment

International Nuclear Information System (INIS)

Siccama, N.B.; Roodbergen, H.A.

1995-08-01

This report presents the results of the calculation of the PACTEL ISP-33 experiment as obtained from the RELAP5 code. The main goal of the ECN contribution to this ISP was to assess RELAP5 on one- and two-phase natural circulation phenomena which occur in Eastern European VVER plants in case of LOCA conditions. Different natural circulation modes were calculated in the simulation of the ISP-33 experiment. The single-phase liquid flow, the steady two-phase flow, and the boiler-condenser single-phase heat removal are calculated well by the RELAP5 code. The phenomena in the transient two-phase flow are difficult to simulate. (orig.)

Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

Energy Technology Data Exchange (ETDEWEB)

Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

1995-09-01

There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

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

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

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)

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.

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

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.

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

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)

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)

Thermal-hydraulic studies on the safety of VVER-440 type nuclear power plants

International Nuclear Information System (INIS)

Tuunanen, J.

1994-01-01

The thesis includes several thermal-hydraulic analyses related to the Loviisa VVER-440 nuclear power plant units. The work consists of experimental studies, analysis of the experiments, analysis of some plant transients and development of a calculational model for calculation of boric concentrations in the reactor. In the first part of thesis, in the case of simulation of boric acid solution behaviour during long-term cooling period of LOCAs, experiments were performed in scaled-down test facilities. The experimental data together with the results of RELAP5/MOD3 simulations were used to develop a model for calculations of boric acid concentrations in the reactor during LOCAs. In the second part, in the case of simulation of horizontal generators, experiments were performed with PACTEL integral test loop to simulate loss of feedwater transients. The PACTEL experiments as well as earlier REWETT-III natural circulation tests, were analyzed with RELAP5/MOD3 Version 5m5 code. The third part of the work consists of simulations of Loviisa VVER reactor pump trip transients with RELAP5/MOD1-Eur, RELAP5/MOD3 and CATHARE codes. (56 refs., 9 figs.)

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)

LOFT/LP-FW-1, Loss of Fluid Test, PWR Response to Loss-of-Feedwater Transient

International Nuclear Information System (INIS)

1989-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The first OECD LOFT experiment was conducted on February 20, 1983. It was designed to evaluate the generic PWR system response during a complete loss-of-feedwater transient. The objective of the experiment was to investigate the performance of primary 'feed and bleed' using a 'bleed' from the PORV and 'feed' from the HPIS to provide decay heat removal and system pressure reduction while maintaining the primary coolant inventory. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

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

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

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)

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)

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)

RELAP5 analysis of PACTEL injection tests

International Nuclear Information System (INIS)

Kimber, G.R.; Lillington, J.N.

2000-01-01

A characteristic feature of advanced reactor designs is their reliance on passive safety systems. It is important to assess both the operation of such systems and the ability of systems codes, such as RELAP5, to model them. In Finland VTT Energy, together with Lappeenranta University of Technology, is using the PACTEL facility for the investigation of passive core cooling systems. In particular, a core make-up tank (CMT) has been installed in the rig to operate in a similar manner to those in many Advanced PWR designs. Three small break tests, GDE-24, GDE-34 and GDE-43 in the PACTEL facility were chosen for modelling with RELAP5. The objective of GDE-24 was to investigate CMT behaviour and in particular the effects of condensation in the CMT. The second test, GDE-34, was similar except that it had a smaller CMT and at the start of the test the water in the CMT and connecting pipework was at an elevated temperature. Test GDE-43 focused on conditions when the driving force for flow through the passive system injection system (PSIS) slowly disappears. Analysis of all tests reported here was carried out with RELAP5/MOD 3.2.1.2. The paper summarises the conclusions of all the tests. A critical part of the study revolved around modelling of the CMT. A model was developed to allow its detailed behaviour to be investigated more easily. This enabled recommendations for improving the condensation modelling in RELAP5 to be made. Apart from the wall condensation modelling issue, the implication of the work is that RELAP5/MOD 3.2.1.2 (a comparatively recent version of the code) is broadly adequate for these applications. (author)

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)

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

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

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.)

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.)

Reactor feedwater pump control device

International Nuclear Information System (INIS)

Nishiyama, Hiroyuki.

1990-01-01

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

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

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.)

Post test analysis of the LOBI BT17 experiment

International Nuclear Information System (INIS)

Farkas, I.T.; Hozer, Z.; Takacs, A.

1994-12-01

The LOBI experimental facility and the BT17 experiment. This experiment represents a loss-of-feedwater transient with feed and bleed procedure. The computational analysis has been performed by the CATHARE thermal hydraulic system code. The results of calculations are in satisfactory agreement with the experimental values. A comparison has been made with a Loss-of-Feedwater test performed on the PMK-2 facility. (author). 16 refs., 22 figs., 5 tabs

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)

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.)

Feedwater system in a nuclear power plant

International Nuclear Information System (INIS)

Shimizu, Tadayuki.

1975-01-01

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

Reactor feedwater facility

Energy Technology Data Exchange (ETDEWEB)

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

1996-04-30

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

ISP33 standard problem on the PACTEL facility

Energy Technology Data Exchange (ETDEWEB)

Purhonen, H.; Kouhia, J. [VTT Energy, Lappeenranta (Finland); Kalli, H. [Lappeenranta Univ. of Technology (Finland)

1995-09-01

ISP33 is the first OECD/NEA/CSNI standard problem related to VVER type of pressurized water reactors. The reference reactor of the PACTEL test facility, which was used to carry out the ISP33 experiment, is the VVER-440 reactor, two of which are located near the Finnish city of Loviisa. The objective of the ISP33 test was to study the natural circulation behaviour of VVER-440 reactors at different coolant inventories. Natural circulation was considered as a suitable phenomenon to focus on by the first VVER related ISP due to its importance in most accidents and transients. The behaviour of the natural circulation was expected to be different compared to Western type of PWRs as a result of the effect of horizontal steam generators and the hot leg loop seals. This ISP was conducted as a blind problem. The experiment was started at full coolant inventory. Single-phase natural circulation transported the energy from the core to the steam generators. The inventory was then reduced stepwise at about 900 s intervals draining 60 kg each time from the bottom of the downcomer. the core power was about 3.7% of the nominal value. The test was terminated after the cladding temperatures began to rise. ATHLET, CATHARE, RELAP5 (MODs 3, 2.5 and 2), RELAP4/MOD6, DINAMIKA and TECH-M4 codes were used in 21 pre- and 20 posttest calculations submitted for the ISP33.

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

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

Preliminary design of RDE feedwater pump impeller

International Nuclear Information System (INIS)

Sri Sudadiyo

2018-01-01

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

Analysis of results from a loss-of-offsite-power-initiated ATWS experiment in the LOFT Facility

International Nuclear Information System (INIS)

Varacalle, D.J.; Giri, A.M.; Koizumi, Y.; Koske, J.E.

1983-01-01

An anticipated transient without scram (ATWS), initiated by loss-of-offsite power, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a scaled safety relief valve (SRV) representative of those in a commercial PWR, while reactor power was reduced by moderator reactivity feedback in a natural circulation mode. The experiment showed that reactor power decreases more rapidly when the primary pumps are tripped in a loss-of-offsite-power ATWS than in a loss-of-feedwater induced ATWS when the primary pumps are left on. During the experiment, the SRV had sufficient relief capacity to control primary system pressure. Natural circulation was effective in removing core heat at high temperature, pressure, and core power. The system transient response predicted using the RELAPS/MOD1 computer code showed good agreement with the experimental data

Simulation of long-term cooling in the VVER-640 power plant after a large break LOKA on the PACTEL facility

International Nuclear Information System (INIS)

Banati, J.

2000-01-01

The present report gives a short introduction to the safety features of the new Russian VVER-640 reactor design. In order to analyze the complex thermal hydraulic phenomena during long-term cooling after a large-break LOCA, experiments will be carried out in the PACTEL facility. For preparation, pre-test calculations were performed using the RELAPS/MOD3.2 computer code. The main part of the report discusses the results obtained by the program. The structure and options used in the input deck, as well as the efforts of code application to the simulation of proposed experiments are reviewed. A short sensitivity study is provided on the calculated results. Finally, conclusions are drawn for the code capabilities to represent the expectable trends in the upcoming tests. (orig.)

49 CFR 230.57 - Injectors and feedwater pumps.

Science.gov (United States)

2010-10-01

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

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

Feedwater device for nuclear power plant

International Nuclear Information System (INIS)

Ikekita, Iwao.

1980-01-01

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

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)

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

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.)

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.

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)

Experiment data report for LOFT anticipated transient-without-scram Experiment L9-3

International Nuclear Information System (INIS)

Bayless, P.D.; Divine, J.M.

1982-05-01

Selected pertinent and uninterpreted data from the third anticipated transient with multiple failures experiment (Experiment L9-3) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large [approx. 1000 MW(e)], commercial PWR operations. Experiment L9-3 simulated a loss-of-feedwater anticipated transient without scram. The loss-of-feedwater accident led to an increase in the primary coolant system temperature and pressure. Both the experiment power-operated relief valve (PORV) and safety relief valve opened and were able to limit and control the pressure transient. The plant was then recovered with the control rods still withdrawn by injecting 7200-ppM borated water, manually cycling the PORV and feeding and bleeding the steam generator

Anticipated transient without SCRAM experiments at LOFT

International Nuclear Information System (INIS)

Grush, W.H.; Harvego, E.A.; Koizumi, Y.; Varacalle, D.J.

1983-01-01

This paper discusses the experimental results for two anticipated transients without scram (ATWS) experiments, and compares computer code predictions with the experimental data. Experiment L9-3 simulated an ATWS in a commercial pressurized water reactor (PWR) initiated by a complete loss of feedwater and Experiment L9-4 simulated a loss-of-offsite-power-initiated (loss of feedwater and trip of the primary coolant pumps) ATWS. The LOFT facility is uniquely suited for ATWS experiments because it is a volumetrically scaled (1/44) experimental PWR designed to simulate the major components and system responses of larger commercial PWRs during both hypothesized loss-of-coolant accidents and anticipated transients. In both of the examined experiments, the primary system transient behavior was dominated by the interactions between the steam generator primary-to-secondary heat removal, the reactor kinetics, and the relief valve actuation. It is demonstrated that the discussed ATWS events can be controlled by properly sized automatic safety systems

A Quick Review on Steam Generator Water Level Tracking Methods and Its Uncertainties

Energy Technology Data Exchange (ETDEWEB)

Park, Ki Moon; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-05-15

The tracking of the SG water level is important for maintaining the heat removal of the reactor and the power plant safety. In addition, the SG water level is important for the reactor trip and the actuation of SG back-up feedwater system as well. The SG water level is mainly controlled by the Feed Water Control System (FWCS) during either normal operation or transients therefore, the selection of the FWCS control parameters is also important. In this paper, methods of SG water level calculation are first reviewed and future works to perform sensitivity study of the SG water level calculation with a system analysis code will be identified. This is partially shown from Loss-of-feedwater experiments carried out in PACTEL and the LOF-10 experiment. The experiment was chosen to test the modeling capabilities of TRACE code for VVER SG. The experiment measured the water level with the pressure differential and the code calculated the water level directly from the code results. In this paper, three previously suggested parameters which can be used as an indicator of the SG water level are briefly introduced: (1) downcomer collapsed water level, (2) water mass inventory and (3) pressure differential. From the review of previous works, it was identified that most of the system analysis code calculates the SG water level directly by using the downcomer collapsed level. In contrast, the pressure difference is measured as used for the SG water level tracking in a real nuclear power plant or experiment.

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)

Generic evaluation of feedwater transients and small break loss-of-coolant accidents in GE-designed operating plants and near-term operating license applications

International Nuclear Information System (INIS)

1980-01-01

The results are presented of a generic evaluation of feedwater transients, small-break loss-of-coolant accidents (LOCAs), and other TMI-2-related events for General Electric Company (GE)-designed operating plants and near-term operating license applications to confirm or establish the bases for the continued safe operation of the operating plants. 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. Additional review of the accident is continuing and further information is being obtained and evaluated. Any new information will be reviewed and modifications will be made as appropriate

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

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)

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

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

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

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.

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

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

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

International Nuclear Information System (INIS)

Anon.

1986-01-01

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

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)

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

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%

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

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

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

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

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)

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.

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

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

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.

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

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

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.

SBLOCA AND LOFW EXPERIMENTS IN A SCALED-DOWN IET FACILITY OF REX-10 REACTOR

Directory of Open Access Journals (Sweden)

YEON-GUN LEE

2013-06-01

Full Text Available This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA and the loss-of-feedwater accident (LOFW in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF, a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

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)

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.

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

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)

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)

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)

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

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.

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

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.

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

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)

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

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

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

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

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

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

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

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

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

thickness is around one hundred nanometers, then it reaches rapidly one micrometer after three hundred hours. The last tests show an important increase of the thickness. The following of iron release in solution confirms this result. During the first three hundred hours the release is linear but with a low slope, then an acceleration is observed from three hundred to nine hundred hours. The flow and the morphology of the oxide are in accordance with a generalized corrosion phenomenon. The addition base metal loss and the oxide formed on the surface confirm a parabolic rate characteristic of generalized corrosion during the first hours. Then, the rate of corrosion follows a linear law representative of flow accelerated corrosion. These tests show a corrosion of carbon steel in several steps in the feedwater system conditions, a general corrosion during three hundred hours and then a more important corrosion. In order to identify each phenomenon, additional tests and characterizations (porosity measurements) will be carried out. (authors)

Passive safety injection experiments and analyses (PAHKO)

International Nuclear Information System (INIS)

Tuunanen, J.

1998-01-01

PAHKO project involved experiments on the PACTEL facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines (Pressure Balancing Line, PBL, and Injection Line, IL). The examined PSIS worked efficiently in SBLOCAs although the flow through the PSIS stopped temporarily if the break was very small and the hot water filled the CMT. The experiments demonstrated the importance of the flow distributor in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable to simulate the overall behaviour of the transients. The detailed analyses of the results showed some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of the PSIS phenomena. (orig.)

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

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)

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

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.

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

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

Experiment data report for LOFT anticipated transient without scram Experiment L9-4

International Nuclear Information System (INIS)

Batt, D.L.; Divine, J.M.; McKenna, K.J.

1982-11-01

Selected pertinent and uninterpreted data from the fourth anticipated transient with multiple failures experiment (Experiment L9-4) conducted on September 24, 1982, in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system's thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large [approx. 1000 MW(e)], commercial PWR operations. Experiment L9-4 simulated a loss-of-offsite-power anticipated transient without reactor scram. The loss-of-offsite-power accident led to an increase in the primary coolant system temperature and pressure. The experiment safety relief valve opened and was able to limit and control the pressure transient. In addition, subsequent heat generation was dissipated by the auxiliary feedwater flow in the secondary coolant system until the reactor was scrammed at experiment termination

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

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

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

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

Aging and low-flow degradation of auxilary feedwater pumps

International Nuclear Information System (INIS)

Adams, M.L.

1992-01-01

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

Aging and low-flow degradation of auxiliary feedwater pumps

International Nuclear Information System (INIS)

Adams, M.L.

1991-01-01

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

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

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

Analysis of ATLAS FLB-EC6 Experiment using SPACE Code

International Nuclear Information System (INIS)

Lee, Donghyuk; Kim, Yohan; Kim, Seyun

2013-01-01

The new code is named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). As a part of code validation effort, simulation of ATLAS FLB(Feedwater Line Break) experiment using SPACE code has been performed. The FLB-EC6 experiment is economizer break of a main feedwater line. The calculated results using the SPACE code are compared with those from the experiment. The ATLAS FLB-EC6 experiment, which is economizer feedwater line break, was simulated using the SPACE code. The calculated results were compared with those from the experiment. The comparisons of break flow rate and steam generator water level show good agreement with the experiment. The SPACE code is capable of predicting physical phenomena occurring during ATLAS FLB-EC6 experiment

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

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

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)

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

International Nuclear Information System (INIS)

No, Young Gyu; Seong, Poong Hyun

2015-01-01

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

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

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

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

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

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)

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)

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.).

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

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)

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

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)

Calculations for the prediction of accident situation

International Nuclear Information System (INIS)

Perneczky, L.; Szabados, L.; Toth, I.

1987-11-01

The report deals with the analysis of the loss of feedwater transient assuming that the pressurizer safety valve remains open throughout the process. The scenario is the Paks NPP counterpart of the loss of feedwater test performed on the PMK-NVH facility. The initial and boundary conditions of the calculation agree with those of the experiment allowing direct comparison with experimental results. (author) 18 refs.; 34 figs

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.

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

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.

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

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

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)

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.

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.).

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)

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

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

A comparison of the RELAP5/MOD3 code with the IIST natural circulation experiments

International Nuclear Information System (INIS)

Ferng, Y.M.; Lee, C.H.

1995-01-01

A series of experiments dealing with variable secondary-side cooling conditions have been conducted at the IIST facility, including the natural circulation experiments under the secondary-side conditions of normal feedwater, loss of feedwater, and full of air. Different cooling conditions at the secondary side directly affect the primary-to-secondary heat transfer and then may influence the heat removal capability of natural circulation in the primary system. The corresponding analytical work is performed using the RELAP5/MOD3 code. Good agreement is reached both qualitatively and quantitatively between the experimental data and calculated results, demonstrating the satisfactory assessment of RELAP5/MOD3 code compared with the IIST natural circulation experiments. The cooling conditions at the secondary side have no significant effect on the heat removal capability of natural circulation as long as sufficient coolant exists on the steam generator secondary side, based on current IIST data and analytical results. Continuous increase of the core temperature and system pressure is also demonstrated experimentally and analytically in the test with the secondary side dry for the sake of deficient heat transfer capability at the steam generator secondary system

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)

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

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)

Investigation of surface oxide morphology in SG feedwater pipes and study of its influence on flow accelerated corrosion rate

International Nuclear Information System (INIS)

Qiu, G.; Alos-Ramos, O.; Monchecourt, D.; Mansour, C.; Delaunay, S.; Trevin, S.

2015-01-01

Flow accelerated corrosion (FAC) affects carbon steel components in the secondary circuits of PWR plants. The mandatory use of the prediction tool BRT-CICERO in all its PWR plants enables EDF to perform efficient inspections programs and minimize the number of leaks in the secondary circuits. Due to the operating conditions, SG feedwater flow regulation (ARE) circuits can be affected by FAC phenomenon. Thickness loss has been reported by several plants during the last 10 years, although significant damage by FAC remains very rare. This paper describes the surface features observed on an ARE straight tube that has orange peel pattern with thickness loss on the one half of its inner surface and a thick fouling layer without much thickness loss on the other. An analysis of the oxide porosity and structure by SEM investigation has been carried out. The origin of fouling layer and its behavior in the ARE circuits environment (oxide solubility, flow stability/turbulence) have been discussed. Finally by comparing with the classic FAC models, an attempt of correlation between the presence of the fouling layer and the lower corrosion rate is proposed. (authors)

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

The weight-loss experience: a qualitative exploration

Directory of Open Access Journals (Sweden)

David Rogerson

2016-05-01

Full Text Available Abstract Background Long-term weight management consists of weight-loss, weight-loss maintenance, and weight-gain stages. Qualitative insights into weight management are now appearing in the literature however research appears to be biased towards explorations of weight-loss maintenance. The qualitative understanding of weight loss, which begets weight-loss maintenance and might establish the experiences and behaviours necessary for successful long-term weight management, is comparatively under-investigated. The aim of this study was to investigate the weight-loss experiences of a sample of participants not aligned to clinical intervention research, in order to understand the weight-loss experiences of a naturalistic sample. Methods Participants (n = 8 with weight-loss (n = 4 and weight-maintenance experiences (n = 4 were interviewed using a semi-structured interview to understand the weight-loss experience. Interview data was analysed thematically using Framework Analysis and was underpinned by realist meta-theory. Results Weight loss was experienced as an enduring challenge, where factors that assisted weight loss were developed and experienced dichotomously to factors that hindered it. Participants described barriers to (dichotomous thinking, environments, social pressures and weight centeredness and facilitators of (mindfulness, knowledge, exercise, readiness to change, structure, self-monitoring and social support their weight-loss goals in rich detail, highlighting that weight loss was a complex experience. Conclusions Weight loss was a difficult task, with physical, social, behavioural and environmental elements that appeared to assist and inhibit weight-loss efforts concurrently. Health professionals might need to better understand the day-to-day challenges of dieters in order to provide more effective, tailored treatments. Future research should look to investigate the psycho-social consequences of weight-loss dieting, in

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.)

Assessment of System Behavior and Actions Under Loss of Electric Power For CANDU

Energy Technology Data Exchange (ETDEWEB)

Kang, San Ha; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

2014-05-15

For the analysis, the CANDU-6 plant in Korea is considered and only the passive components are operable. The other systems are assumed to be at failed condition due to the loss of electric power. At this accident, only the inventories remained in the primary heat transport system (PHTS) and steam generator can be used for the decay heat removal. Due to the transfer of decay heat, the inventory of steam generator secondary side is discharged to the air through passive operation of main steam safety valves (MSSVs). After the steam generators are dried, the PHTS is over-pressurized and the coolant is discharged to fuelling machine vault through passive operation of degasser condenser tank relief valves (DCRVs). Under this situation, the maintenance of the integrity of PHTS is important for the protection of radionuclides release to the environment. Thus, deterministic analysis using CATHENA code is carried out for the simulation of the accident and the appropriate operator action is considered. The loss of electric power results in the depletion of steam generator inventory which is necessary for the decay heat removal. If only the passive system is credited, the PT can be failed after the steam generator is depleted. For the prevention of the PT failure, the feedwater should be supplied to the steam generator before 4,800s after the accident. The feedwater can be supplied using water in dousing tank if the steam generators are depressurized. The decay heat from the core is removed through natural circulation if the feedwater can be supplied continuously.

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

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

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 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

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)

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

About the complete loss of functions assumed by redundant systems

International Nuclear Information System (INIS)

Boaretto, Y.; Cayol, A.; Fourest, M.; Guimbail, H.

1980-04-01

Are to be taken into account situations resulting from loss of redundant safety systems. Two ways of approach were to be probed: evaluation of the failure probability and analysis of the consequences of those situations. The first way leads to improve reliability of concerned systems, the second way to set up mitigating means. Before TMI-2 occured, safety advices had already been issued about three kinds of situations: anticipated transients without scram, loss of ultimate heat sink, simultaneous loss of out-and inside power supplies. That, in some cases, something had to be done to improve safety showed the rightness of the concern. Next step is the study of the loss of both normal and emergency feedwater: The regulatory request has been issued on September 1979

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

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.

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

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

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

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

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)

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

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...

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

LOFT/L9-3, Loss of Fluid Test, Anticipated Transients with Multiple Failures

International Nuclear Information System (INIS)

1992-01-01

1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the third of the NRC L9 series of experiments on Anticipated Transients with Multiple Failures. Loss-of-feedwater effects were studied. The experiment was conducted on 7 April 1982

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.

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.

Computational modeling and analysis of heavy water losses in boiler blow down with different positions of BBW-V100 at KANUPP

International Nuclear Information System (INIS)

Maqbool, M. U.

2012-01-01

The term blowdown is referred to the boilers and steam generators. Blowing down water from the steam generators maintains the chemistry of the feedwater and helps prevent scaling or sludge formation. In a nuclear power plant, the primary loop contains some activity in the form of tritium content. In boilers, primary and secondary systems interface and due to the pressure difference there is always a chance of mixing of primary and secondary fluids in event of tube leak. This primary fluid i.e., heavy water in our case can be lost through the blowdown lines after mixing with the feedwater. This thesis is a computational work for the determination of heavy water losses through the blowdown lines. (author)

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

Application of the integrated analysis of safety (ISA) to sequences of Total loss of feed water in a PWR Reactor

International Nuclear Information System (INIS)

Moreno Chamorro, P.; Gallego Diaz, C.

2011-01-01

The main objective of this work is to show the current status of the implementation of integrated analysis of safety (ISA) methodology and its SCAIS associated tool (system of simulation codes for ISA) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

Influence of the CVCS Modelling on Results of the Loss of Offsite Power (LOOP) Safety Analysis for NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Bajs, T.; Debrecin, N.

2006-01-01

A Loss of Offsite Power (LOOP) transient scenario is based on a complete loss of non-emergency AC power that results in the loss of all power to the plant auxiliaries, i.e., the Reactor Coolant Pumps (RCPs), condensate pumps, etc. An actual LOOP event would cause a loss of all feedwater, a loss of forced Reactor Coolant System (RCS) flow and a reactor trip within less than 2 seconds as a result of either loss of power to the rod cluster assembly gripper coils or any RCS flow trips. For safety analysis purposes the LOOP event is conservatively modelled as a Loss of Normal Feedwater (LONF) transient with a subsequent loss of offsite power as a result of a reactor trip. The reactor trip followed by RCP trip are delayed until a low-low Steam Generator (SG) level signal is reached. This is a more conservative scenario than the LOOP event because the least amount of SG secondary side water mass available for heat removal and the increased amount of the stored energy in the primary circuit at the time of the loss of RCS flow result. The standard LOOP safety analysis is aimed to demonstrate the natural circulation capability of the RCS to remove residual and decay heat from the core aided by Auxiliary Feedwater in the secondary system. In addition to this goal the presented work is aimed to resolve the potential safety issue resulting from the influence of the Chemical and Volume Control System (CVCS) operation during LOOP event for NPP Krsko. The potential safety concern for the LOOP analysis is that the loss of instrument air system may occur thus leading to the CVCS charging and letdown flow imbalance. A net RCS inventory addition may result with water solid pressurizer condition. Water discharge through the pressurizer relief and safety valves could lead to overpressurization of the Pressurizer Relief Tank (PRT) and rupture of the PRT rupture disks. Additional concern is that pressurizer relief and safety valves may fail to properly reseat when exposed to water relief

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

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.

Diagnosis of Feedwater Heater Performance Degradation using Fuzzy Approach

International Nuclear Information System (INIS)

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

2014-01-01

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

Diagnosis of Feedwater Heater Performance Degradation using Fuzzy Approach

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

'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)

Alpha particle loss in the TFTR DT experiments

International Nuclear Information System (INIS)

Zweben, S.J.; Darrow, D.S.; Herrmann, H.W.

1995-01-01

Alpha particle loss was measured during the TFTR DT experiments using a scintillator detector located at the vessel bottom in the ion grad-B drift direction. The DT alpha particle loss to this detector was consistent with the calculated first-orbit loss over the whole range of plasma current I=0.6-2.7 MA. In particular, the alpha particle loss rate per DT neutron did not increase significantly with fusion power up to 10.7 MW, indicating the absence of any new ''collective'' alpha particle loss processes in these experiments

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)

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

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

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.)

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

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.

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

selected, out of them each covers a group of ana-logical processes: operating mode with feedwater loss; operating mode with station black-out; operating mode during an earthquake

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

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

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)

A study on emergency response guideline during the loss of steam generator secondary heat sink in pressurizer water reactor

International Nuclear Information System (INIS)

Yoon, D. J.; Lee, J. Y.; Song, D. S.

1999-01-01

A loss of secondary heat sink can occur as a result of several different initiating events, which are a loss of main feedwater during power operation, a loss of off-site power, or any other scenario for which main feedwater is isolated or lost. At this point the opening and closing of the PORV or safety valves will result in a loss of RCS inventory similar in nature to a small break loss of coolant accident. If operator action is not taken, the pressurizer PORV or safety valves will continue to cycle open and closed at the valve setpoint pressure removing RCS inventory and a limited amount of core decay heat until eventually enough inventory will be lost to result in core uncovery. We conclude that a requirement to successfully initiate bleed and feed on steam generator dryout, without any significant core uncovery expected to occur, is that the PORV flow to power ratio must exceed 140 (lbm/hr)/Mwt. For all plants whose PORV capacity is less than 140 (lbm/hr)/Mwt, since symptoms of SG dryout cannot be used to initiate bleed and feed, increasing RCS pressure and temperature or pressure greater than 2335 psig cannot be used. The only alternative symptom available is SG narrow range level. Since Kori 1,2,3 and 4' PORV capacity is more than the criteria, the bleed and feed operation can be initiated at steam generator dryout

Experiment data report for LOFT large-break loss-of-coolant experiment L2-5

International Nuclear Information System (INIS)

Bayless, P.D.; Divine, J.M.

1982-08-01

Selected pertinent and uninterpreted data from the third nuclear large break loss-of-coolant experiment (Experiment L2-5) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large [approx. 1000 MW(e)] commercial PWR operations. Experiment L2-5 simulated a double-ended offset shear of a cold leg in the primary coolant system. The primary coolant pumps were tripped within 1 s after the break initiation, simulating a loss of site power. Consistent with the loss of power, the starting of the high- and low-pressure injection systems was delayed. The peak fuel rod cladding temperature achieved was 1078 +- 13 K. The emergency core cooling system re-covered the core and quenched the cladding. No evidence of core damage was detected

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)

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.

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

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

A comparative simulation of feed and bleed operation during the total loss of feedwater event by RELAP5/MOD3 and CEFLASH-4AS/REM computer codes

International Nuclear Information System (INIS)

Kwon, Y.M.; Ro, T.S.; Song, J.H.

1995-01-01

The Ulchin 3 and 4 nuclear power plants, which are two-loop 2,825 MW(thermal) pressurized water reactors designed by the Korea Atomic Energy Research Institute, adopted a safety depressurization system (SDS) to mitigate the beyond-design-basis event of a total loss of feedwater (TLOFW). A comparative simulation by the CEFLASH-4AS/REM and RELAP5/MOD3 computer codes for the TLOFW event without operator recovery and the TLOFW event with feed and bleed (F and B) operation is performed for Ulchin 3 and 4. In the analyses, the SDS bleed paths are modeled by orifices located on the top of the pressurizer, where the analytical area of the bleed path is based on the Ulchin 3 and 4 SDS design flow capacity. An additional case, where the SDS piping and valves are modeled explicitly, is considered for the RELAP5 analysis. The predictions by the CEFLASH-4AS/REM of the transient two-phase system behavior show good qualitative and quantitative agreement with those by the RELAP5 simulation. The RELAP5 case with explicit piping results in less repressurization and lower reactor coolant system pressure than that of the case without explicit SDS modeling. However, the two cases of RELAP5 analyses result in essentially the same transient scenarios for TLOFW with F and B operation. The results of the simulation demonstrate the validity of the Ulchin 3 and 4 design approach, which employs CEFLASH-4AS/REM computer code and SDS bleed paths modeled by orifices located on the top of the pressurizer. The results also indicate that the decay heat removal and core inventory makeup function can be successfully accomplished by F and B operation by using the SDS for Ulchin 3 and 4

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

Loss of preferred power events in German BWRs and PWRs

International Nuclear Information System (INIS)

Frisch, W.

1984-01-01

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

Loss of Power and Water Hammer Event at San Onofre, Unit 1, on November 21, 1985

International Nuclear Information System (INIS)

1986-01-01

On November 21, 1985, Southern California Edison's Onofre Nuclear Generating Station, Unit 1, located south of San Clemente, California, experienced a partial loss of inplant ac electrical power while the plant was operating at 60% power. Following a manual reactor trip, the plant lost all inplant ac power for 4 minutes and experienced a severe incidence of water hammer in the feedwater system which caused a leak, damaged plant equipment, and challenged the integrity of the plant's heat sink. The most significant aspect of the event involved the failure of five safety-related check valves in the feed-water system whose failure occurred in less than year, without detection, and jeopardized the integrity of safety systems. The event involved a number of equipment malfunctions, operator errors, and procedural deficiencies. This report documents the findings and conclusions of an NRC Incident Investigation Team sent to San Onofre by the NRC Executive Director for Operations in conformance with NRC's recently established Incident Investigation Program

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

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)

Thermal-hydraulic experiment for safe and stable operation of a PIUS-type reactor

International Nuclear Information System (INIS)

Tasaka, K.; Imai, S.; Masaoka, H.; Irianto, I.D.; Kohketsu, H.; Tamaki, M.; Anoda, Y.; Murata, H.; Kukita, Y.

1992-01-01

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 normal operation simulation and a loss of feedwater test for an accident condition simulation, using a small scale atmospheric pressure test loop which simulated the PIUS principle. (author)

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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.

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)

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

Elastic-plastic response of a piping system due to simulated double-ended guillotine break events

International Nuclear Information System (INIS)

Kussmaul, K.; Diem, H.; Hunger, H.; Katzenmeier, G.

1987-01-01

From the blowdown experiments performed on the HDR feedwater line with feedwater check valve the conclusion can be drawn that high transient loads of up to plastic strains of 3%, acting on an initially integer piping system, can be sustained without loss of integrity for a low number of load cycles due to the plasticizing capacity of the pipework materials nowadays used in reactor technology. In the experiments carried out with ferritic piping of ND 400 pressure peaks up to about 31,5 MPa were achieved which resulted in excessive strains of up to 3%. By nonlinear finite element computations (ABAQUS) it was possible to describe the elastic-plastic behaviour of the piping in a good approximation. (orig./GL)

Experiences of time loss among videogame players: an empirical study.

Science.gov (United States)

Wood, Richard T A; Griffiths, Mark D; Parke, Adrian

2007-02-01

Playing videogames is now a major leisure pursuit, yet research in the area is comparatively sparse. Previous correlational evidence suggests that subjective time loss occurs during playing videogames. This study examined experiences of time loss among a relatively large group of gamers (n = 280). Quantitative and qualitative data were collected through an online survey. Results showed that time loss occurred irrespective of gender, age, or frequency of play, but was associated with particular structural characteristics of games such as their complexity, the presence of multi-levels, missions and/or high scores, multiplayer interactions, and plot. Results also demonstrated that time loss could have both positive and negative outcomes for players. Positive aspects of time loss included helping players to relax and temporarily escape from reality. Negative aspects included the sacrificing of other things in their lives, guilty feelings about wasted time, and social conflict. It is concluded that for many gamers, losing track of time is a positive experience and is one of the main reasons for playing videogames.

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

Thermal-Hydraulic Experiment To Test The Stable Operation Of A PIUS Type Reactor

International Nuclear Information System (INIS)

Irianto, Djoko; Kanji, T.; Kukita, Y.

1996-01-01

An advanced type of reaktor concept as the Process Inherent Ultimate Safety (PIUS) reactor was based on intrinsically passive safety considerations. The stable operation of a PIUS type reactor is based on the automation of circulation pump speed. An automatic circulation pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed the PIUS-type reactor. In principle this control system maintains the fluid temperature at the axial center of the lower density lock at average of the fluid temperatures below and above the lower density lock. This control system will 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 using atmospheric pressure thermal-hydraulic test loop which simulated the PIUS principle. The experiments such as: start-up and power ramping tests for normal operation simulation and loss of feedwater test for an accident condition simulation, carried out in JAERI

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)

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)

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.

Cultural conflicts in the weight loss experience of overweight Latinos.

Science.gov (United States)

Diaz, V A; Mainous, A G; Pope, C

2007-02-01

In spite of the high prevalence of obesity in the Latino population, there is limited recent information that can be used by health-care providers to develop culturally appropriate weight loss strategies for this population. Therefore, we describe weight loss experiences, attitudes and barriers in overweight Latino adults. Qualitative study using focus group methodology. Twenty-one overweight adults (body mass index >/=25, age >/=20 years) self-identified as Latinos. Subjects participated in one of three focus groups. Reccurring themes within group discussions were identified by three independent investigators, one who was ethnicity concordant. Themes included the presence of mixed messages when determining one's appropriate weight, with participants' desire to lose weight to be healthy (based on professional advice and personal experience) conflicting with the cultural idea that being overweight is healthy. Participants described discordance when adapting to the mainstream, leading to the loss of healthy traditional habits. Participants expressed interest in weight loss and familiarity with dieting and weight loss interventions. They desired culturally appropriate nutrition education and reassurance regarding healthy dieting from health-care providers. The importance of interactions with peers during education was another relevant theme, and participants were overwhelmingly positive about group education. To improve health promotion for Latinos, cultural factors distinctive to this underserved population, and barriers they articulate, should be considered when developing weight loss interventions.

Weight loss experiences of obese perimenopausal women with metabolic syndrome.

Science.gov (United States)

Su, Mei-Chen; Lin, Hung-Ru; Chu, Nain-Feng; Huang, Chih-Hsung; Tsao, Lee-Ing

2015-07-01

To develop a descriptive theory for the weight loss experiences of obese perimenopausal women with metabolic syndrome. Obesity and metabolic syndrome both pose a threat to the health of perimenopausal women; therefore, understanding perimenopausal women's subjective feelings and experiences is beneficial to establishing effective prevention strategies. However, studies have rarely explored these relevant experiences. A qualitative study using the grounded theory method to establish a descriptive theory. Eighteen obese perimenopausal women with metabolic syndrome aged 45-60 years participated in comprehensive interviews. 'Crossing the gaps to making life modifications' was the core category, and 'the awareness of weight gain and health alarm' was the antecedent condition. In the weight loss experience, the following three interaction categories were identified: (1) 'experiencing bad feelings,' (2) 'encountering obstacles' and (3) 'making efforts to transition to a new life.' Some women adhered to new life habits through perceiving social support and by using self-incentives. Finally, women enjoyed and mastered self-monitoring of their health in their new life, and practiced new changes as part of their life. However, some participants felt that making changes to their life was too time-consuming. Therefore, these women chose to live with their abnormal health without making changes. Obese perimenopausal women with metabolic syndrome experienced various gaps in their weight loss process. Although they struggled with many obstacles, these women were able to learn from their experiences and face their health challenges. These findings can guide healthcare professionals to provide appropriate interventions to understand the hidden health problems of this particular group of women. Healthcare professionals should develop a set of plans by which women receive a complete weight loss program and support from professionals and family. © 2015 John Wiley & Sons Ltd.

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 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 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

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)

The Emotional Experience of School Change: Resistance, Loss, and Grief.

Science.gov (United States)

Marshak, David

1996-01-01

Ignoring the emotional experience of school change may unintentionally sabotage rational planning. Reinventing schools means attending to educators' emotional experience, particularly their expectations, sense of loss, and resulting grief. School norms must be transformed so that teachers and administrators can have meaningful conversations about…

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.)

Experience and optimisation of ethanolamine treatment for a PWR secondary system

International Nuclear Information System (INIS)

Eeden, Nestor-van; Matthee, Felix; Montshiwagae, Maleke

2012-09-01

concentration. In 2008, ETA dosing control was automated, which improved the ability to control the concentration in the feedwater. A significant reduction in corrosion product transport (CPT) on both KNPS units is clearly attributable to ETA implementation. This is demonstrated by feedwater iron measurements and also by the mass of material removed by steam generator sludge lancing. This paper discusses 14 years of operating experience under the ETA Chemistry Regime. The rationale that led to the evolution of the chemistry specifications is described. Information is presented on the effects of the various specification changes. ETA influence on SG chemistry, effect on demineralizers and corrosion product monitoring. Operation with ETA has exposed some undesirable effects on ion exchange resins that can lead to deterioration of secondary chemistry. These effects that were at first not fully appreciated are now better understood. At KNPS this operating experience is being used to optimise the secondary chemistry to achieve the maximum advantage of FAC and CPT mitigation while also minimising impurity intrusion. (authors)

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.)

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)

Screening Criteria for Loss of Room Cooling Failure

Energy Technology Data Exchange (ETDEWEB)

Hwang, Mee Jeong; Yang, Joon Eon; Yoon, Churl

2007-01-15

In this report, we estimated the temperature of the pump rooms and reviewed the operability of the components under the loss of the HVAC (Heating, Ventilation, and Air Condition) system. The issues relevant to the HVAC system in the PSA (Probabilistic Safety Assessment) FT (Fault Tree) model are as follows: (1) Does the loss of the HVAC system bring about a function failure of other components? (2) Can the operator take action to reduce the temperature of the room in case of a HVAC function failure? At present, we do not know whether a component will lose its function or not under the loss of the HVAC. ASME Standard describes that a recovery action can be credited if the related recovery action is included in the procedure or there are similar recovery experiences in the plant. However, there is no description about the recovery action of the HVAC in the EOP (Emergency Operation Procedure) of the UCN3, 4 under the situation of a loss of the HVAC. Even though we consider this assumption positively, it would be limited to the rooms such as the Switchgear Room, Inverter Room, and Main Control Room etc. where a real recovery action can be performed easily. However, if we consider the HVAC failure in the PSA FT model according to the above background, the problem is that the unavailability induced from the loss of a HVAC is highly unrealistically. From a viewpoint of the PSA, it is not true that the related system always fails even though the HVAC system fails. Therefore, we reviewed the necessity of the HVAC model through the identification of the operable temperature of the components' within the pump room and the change of the temperature of the pump room under the situation of a loss of the HVAC system. In this paper, we performed a heat up calculation for the Auxiliary Feedwater Motor Operated Pump (AFW MDP) room, PAB-077-11A with CFX 10 and RATT when the HVAC system is failed. We also reviewed the operability of the components under a loss of the HVAC. Room

Screening Criteria for Loss of Room Cooling Failure

International Nuclear Information System (INIS)

Hwang, Mee Jeong; Yang, Joon Eon; Yoon, Churl

2007-01-01

In this report, we estimated the temperature of the pump rooms and reviewed the operability of the components under the loss of the HVAC (Heating, Ventilation, and Air Condition) system. The issues relevant to the HVAC system in the PSA (Probabilistic Safety Assessment) FT (Fault Tree) model are as follows: (1) Does the loss of the HVAC system bring about a function failure of other components? (2) Can the operator take action to reduce the temperature of the room in case of a HVAC function failure? At present, we do not know whether a component will lose its function or not under the loss of the HVAC. ASME Standard describes that a recovery action can be credited if the related recovery action is included in the procedure or there are similar recovery experiences in the plant. However, there is no description about the recovery action of the HVAC in the EOP (Emergency Operation Procedure) of the UCN3, 4 under the situation of a loss of the HVAC. Even though we consider this assumption positively, it would be limited to the rooms such as the Switchgear Room, Inverter Room, and Main Control Room etc. where a real recovery action can be performed easily. However, if we consider the HVAC failure in the PSA FT model according to the above background, the problem is that the unavailability induced from the loss of a HVAC is highly unrealistically. From a viewpoint of the PSA, it is not true that the related system always fails even though the HVAC system fails. Therefore, we reviewed the necessity of the HVAC model through the identification of the operable temperature of the components' within the pump room and the change of the temperature of the pump room under the situation of a loss of the HVAC system. In this paper, we performed a heat up calculation for the Auxiliary Feedwater Motor Operated Pump (AFW MDP) room, PAB-077-11A with CFX 10 and RATT when the HVAC system is failed. We also reviewed the operability of the components under a loss of the HVAC. Room

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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.)

Grief Interrupted: The Experience of Loss Among Incarcerated Women

OpenAIRE

Harner, Holly M.; Hentz, Patricia M.; Evangelista, Maria Carmela

2010-01-01

Incarcerated women face a number of stressors apart from the actual incarceration. Nearly half of all women in prison experience the death of a loved one during their incarceration. Our purpose for this study was to explore the experience of grief and loss among incarcerated women using a phenomenological method. Our study approach followed van Manen's method of phenomenology and Munhall's description of existential lifeworlds. Our analysis revealed four existential lifeworlds: temporality: f...

Dealing with death: medical students' experiences with patient loss.

Science.gov (United States)

Pessagno, Regina; Foote, Carrie E; Aponte, Robert

This article explores medical students' experiences and coping strategies when confronting patient loss in their 3rd and 4th years of their programs. Much of the literature on the impact of patient losses focuses on physicians. This article joins a handful of works aimed at how medical students experience and cope with patient loss. In-depth interviews with 20 medical students provided rich descriptions of their varying experiences coping with death. Consistent with previous work, students experience substantial emotional stress coping with patient deaths, though some were more difficult to bear than others, such as when the dying patient was a child or when treatment errors could have contributed to deaths. Common coping mechanisms included talking through their emotions, thrusting themselves into continuing their rounds, crying, participating in infant death rituals, and turning to religion. When deaths occurred, senior personnel who exhibited empathy toward the deceased and tolerance toward the students' emotional responses were lauded and made the process easier. Also emotionally daunting, in many instances, was dealing with the families of dying patients. Most of the students did not view death as a failure, contrary to much earlier literature, except in instances in which human error or decision making may have played a part in causing the death of a patient.

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)

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

Experience theory, or How desserts are like losses.

Science.gov (United States)

Martin, Jolie M; Reimann, Martin; Norton, Michael I

2016-11-01

Although many experiments have explored risk preferences for money, few have systematically assessed risk preferences for everyday experiences. We propose a conceptual model and provide convergent evidence from 7 experiments to suggest that, in contrast to a typical "zero" reference point for choices on money, reference points for choices of experiences are set at more extreme outcomes, leading to concave utility for negative experiences but convex utility for positive experiences. As a result, people are more risk-averse for negative experiences such as disgusting foods-as for monetary gains-but more risk-seeking for positive experiences such as desserts-as for monetary losses. These risk preferences for experiences are robust to different methods of elicitation. (PsycINFO Database Record (c) 2016 APA, all rights reserved).

Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

Energy Technology Data Exchange (ETDEWEB)

Moreno Chamorro, P.; Gallego Diaz, C.

2011-07-01

The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

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)

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)

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

International Nuclear Information System (INIS)

1992-07-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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.

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

Calculation of the time behavior of a PWR NPP during a loss of feedwater ATWS case

International Nuclear Information System (INIS)

Hoeld, A.

1988-01-01

Event tree analyses of plant internal accidents play an important role within the safety evaluations of nuclear power reactors. The consequences after normal and abnormal operational perturbations have to be studied with respect to the safety situation of the entire plant and the possibility of additional failures in the reactor scram system be taken into account. In the analysis of anticipated transients with or without reactor scram (non-ATWS or ATWS-cases), it can, according to their initiating events, be distinguished between three important categories, namely - loss of off-site and on-site power (LOOP), - turbine-trip without opening of the bypass station, - loss of main feed water (LOFW). The last case with the additional assumption of a failure in the control rod drive will be subject of this presentation, calculating the dynamic behavior of a PWR NPP (with an end of cycle core, EOC) after such a LOFW/ATWS accident by the transient code combination ALMOD-4/UTSG-2. A short characterization of this combination will be given before consequences of such an accident and the interactions of the different plant parameters are discussed in more detail on basis of the corresponding calculation

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

International Nuclear Information System (INIS)

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

1993-02-01

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

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

Marginal motherhood: the ambiguous experience of pregnancy-loss in Cameroon

NARCIS (Netherlands)

van der Sijpt, E.

2007-01-01

There has been much international and scholarly attention for, on the one hand, ‘overpopulation’ or ‘high fertility rates’; and, on the other hand, experiences of infertility. Little light has been shed, however, on the marginal ‘in-between’ situation of women experiencing reproductive loss, that

LOFT advanced densitometer for nuclear loss-of-coolant experiments

International Nuclear Information System (INIS)

Johnson, L.O.; Lassahn, G.D.; Wood, D.B.

1979-01-01

A ''nuclear hardened'' gamma densitometer, a device which uses radiation attenuation to measure fluid density in the presence of a background radiation field, is described. Data from the nuclear hardened gamma densitometer are acquired by time sampling the coolant fluid piping and fluid attenuated source energy spectrum. The data are used to calculate transient coolant fluid cross sectional average density to analyze transient mass flow and other thermal-hydraulic characteristics during the Loss-of-Fluid Test (LOFT) loss-of-coolant experiments. The nuclear hardened gamma densitometer uses a pulse height analysis or energy discrimination, pulse counting technique which makes separation of the gamma radiation source signal from the reactor generated gamma radiation background noise signal possible by processing discrete pulses which retain their pulse amplitude information

Simulation of total loss of feed water in ATLAS test facility using SPACE code

Energy Technology Data Exchange (ETDEWEB)

Kim, Minhee; Kim, Seyun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of). Central Research Inst.

2017-08-15

A total loss of feedwater (TLOFW) with additional failures in ATLAS test facility was analyzed using SPACE code, which is an advanced thermal-hydraulic system analysis code developed by the Korea nuclear industry. Partial failure of the safety injection pumps (SIPs) and the pilot-operated safety relief valves (POSRVs) of pressurizer were selected as additional failures. In order to assess the capability of SPACE code, partial failure was modeled, and compared with results of OECD-ATLAS A3.1 results. Reasonably good agreement with major thermal-hydraulic parameters was obtained by analyzing the transient behavior. From the results, this indicated that SPACE code has capabilities to design extension conditions, and feed and bleed operation using POSRVs and SIPs were effective for RCS cooling capability during TLOFW.

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

ROSA/LSTF experiment report for RUN SB-CL-24 repeated core heatup phenomena during 0.5% cold leg break LOCA

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Mitsuhiro; Anoda, Yoshinari [Department of Reactor Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

2000-03-01

A small break loss-of-coolant accident (SBLOCA) in a Westinghouse-type four-loop PWR was simulated in an experiment (SB-CL-24) conducted at the Large-Scale Test Facility (LSTF) with an intention to study repeated core heatup during a long-term cooldown process. The experiment was conducted on February 28, 1990 with specified test conditions including failure assumptions both on the high pressure injection (HPI) and the auxiliary feedwater systems, and the intentional secondary system depressurization as an operator action. The secondary depressurization contributed to promote the primary depressurization and the actuation of accumulator injection system (AIS). A temporary core heatup was observed in each of three loopseal clearing (LSC) processes. A significant core heatup occurred in the following boil-off process after loss of the secondary coolant mass and the AIS termination due to increase of the primary pressure. By additional opening of the pressurizer relief valves and safety valves, the primary pressure rapidly decreased to result in the low pressure injection (LPI) which cooled the heated core. This report summarizes results of the experiment (SB-CL-24) in addition to typical responses of some accident indication systems including the core exit thermocouples (CETs) and the water level meters in the primary system. (author)

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

International Nuclear Information System (INIS)

Dai Anguo; Tang Jiahuan; Qian Huifu; Gao Zhikang

1987-12-01

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

Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments

Science.gov (United States)

Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride

The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

LOFT fuel module structural response during loss-of-coolant experiments

International Nuclear Information System (INIS)

Saffell, B.F. Jr.; Selcho, H.S.

1979-01-01

The structural response of the reactor fuel modules installed in the Loss-of-Fluid Test (LOFT) facility have been analyzed for subcooled blowdown loading conditions associated with loss-of-coolant experiments (LOCE). Three independent analyses using the WHAM, SHOCK, and SAP computer codes have been interfaced to calculate the transient mechanical behavior of the LOFT fuel. Test data from two LOCEs indicate the analysis method is conservative. Structural integrity of the fuel modules has been assessed by monitoring guide tube temperatures and control rod drop times during the LOCEs. The analysis and experimental test data indicate the fuel module structural integrity will be maintained for the duration of the LOFT experimental program

RELAP5/MOD2 analysis of LOFT Experiment L9-3

International Nuclear Information System (INIS)

Birchley, J.C.

1992-04-01

An analysis has been performed of LOFT Experiment L9-3, a loss-of-feedwater anticipated transient without trip, in order to support the validation of RELAP5/MOD2. Experiment L9-3 exhibited a rapid boildown of the steam generator, following the loss of feed, with the reactor remaining close to its initial power until the steam generator tubes became sufficiently uncovered for primary to secondary heat transfer to be significantly reduced. The ensuing heat up of the primary fluid resulted in a reduction in power induced by the moderator feedback. The primary system pressure increased to the safety relief valve setpoint, before the fall in reactor power allowed the mismatch between primary system heat input and heat removal via the steam generator to be accommodated by cycling of the pilot operated relief valve (PORV). Comparison between calculation and data shows generally good agreement, though with discrepancies in some areas. Weaknesses in the code's treatment of interphase drag and in the representation of the pressuriser spray are indicated, although a shortage of definitive data, particularly in the steam generator, may also be a factor. The overprediction of interphase drag led to a tendency to underpredict the initial inventory in the steam generator and also, perhaps, to overpredict the steam generator heat transfer while the tubes were being uncovered. There is indication that the pressuriser vapour region conditions were close to equilibrium during spray operation. The point kinetics model in RELAP5/MOD2 proved a viable means of representing the power history for this transient

Elastic-plastic response of a piping system due to simulated double-ended guillotine break events

International Nuclear Information System (INIS)

Kussmaul, K.; Diem, H.; Hunger, H.; Katzenmeier, G.

1987-01-01

From the blowdown experiments performed on the HDR feedwater line with feedwater check valve the conclusion can be drawn that high transient loads of up to plastic strains of 3%, acting on an initially integer piping system, can be sustained without loss of integrity for a low number of load cycles due to the plasticizing capacity of the pipework materials nowadays used in the reactor technology. In the experiments carried out with ferritic piping of ND 400 pressure peaks up to about 31,5 mPA were achieved which resulted in excessive strains of up to 3%. By nonlinear finite element computations (ABAQUS) it was possible to describe the elastic-plastic behaviour of the piping in a good approximation. On account of the safety margins proved in the experiments, potential inaccuracies in theoretical structure analyses are recommended so as to be on the safe side. On the other hand, it appears that designing pipework with reference to elastic stress categories does not adequately take into account the actual reserves of the pipework material

Core heatup prediction during SB LOCA with RELAP5/MOD3.2.2 Gamma

International Nuclear Information System (INIS)

Parzer, I.; Mavko, B.; Petelin, S.

2001-01-01

The paper focuses on the phenomena leading to core uncovering and heatup during the SB LOCA and the ability of RELAP5/MOD3.2.2 Gamma to predict core overheating. The code prediction has been compared to the three experiments, one conducted on the separate effect test facility NEPTUN in Switzerland and the other two conducted on two integral test facilities, PMK-2 in Hungary and PACTEL facility in Finland. In the case of a series of boiloff experiments performed on the NEPTUN test facility the influence of the two correlations available in MOD3.2.2 Gamma for determining interphase drag has been studied. In the case of IAEA-SPE-4 experiment simulation on PMK-2 facility the main goal of the analysis was to study the adequate modeling of the hexagonal core channel with 19-rod bundle and the phenomena during the core uncovering. The third analyzed experiment, OECD-ISP-33, was performed on PACTEL facility to study different natural circulation modes during SB LOCA. The analysis also focused on the final stage of this SB LOCA experiment, when core dryout and heatup was observed due to gradual emptying of the primary system. Following the experience the appropriate modeling options have been used to achieve better representation of the important phenomena during the SB LOCA.(author)

Experiences of adult patients hearing loss postlingually with Cochlear Implant

Directory of Open Access Journals (Sweden)

Teresa María Lizcano Tejado

2013-09-01

Full Text Available Hearing loss is a significant public health problem. The incidence is difficult to establish because of the lack of data in people under age three, but is estimated about 1 per thousand for severe and profound hearing loss.A cochlear implant (CI is a device that converts sounds into electrical energy that triggers a sensation of hearing. The IC is indicated in patients with severe bilateral sensorineural hearing loss with null or poor benefit use of hearing aids.The general objective of this project is to understand the experiences of adult patients with severe-profound sensorineural hearing loss with IC postlingually throughout the implementation process.A personal vision of those implemented will allow us to learn how to face the possibility to hear and interact with their environment, applying this information to improve health care provided to them and identifying those areas where such assistance should be improved. Also allow us to compare the initial expectations and have been achieved, creating realistic expectations for future candidates.For its development we have designed a qualitative study, based on the principles and procedures of grounded theory, semistructured interviews, participant observation and discussion groups.The data will be analyzed using the software Nudist ViVo 9.

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)

Grief Interrupted: The Experience of Loss Among Incarcerated Women

Science.gov (United States)

Harner, Holly M.; Hentz, Patricia M.; Evangelista, Maria Carmela

2011-01-01

Incarcerated women face a number of stressors apart from the actual incarceration. Nearly half of all women in prison experience the death of a loved one during their incarceration. Our purpose for this study was to explore the experience of grief and loss among incarcerated women using a phenomenological method. Our study approach followed van Manen's method of phenomenology and Munhall's description of existential lifeworlds. Our analysis revealed four existential lifeworlds: temporality: frozen in time; spatiality: no place, no space to grieve; corporeality: buried emotions; and relationality: never alone, yet feeling so lonely. The findings generated from this study can help mental health providers as well as correctional professionals develop policies and programs that facilitate the grief process of incarcerated women within the confines of imprisonment. PMID:20581074

Beam loss studies on silicon strip detector modules for the CMS experiment

CERN Document Server

Fahrer, Manuel

2006-01-01

The large beam energy of the LHC demands for a save beam abort system. Nevertheless, failures cannot be excluded with last assurance and are predicted to occur once per year. As the CMS experiment is placed in the neighboured LHC octant, it is affected by such events. The effect of an unsynchronized beam abort on the silicon strip modules of the CMS tracking detector has been investigated in this thesis by performing one accelerator and two lab experiments. The dynamical behaviour of operational parameters of modules and components has been recorded during simulated beam loss events to be able to disentangle the reasons of possible damages. The first study with high intensive proton bunches at the CERN PS ensured the robustness of the module design against beam losses. A further lab experiment with pulsed IR LEDs clarified the physical and electrical processes during such events. The silicon strip sensors on a module are protected against beam losses by a part of the module design that originally has not been...

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

Operating Experiences of a Loss of Voltage Monitoring Program

Energy Technology Data Exchange (ETDEWEB)

Lee, Eun-Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

2015-10-15

Loss of voltage (LOV) events continue to occur due to inadequate work management and random human errors. On February 26, 2015, regulators analyzed the root causes of LOV events and presented the results for the nuclear industry. Currently, KHNP uses a risk monitoring program, which is named 'LOV Monitor', for LOV prevention during pilot plant outages. This review introduces the operation experiences of LOV Monitor based on the evaluation results of a real event. The operation experiences of LOV Monitor in the pilot plants confirmed that this program could detect and reduce LOV possibilities from scheduling errors such as the simultaneous maintenance of energized trains and de-energized trains considering the physical conditions of the power circuit breakers. However, a maintenance culture that heeds the risk monitoring result must be strengthened in order to obtain substantial effects through applying LOV Monitor to the outage.

Cavitation preventing device in a nuclear power plant