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Sample records for integrity steam generator

  1. Steam generator tube integrity program

    International Nuclear Information System (INIS)

    Dierks, D.R.; Shack, W.J.; Muscara, J.

    1996-01-01

    A new research program on steam generator tubing degradation is being sponsored by the U.S. Nuclear Regulatory Commission (NRC) at Argonne National Laboratory. This program is intended to support a performance-based steam generator tube integrity rule. Critical areas addressed by the program include evaluation of the processes used for the in-service inspection of steam generator tubes and recommendations for improving the reliability and accuracy of inspections; validation and improvement of correlations for evaluating integrity and leakage of degraded steam generator tubes, and validation and improvement of correlations and models for predicting degradation in steam generator tubes as aging occurs. The studies will focus on mill-annealed Alloy 600 tubing, however, tests will also be performed on replacement materials such as thermally-treated Alloy 600 or 690. An overview of the technical work planned for the program is given

  2. Steam-Generator Integrity Program/Steam-Generator Group Project

    International Nuclear Information System (INIS)

    1982-10-01

    The Steam Generator Integrity Program (SGIP) is a comprehensive effort addressing issues of nondestructive test (NDT) reliability, inservice inspection (ISI) requirements, and tube plugging criteria for PWR steam generators. In addition, the program has interactive research tasks relating primary side decontamination, secondary side cleaning, and proposed repair techniques to nondestructive inspectability and primary system integrity. The program has acquired a service degraded PWR steam generator for research purposes. This past year a research facility, the Steam Generator Examination Facility (SGEF), specifically designed for nondestructive and destructive examination tasks of the SGIP was completed. The Surry generator previously transported to the Hanford Reservation was then inserted into the SGEF. Nondestructive characterization of the generator from both primary and secondary sides has been initiated. Decontamination of the channelhead cold leg side was conducted. Radioactive field maps were established in the steam generator, at the generator surface and in the SGEF

  3. Integrity Analysis of Damaged Steam Generator Tubes

    International Nuclear Information System (INIS)

    Stanic, D.

    1998-01-01

    Variety of degradation mechanisms affecting steam generator tubes makes steam generators as one of the critical components in the nuclear power plants. Depending of their nature, degradation mechanisms cause different types of damages. It requires performance of extensive integrity analysis in order to access various conditions of crack behavior under operating and accidental conditions. Development and application of advanced eddy current techniques for steam generator examination provide good characterization of found damages. Damage characteristics (shape, orientation and dimensions) may be defined and used for further evaluation of damage influence on tube integrity. In comparison with experimental and analytical methods, numerical methods are also efficient tools for integrity assessment. Application of finite element methods provides relatively simple modeling of different type of damages and simulation of various operating conditions. The stress and strain analysis may be performed for elastic and elasto-plastic state with good ability for visual presentation of results. Furthermore, the fracture mechanics parameters may be calculated. Results obtained by numerical analysis supplemented with experimental results are the base for definition of alternative plugging criteria which may significantly reduce the number of plugged tubes. (author)

  4. Status of steam generator tubing integrity at Jaslovske Bohunice NPP

    International Nuclear Information System (INIS)

    Cepcek, S.

    1997-01-01

    Steam generator represents one of the most important component of nuclear power plants. Especially, loss of tubing integrity of steam generators can lead to the primary coolant leak to secondary circuit and in worse cases to the unit shut down or to the PTS events occurrence. Therefore, to ensure the steam generator tubing integrity and the current knowledge about tube degradation propagation and development is of the highest importance. In this paper the present status of steam generator tubing integrity in operated NPP in Slovak Republic is presented

  5. Steam generator tube integrity program. Phase I report

    International Nuclear Information System (INIS)

    Alzheimer, J.M.; Clark, R.A.; Morris, C.J.; Vagins, M.

    1979-09-01

    The results are presented of the pressure tests performed as part of Phase I of the Steam Generator Tube Integrity (SGTI) program at Battelle Pacific Northwest Laboratory. These tests were performed to establish margin-to-failure predictions for mechanically defected Pressurized Water Reactor (PWR) steam generator tubing under operating and accident conditions. Defect geometries tested were selected because they simulate known or expected defects in PWR steam generators. These defect geometries are Electric Discharge Machining (EDM) slots, elliptical wastage, elliptical wastage plus through-wall slot, uniform thinning, denting, denting plus uniform thinning, and denting plus elliptical wastage. All defects were placed in tubing representative of that currently used in PWR steam generators

  6. Steam generator tube integrity program: Phase II, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Kurtz, R.J.; Bickford, R.L.; Clark, R.A.; Morris, C.J.; Simonen, F.A.; Wheeler, K.R.

    1988-08-01

    The Steam Generator Tube Integrity Program (SGTIP) was a three phase program conducted for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL). The first phase involved burst and collapse testing of typical steam generator tubing with machined defects. The second phase of the SGTIP continued the integrity testing work of Phase I, but tube specimens were degraded by chemical means rather than machining methods. The third phase of the program used a removed-from-service steam generator as a test bed for investigating the reliability and effectiveness of in-service nondestructive eddy-current inspection methods and as a source of service degraded tubes for validating the Phase I and Phase II data on tube integrity. This report describes the results of Phase II of the SGTIP. The object of this effort included burst and collapse testing of chemically defected pressurized water reactor (PWR) steam generator tubing to validate empirical equations of remaining tube integrity developed during Phase I. Three types of defect geometries were investigated: stress corrosion cracking (SCC), uniform thinning and elliptical wastage. In addition, a review of the publicly available leak rate data for steam generator tubes with axial and circumferential SCC and a comparison with an analytical leak rate model is presented. Lastly, nondestructive eddy-current (EC) measurements to determine accuracy of defect depth sizing using conventional and alternate standards is described. To supplement the laboratory EC data and obtain an estimate of EC capability to detect and size SCC, a mini-round robin test utilizing several firms that routinely perform in-service inspections was conducted.

  7. Steam generator tube integrity program: Phase II, Final report

    International Nuclear Information System (INIS)

    Kurtz, R.J.; Bickford, R.L.; Clark, R.A.; Morris, C.J.; Simonen, F.A.; Wheeler, K.R.

    1988-08-01

    The Steam Generator Tube Integrity Program (SGTIP) was a three phase program conducted for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL). The first phase involved burst and collapse testing of typical steam generator tubing with machined defects. The second phase of the SGTIP continued the integrity testing work of Phase I, but tube specimens were degraded by chemical means rather than machining methods. The third phase of the program used a removed-from-service steam generator as a test bed for investigating the reliability and effectiveness of in-service nondestructive eddy-current inspection methods and as a source of service degraded tubes for validating the Phase I and Phase II data on tube integrity. This report describes the results of Phase II of the SGTIP. The object of this effort included burst and collapse testing of chemically defected pressurized water reactor (PWR) steam generator tubing to validate empirical equations of remaining tube integrity developed during Phase I. Three types of defect geometries were investigated: stress corrosion cracking (SCC), uniform thinning and elliptical wastage. In addition, a review of the publicly available leak rate data for steam generator tubes with axial and circumferential SCC and a comparison with an analytical leak rate model is presented. Lastly, nondestructive eddy-current (EC) measurements to determine accuracy of defect depth sizing using conventional and alternate standards is described. To supplement the laboratory EC data and obtain an estimate of EC capability to detect and size SCC, a mini-round robin test utilizing several firms that routinely perform in-service inspections was conducted

  8. Steam 80 steam generator instrumentation

    International Nuclear Information System (INIS)

    Carson, W.H.; Harris, H.H.

    1980-01-01

    This paper describes two special instrumentation packages in an integral economizer (preheater) steam generator of one of the first System 80 plants scheduled to go into commercial operation. The purpose of the instrumentation is to obtain accurate operating information from regions of the secondary side of the steam generator inaccessible to normal plant instrumentation. In addition to verification of the System 80 steam generator design predictions, the data obtained will assist in verification of steam generator thermal/hydraulic computer codes developed for generic use in the industry

  9. Steam generator with integral downdraft dryer

    Energy Technology Data Exchange (ETDEWEB)

    Hochmuth, F.W.

    1992-02-01

    On June 30, 1989, a financial assistance award was granted by the United State Department of Energy, the purpose of which was to study and evaluate the technical aspect, the economic viability, and commercial possibilities of a new furnace design for burning high moisture cellulose type fuels. The new design is an invention by F.W. Hochmuth, P.Eng. and has received United States Patents Nos. 4,480, 557 and 4,502,397. It was conceived as a method to improve the general operation and efficiency of waste wood burning boilers, to avoid the use of stabilizing fuels such as oil or gas, and to reduce objectionable stack emissions. A further objective was to obtain such benefits at relatively low cost by integrating all new material requirements within the furnace itself thereby avoiding the need for costly external equipment. The proposed integral down-draft dryer avoids the use of external dryer systems that are very expensive, have high power consumption, and require a large amount of maintenance. This document provides the details of this invention.

  10. ANL/CANTIA code for steam generator tube integrity assessment

    International Nuclear Information System (INIS)

    Revankar, S.T.; Wolf, B.; Majumdar, S.; Riznic, J.R.

    2009-01-01

    Steam generator (SG) tubes have an important safety role in CANDU type reactors and Pressurized Water Reactors (PWR) because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear plant. The SG tubes are susceptible to corrosion and damage. A failure of a single steam generator tube, or even a few tubes, would not be a serious safety-related event in a CANDU reactor. The leakage from a ruptured tube is within makeup capacity of the primary heat transport system, so that as long as the operator takes the correct actions, the off-site consequences will be negligible. A sufficient safety margin against tube rupture used to be the basis for a variety of maintenance strategies developed to maintain a suitable level of plant safety and reliability. Several through-wall flaws may remain in operation and potentially contribute to the total primary-to-secondary leak rate. Assessment of the conditional probabilities of tube failures, leak rates, and ultimately risk of exceeding licensing dose limits has been used for steam generator tube fitness-for-service assessment. The advantage of this type of analysis is that it avoids the excessive conservatism typically present in deterministic methodologies. However, it requires considerable effort and expense to develop all of the failure, leakage, probability of detection, and flaw growth distributions and models necessary to obtain meaningful results from a probabilistic model. The Canadian Nuclear Safety Commission (CNSC) recently developed the CANTIA methodology for probabilistic assessment of inspection strategies for steam generator tubes as a direct effect on the probability of tube failure and primary-to-secondary leak rate Recently Argonne National Laboratory has developed tube integrity and leak rate models under Integrated Steam Generator Tube Integrity Program (ISGTIP-2). These models have been incorporated in the ANL/CANTIA code. This paper presents the ANL

  11. Steam Generator Tube Integrity Program: Surry Steam Generator Project, Hanford site, Richland, Benton County, Washington: Environmental assessment

    International Nuclear Information System (INIS)

    1980-03-01

    The US Nuclear Regulatory Commission (NRC) has placed a Nuclear Regulatory Research Order with the Richland Operations Office of the US Department of Energy (DOE) for expanded investigations at the DOE Pacific Northwest Laboratory (PNL) related to defective pressurized water reactor (PWR) steam generator tubing. This program, the Steam Generator Tube Integrity (SGTI) program, is sponsored by the Metallurgy and Materials Research Branch of the NRC Division of Reactor Safety Research. This research and testing program includes an additional task requiring extensive investigation of a degraded, out-of-service steam generator from a commercial nuclear power plant. This comprehensive testing program on an out-of-service generator will provide NRC with timely and valuable information related to pressurized water reactor primary system integrity and degradation with time. This report presents the environmental assessment of the removal, transport, and testing of the steam generator along with decontamination/decommissioning plans

  12. Steam generator

    International Nuclear Information System (INIS)

    Fenet, J.-C.

    1980-01-01

    Steam generator particularly intended for use in the coolant system of a pressurized water reactor for vaporizing a secondary liquid, generally water, by the primary cooling liquid of the reactor and comprising special arrangements for drying the steam before it leaves the generator [fr

  13. Research perspectives on the evaluation of steam generator tube integrity

    International Nuclear Information System (INIS)

    Muscara, J.; Diercks, D. R.; Majumdar, S.; Kupperman, D. S.; Bakhtiari, S.; Shack, W. J.

    2001-01-01

    Industry effects have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, SCC and intergranular attack have proved more difficult to manage. Although steam generator replacements are proceeding there is substantial industry interest in operating with degraded steam generators, and significant numbers of plants will continue to do so. In most cases degradation of steam generator tubing by stress corrosion cracking is still managed by plug or repair on detection, because current NDE techniques for characterization of flaws are not accurate enough to permit continued operation. This paper reviews some of the historical background that underlies current steam generator degradation management strategies and outlines some of the additional research that must be done to provide more effective management of degradation in current generators and provide greater assurance of satisfactory performance in replacement steam generators

  14. Evaluation of material integrity on electricity generator water steam cycles component (Main Steam Pipe)

    International Nuclear Information System (INIS)

    Sudardjo; Histori; Triyadi, Ari

    1998-01-01

    The evaluation of material integrity on electricity generator component has been done. That component was main steam pipe of Unit II Suralaya Coal Fired Power Plant. evaluation was done by replication technique. The damage was found are two porosity's, from two point samples of six points sample population. Based on cavity evaluation in steels, which proposed by Neubauer and Wedel that porosity's still at class A damage. For class A damage, its means no remedial action would be required until next major scheduled maintenance outage. That porosity's was grouped on isolated cavities and not need ti repair that main steam pipe component less than three year after replication test

  15. Steam Generator tube integrity -- US Nuclear Regulatory Commission perspective

    International Nuclear Information System (INIS)

    Murphy, E.L.; Sullivan, E.J.

    1997-01-01

    In the US, the current regulatory framework was developed in the 1970s when general wall thinning was the dominant degradation mechanism; and, as a result of changes in the forms of degradation being observed and improvements in inspection and tube repair technology, the regulatory framework needs to be updated. Operating experience indicates that the current U.S. requirements should be more stringent in some areas, while in other areas they are overly conservative. To date, this situation has been dealt with on a plant-specific basis in the US. However, the NRC staff is now developing a proposed steam generator rule as a generic framework for ensuring that the steam generator tubes are capable of performing their intended safety functions. This paper discusses the current U.S. regulatory framework for assuring steam generator (SG) tube integrity, the need to update this regulatory framework, the objectives of the new proposed rule, the US Nuclear Regulatory Commission (NRC) regulatory guide (RG) that will accompany the rule, how risk considerations affect the development of the new rule, and some outstanding issues relating to the rule that the NRC is still dealing with

  16. WWER steam generator tube structural and leakage integrity

    International Nuclear Information System (INIS)

    Splichal, K.; Krhounek, Vl.; Otruba, J.; Ruscak, M.

    1998-01-01

    The integrity of heat exchange tubes may influence the lifetime of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirements are to assure very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evaluation and heat exchange tubes plugging. The stress corrosion cracking and pitting are the main corrosion damages of WWER heat exchange tubes and are initiated from the outer surface. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through wall cracks, oriented preferentially in the axial direction. The paper presents the leakage and plugging limits for WWER steam generators, which have been determined from leak tests and burst tests. The tubes with axial part-through and through-wall defects have been used. The permissible value of primary to secondary leak rate was evaluated with respect to permissible axial through-wall defect size of WWER 440 and 1000 steam generator tubes. Blocking of the tube cracks by corrosion product particles and other compounds reduces the primary to secondary leak rate. The plugging limits involve the following factors: permissible tube wall thickness which determine further operation of the tubes with defects and assures their integrity under operating conditions and permissible size of a through-wall crack which is sufficiently stable under normal and accident conditions in relation to the critical crack length. For the evaluation of burst test of heat exchange tubes with longitudinal through-wall defects the instability criterion has been used and the dependence of the normalised burst pressure on the normalised length of an axial through-wall defect has been determined. The validity of the criterion of instability for WWER tubes with through

  17. Research perspectives on the evaluation of steam generator tube integrity

    International Nuclear Information System (INIS)

    Muscara, J.; Diercks, D.R.; Majumdar, S.; Kupperman, D.S.; Bakhtiari, S.; Shack, W.J.

    2002-01-01

    Industry efforts have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, stress corrosions cracking (SCC) and intergranular attack have proved more difficult to manage. Although steam generator replacements are proceeding, there is substantial industry interest in operating with degraded steam generators, and significant numbers of plants will continue to do so. In most cases degradation of steam generator tubing by stress corrosion cracking is still managed by 'plug or repair on detection' because current NDE techniques for characterization of flaws and the knowledge of SCC crack growth rates are not accurate enough to permit continued operation. Replacement generators with improved designs and materials have performed well to date, but previous experience with the appearance of some types of SCC in Alloy 600 after 10 years or more of operation and laboratory results suggest additional understanding of corrosion performance of these materials is needed. This paper reviews some of the historical background that underlies current steam generator degradation management strategies and outlines some of the additional research that must be done to provide more effective management of degradation in current generators and provide greater assurance of satisfactory performance in replacement steam generators. (author)

  18. Thermal-hydraulic study of integrated steam generator in PWR

    International Nuclear Information System (INIS)

    Osakabe, Masahiro

    1989-01-01

    One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR. The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated. (author)

  19. An integrated leak detection system for the ALMR steam generator

    International Nuclear Information System (INIS)

    Dayal, Y.; Gaubatz, D.C.; Wong, K.K.; Greene, D.A.

    1995-01-01

    The steam generator (SG) of the Advanced Liquid Metal Reactor (ALMR) system serves as a heat exchanger between the shell side secondary loop hot liquid sodium and the tube side water/steam mixture. A leak in the tube will result in the injection of the higher pressure water/steam into the sodium and cause an exothermic sodium-water reaction. An initial small leak (less than 1 gm/sec) can escalate into an intermediate size leak in a relatively short time by self enlargement of the original flaw and by initiating leaks in neighboring tubes. If not stopped, complete rupture of one or more tubes can cause injection rates of thousands of gm/sec and result in the over pressurization of the secondary loop rupture disk and dumping of the sodium to relieve pressure. The down time associated with severe sodium-water reaction damage has great adverse economic consequence. An integrated leak detection system (ILDS) has been developed which utilizes both chemical and acoustic sensors for improved leak detection. The system provides SG leak status to the reactor operator and is reliable enough to trigger automatic control action to protect the SG. The ILDS chemical subsystem uses conventional in-sodium and cover gas hydrogen detectors and incorporates knowledge based effects due to process parameters for improved reliability. The ILDS acoustic subsystem uses an array of acoustic sensors and incorporates acoustic beamforming technology for highly reliable and accurate leak identification and location. The new ILDS combines the small leak detection capability of the chemical system with the reliability and rapid detection/location capability of the acoustic system to provide a significantly improved level of protection for the SG over a wide range of operation conditions. (author)

  20. Evaluation of steam generator tube integrity during earthquakes

    Energy Technology Data Exchange (ETDEWEB)

    Kusakabe, Takaya; Kodama, Toshio [Mitsubishi Heavy Industries Ltd., Kobe (Japan). Kobe Shipyard and Machinery Works; Takamatsu, Hiroshi; Matsunaga, Tomoya

    1999-07-01

    This report shows an experimental study on the strength of PWR steam generator (SG) tubes with various defects under cyclic loads which simulate earthquakes. The tests were done using same SG tubing as actual plants with axial and circumferential defects with various length and depth. In the tests, straight tubes were loaded with cyclic bending moments to simulate earthquake waves and number of load cycles at which tube leak started or tube burst was counted. The test results showed that even tubes with very long crack made by EDM more than 80% depth could stand the maximum earthquake, and tubes with corrosion crack were far stronger than those. Thus the integrity of SG tubes with minute potential defects was demonstrated. (author)

  1. Steam generators

    International Nuclear Information System (INIS)

    Hayden, R.L.J.

    1979-01-01

    Steam generators for nuclear reactors are designed so that deposition of solids on the surface of the inlet side of the tubesheet or the inlet header with the consequent danger of corrosion and eventual tube failure is obviated or substantially reduced. (U.K.)

  2. Integrated steam generation process and system for enhanced oil recovery

    Energy Technology Data Exchange (ETDEWEB)

    Betzer-Zilevitch, M. [Ex-Tar Technologies Inc., Calgary, AB (Canada)

    2010-07-01

    A method of producing steam for the extraction of heavy bitumens was presented. The direct contact steam generation (DCSG) method is used for the direct heat transfer between combustion gas and contaminated liquid phase water to generate steam. This paper presented details of experimental and field studies conducted to demonstrate the DCSG. Results of the study demonstrated that pressure and temperature are positively correlated. As pressure increases, the flow rate of the discharged mass decreases and the steam ratio decreases. As pressure increases, the condensate and distillate flow rates increases while water vapor losses in the non-condensable gases decrease. The study indicated that for a 10 bar pressurized system producing 9.6 mt per hour of 10,000 kpa steam and 9.6 mt per hour of distillate BFW, 70 percent of the combustion energy should be recovered to generate 10,000 kpa pressure steam for EOR. Combustion energy requirements were found to decrease when pressure decreases. 11 refs., 5 tabs., 8 figs.

  3. Application of Integrated Neural Network Method to Fault Diagnosis of Nuclear Steam Generator

    International Nuclear Information System (INIS)

    Zhou Gang; Yang Li

    2009-01-01

    A new fault diagnosis method based on integrated neural networks for nuclear steam generator (SG) was proposed in view of the shortcoming of the conventional fault monitoring and diagnosis method. In the method, two neural networks (ANNs) were employed for the fault diagnosis of steam generator. A neural network, which was used for predicting the values of steam generator operation parameters, was taken as the dynamics model of steam generator. The principle of fault monitoring method using the neural network model is to detect the deviations between process signals measured from an operating steam generator and corresponding output signals from the neural network model of steam generator. When the deviation exceeds the limit set in advance, the abnormal event is thought to occur. The other neural network as a fault classifier conducts the fault classification of steam generator. So, the fault types of steam generator are given by the fault classifier. The clear information on steam generator faults was obtained by fusing the monitoring and diagnosis results of two neural networks. The simulation results indicate that employing integrated neural networks can improve the capacity of fault monitoring and diagnosis for the steam generator. (authors)

  4. Hydrogen generation utilizing integrated CO2 removal with steam reforming

    Science.gov (United States)

    Duraiswamy, Kandaswamy; Chellappa, Anand S

    2013-07-23

    A steam reformer may comprise fluid inlet and outlet connections and have a substantially cylindrical geometry divided into reforming segments and reforming compartments extending longitudinally within the reformer, each being in fluid communication. With the fluid inlets and outlets. Further, methods for generating hydrogen may comprise steam reformation and material adsorption in one operation followed by regeneration of adsorbers in another operation. Cathode off-gas from a fuel cell may be used to regenerate and sweep the adsorbers, and the operations may cycle among a plurality of adsorption enhanced reformers to provide a continuous flow of hydrogen.

  5. Strategy for assessment of WWER steam generator tube integrity. Report prepared within the framework of the coordinated research project on verification of WWER steam generator tube integrity

    International Nuclear Information System (INIS)

    2007-12-01

    Steam generator heat exchanger tube degradations happen in WWER Nuclear Power Plant (NPP). The situation varies from country to country and from NPP to NPP. More severe degradation is observed in WWER-1000 NPPs than in case of WWER-440s. The reasons for these differences could be, among others, differences in heat exchanger tube material (chemical composition, microstructure, residual stresses), in thermal and mechanical loadings, as well as differences in water chemistry. However, WWER steam generators had not been designed for eddy current testing which is the usual testing method in steam generators of western PWRs. Moreover, their supplier provided neither adequate methodology and criteria nor equipment for planning and implementing In-Service Inspection (ISI). Consequently, WWER steam generator ISI infrastructure was established with delay. Even today, there are still big differences in the eddy current inspection strategy and practice as well as in the approach to steam generator heat exchanger tube structural integrity assessment (plugging criteria for defective tubes vary from 40 to 90% wall thickness degradation). Recognizing this situation, the WWER operating countries expressed their need for a joint effort to develop methodology to establish reasonable commonly accepted integrity assessment criteria for the heat exchanger tubes. The IAEA's programme related to steam generator life management is embedded into the systematic activity of its Technical Working Group on Life Management of Nuclear Power Plants (TWG-LMNPP). Under the advice of the TWG-LMNPP, an IAEA coordinated research project (CRP) on Verification of WWER Steam Generator Tube Integrity was launched in 2001. It was completed in 2005. Thirteen organizations involved in in-service inspection of steam generators in WWER operating countries participated: Croatia, Czech Republic, Finland, France, Hungary, Russian Federation, Slovakia, Spain, Ukraine, and the USA. The overall objective was to

  6. Structural and leakage integrity assessment of WWER steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Splichal, K.; Otruba, J. [Nuclear Research Inst., Rez (Switzerland)

    1997-12-31

    The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction. 10 refs.

  7. Structural and leakage integrity assessment of WWER steam generator tubes

    International Nuclear Information System (INIS)

    Splichal, K.; Otruba, J.

    1997-01-01

    The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction

  8. Structural and leakage integrity assessment of WWER steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Splichal, K; Otruba, J [Nuclear Research Inst., Rez (Switzerland)

    1998-12-31

    The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction. 10 refs.

  9. Define optimal conditions for steam generator tube integrity and an extended steam generator service life

    International Nuclear Information System (INIS)

    Lu, Y.C.

    2007-01-01

    Steam generator (SG) tubing materials are susceptible to corrosion degradation in certain electrochemical corrosion potential regions in the presence of some aggressive ions. Because of the hideout of impurities, the local chemistry conditions in areas under sludge and inside SG crevices may be very aggressive with high concentrations of chlorides and other impurities. These areas are the locations where SG tubing materials are susceptible to degradation such as pitting, crevice corrosion, intergranular attack (IGA) and stress corrosion cracking (SCC). The corrosion susceptibility of each SG alloy is different and is a function of the electrochemical corrosion potential (ECP) and chemical environment. Electrochemical corrosion behaviors of major SG tube alloys were studied under some plausible aggressive crevice chemistry conditions. The possible hazardous conditions leading to SG tube degradation and the conditions, which can minimize SG tube degradation have been determined. Optimal operating conditions in the form of a 'Recommended ECP/pH zone' for minimizing corrosion degradation have been defined for all major SG tube materials, including Alloys 600, 800, 690 and 400, under CANDU SG operating and startup conditions. SCC tests and accelerated corrosion tests were carried out to verify and revise the recommended ECP/pH zones. This information is being incorporated into ChemAND, a system health monitor for plant chemistry management developed by AECL, which alloys utilities to evaluate the status of the SG alloys and to minimize SG material degradation by appropriate SG water chemistry management. (author)

  10. Assesment of integrity of WWER steam generator tubes

    International Nuclear Information System (INIS)

    Splichal, K.; Brozova, A.; Zdarek, J.

    1992-01-01

    Full text: The leak rates measurement project was held to give experimental data enabling the Czechoslovak Atomic Agency Inspection to decree the change in the Technical Specification allowable limit of steam generator activity release on secondary side. The WWER types of nuclear power plants in Czechoslovakia have horizontal steam generators. The tubes studying in frame of the project belong to steam generator WWER- 440 type, the diameter of tube is 16 mm, the wall thickness 1.4 mm. The subject of the project was the measurement of service leak rates of typical in service cracks. Secondary side stress corrosion cracks were determined as the typical crack created in service condition. These cracks were prepared in tubes artificially by exposition in chloride environment accompanied by an internal stress. The experimental device consisted of a pressure vessel connected with pressure water loop, a cooling vessel for leakage medium and a measuring vessel. The leak rates were determined as a slope of plots the leakage volume - time. Inside the pressure vessel the steam generator operation environment was simulated. It means: primary side of tube 12.5 MPa, Z90 deg. C, secondary side -4.6MPa, 250 deg. C, water service quality. We observed reduce of leak rate in course of time in each experiment. We suppose the tubes were stopped up by deposits formed in manufacturing of crack and in experiment. Our opinion has been proved by fractography. Project results in recommendation for in service leak rate limit based on safety factors with respect to critical crack lengths and for determination of tube plugging criteria. (author)

  11. Steam generator asset management: integrating technology and asset management

    International Nuclear Information System (INIS)

    Shoemaker, P.; Cislo, D.

    2006-01-01

    Asset Management is an established but often misunderstood discipline that is gaining momentum within the nuclear generation industry. The global impetus behind the movement toward asset management is sustainability. The discipline of asset management is based upon three fundamental aspects; key performance indicators (KPI), activity-based cost accounting, and cost benefits/risk analysis. The technology associated with these three aspects is fairly well-developed, in all but the most critical area; cost benefits/risk analysis. There are software programs that calculate, trend, and display key-performance indicators to ensure high-level visibility. Activity-based costing is a little more difficult; requiring a consensus on the definition of what comprises an activity and then adjusting cost accounting systems to track. In the United States, the Nuclear Energy Institute's Standard Nuclear Process Model (SNPM) serves as the basis for activity-based costing. As a result, the software industry has quickly adapted to develop tracking systems that include the SNPM structure. Both the KPI's and the activity-based cost accounting feed the cost benefits/risk analysis to allow for continuous improvement and task optimization; the goal of asset management. In the case where the benefits and risks are clearly understood and defined, there has been much progress in applying technology for continuous improvement. Within the nuclear generation industry, more specialized and unique software systems have been developed for active components, such as pumps and motors. Active components lend themselves well to the application of asset management techniques because failure rates can be established, which serves as the basis to quantify risk in the cost-benefits/risk analysis. A key issue with respect to asset management technologies is only now being understood and addressed, that is how to manage passive components. Passive components, such as nuclear steam generators, reactor vessels

  12. Structural and leakage integrity assessment of WWER steam generator tubes

    International Nuclear Information System (INIS)

    Splichal, K.; Otruba, J.; Keilova, E.; Krhounek, V.; Turek, J.

    1996-01-01

    The leakage and plugging limits were derived for WWER steam generators based on leak and burst tests using tubes with axial part-through and through-wall defects. The following conclusions were arrived at: (i) The permissible primary-to-secondary leak rate with respect to the permissible through-wall defect size of WWER-440 and WWER-1000 steam generator tubes is 8 l/h. (ii) The primary-to-secondary leak rate is reduced by the blocking of the tube cracks by corrosion product particles and other substances. (iii) The rate of crack penetration through the tube wall is higher than the crack widening. (iv) The validity of the criterion of instability for tubes with through-wall cracks was confirmed experimentally. For the WWER-440 and WWER-1000 steam generators, the critical size of axial through-wall cracks, for the threshold primary-to-secondary pressure difference, is 13.8 and 12.0 mm, respectively. (v) The calculated leakage for the rupture of one tube and for the assumed extreme defects is two orders and one order of magnitude, respectively, higher than the proposed primary water leakage limit of 8 l/h. (vi) The experiments gave evidence that the use of the permissible thinning limit of 80% for the heat exchange tube plugging does not bring about uncontrollable leakage or unstable crack growth. This is consistent with experience gained at WWER-440 type nuclear power plants. 4 tabs., 5 figs., 9 refs

  13. Steam generator life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

    1998-01-01

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermal hydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. This report shows how recent advances in cleaning technology are integrated into a life management strategy, discusses downcomer flow measurement as a means of monitoring steam generator condition, and describes recent advances in hideout return as a life management tool. The research and development program, as well as operating experience, has identified

  14. Steam generator tube integrity requirements and operating experience in the United States

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    2009-01-01

    Steam generator tube integrity is important to the safe operation of pressurized-water reactors. For ensuring tube integrity, the U.S. Nuclear Regulatory Commission uses a regulatory framework that is largely performance based. This performance-based framework is supplemented with some prescriptive requirements. The framework recognizes that there are three combinations of tube materials and heat treatments currently used in the United States and that the operating experience depends, in part, on the type of material used. This paper summarizes the regulatory framework for ensuring steam generator tube integrity, it highlights the current status of steam generators, and it highlights some of the steam generator issues and challenges that exist in the United States. (author)

  15. Steam generators - problems and prognosis

    International Nuclear Information System (INIS)

    Tapping, R.L.

    1997-05-01

    Steam-generator problems, largely a consequence of corrosion and fouling, have resulted in increased inspection requirements and more regulatory attention to steam-generator integrity. In addition, utilities have had to develop steam-generator life-management strategies, including cleaning and replacement, to achieve design life. This paper summarizes the pertinent data to 1993/1994, and presents an overview of current steam-generator management practices. (author)

  16. Main results of assessing integrity of RNPP-3 steam generator heat exchange tubes in accident management

    International Nuclear Information System (INIS)

    Shugajlo, Al-j P.; Mustafin, M.A.; Shugajlo, Al-r P.; Ryzhov, D.I.; Zhabin, O.I.

    2017-01-01

    Tubes integrity evaluation under accident conditions considering drain of SG and current technical state of steam exchange tubes is an important question regarding SG long-term operation and improvement of accident management strategy.The main investigation results prepared for heat exchange surface of RNPP-3 steam generator are presented in this research aimed at assessing integrity of heat exchange tubes under accident conditions, which lead to full or partial drain of heat exchange surface, in particular during station blackout.

  17. Some aspects affecting fast reactor steam generator integrity considered from a utility viewpoint

    Energy Technology Data Exchange (ETDEWEB)

    Bolt, P R

    1975-07-01

    The important conditions affecting fast reactor steam generator integrity are discussed. In addition to the need for high integrity levels when the steam generator is first delivered to the power station site, the equally important aspect of demonstrating retention of continued high levels of integrity throughout the operating life of the station is described. The functional and related conditions that are believed important to the selection of a design type which can offer adequately high levels of integrity are given. Some of the data needs of a utility concerned with fast reactor S.G.U. design assessment are described, particular emphasis being given to areas believed to have a significant effect on steam generator reliability and integrity. (author)

  18. Design study on steam generator integration into the VVER reactor pressure vessel

    International Nuclear Information System (INIS)

    Hort, J.; Matal, O.

    2004-01-01

    The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

  19. HTGR steam generator development

    International Nuclear Information System (INIS)

    Schuetzenduebel, W.G.; Hunt, P.S.; Weber, M.

    1976-01-01

    More than 40 gas-cooled reactor plants have produced in excess of 400 reactor years of operating experience which have proved a reasonably high rate of gas-cooled reactor steam generator availability. The steam generators used in these reactors include single U-tube and straight-tube steam generators as well as meander type and helically wound or involute tube steam generators. It appears that modern reactors are being equipped with helically wound steam generators of the once-through type as the end product of steam generator evolution in gas-cooled reactor plants. This paper provides a general overview of gas-cooled reactor steam generator evolution and operating experience and shows how design criteria and constraints, research and development, and experience data are factored into the design/development of modern helically wound tube steam generators for the present generation of gas-cooled reactors

  20. Steam generator performance improvements for integral small modular reactors

    Directory of Open Access Journals (Sweden)

    Muhammad Ilyas

    2017-12-01

    Results and Conclusions: The results are compared with helical-coiled SGs being used in IRIS (International Reactor Innovative and Secure. The results show that the tube length reduces up to 11.56% keeping thermal and hydraulic conditions fixed. In the case of fixed size, the steam outlet temperature increases from 590.1 K to 597.0 K and the capability of power transfer from primary to secondary also increases. However, these advantages are associated with some extra pressure drop, which has to be compensated.

  1. Steam generator tube extraction

    International Nuclear Information System (INIS)

    Delorme, H.

    1985-05-01

    To enable tube examination on steam generators in service, Framatome has now developed a process for removing sections of steam generator tubes. Tube sections can be removed without being damaged for treating the tube section expanded in the tube sheet

  2. Evaluation of material integrity on electricity power steam generator cycles (turbine casing) component

    International Nuclear Information System (INIS)

    Histori; Benedicta; Farokhi; S A, Soedardjo; Triyadi, Ari; Natsir, M

    1999-01-01

    The evaluation of material integrity on power steam generator cycles component was done. The test was carried out on casing turbine which is made from Inconel 617. The tested material was taken from t anjung Priok plant . The evaluation was done by metallography analysis using microscope with magnification of 400. From the result, it is shown that the material grains are equiaxed

  3. Operational control and maintenance integrity of typical and atypical coil tube steam generating systems

    Energy Technology Data Exchange (ETDEWEB)

    Beardwood, E.S.

    1999-07-01

    Coil tube steam generators are low water volume to boiler horsepower (bhp) rating, rapid steaming units which occupy substantially less space per boiler horsepower than equivalent conventional tire tube and water tube boilers. These units can be retrofitted into existing steam systems with relative ease and are more efficient than the generators they replace. During the early 1970's they became a popular choice for steam generation in commercial, institutional and light to medium industrial applications. Although these boiler designs do not require skilled or certified operators, an appreciation for a number of the operational conditions that result in lower unscheduled maintenance, increased reliability and availability cycles would be beneficial to facility owners, managers, and operators. Conditions which afford lower operating and maintenance costs will be discussed from a practical point of view. An overview of boiler design and operation is also included. Pitfalls are provided for operational and idle conditions. Water treatment application, as well as steam system operations not conducive to maintaining long term system integrity; with resolutions, will be addressed.

  4. Steam generator tubes integrity: In-service-inspection

    International Nuclear Information System (INIS)

    Comby, R.J.

    1997-01-01

    The author's approach to tube integrity is in terms of looking for flaws in tubes. The basis for this approach is that no simple rules can be fixed to adopt a universal inspection methodology because of various concepts related to experience, leak acceptance, leak before break approach, etc. Flaw specific management is probably the most reliable approach as a compromise between safety, availability and economic issues. In that case, NDE capabilities have to be in accordance with information required by structural integrity demonstration. The author discusses the types of probes which can be used to search for flaws in addition to the types of flaws which are being sought, with examples of specific analysis experiences. The author also discusses the issue of a reporting level as it relates to avoiding false calls, classifying faults, and allowing for automation in analysis

  5. Steam generator with integral downdraft dryer. Final project report

    Energy Technology Data Exchange (ETDEWEB)

    Hochmuth, F.W.

    1992-02-01

    On June 30, 1989, a financial assistance award was granted by the United State Department of Energy, the purpose of which was to study and evaluate the technical aspect, the economic viability, and commercial possibilities of a new furnace design for burning high moisture cellulose type fuels. The new design is an invention by F.W. Hochmuth, P.Eng. and has received United States Patents Nos. 4,480, 557 and 4,502,397. It was conceived as a method to improve the general operation and efficiency of waste wood burning boilers, to avoid the use of stabilizing fuels such as oil or gas, and to reduce objectionable stack emissions. A further objective was to obtain such benefits at relatively low cost by integrating all new material requirements within the furnace itself thereby avoiding the need for costly external equipment. The proposed integral down-draft dryer avoids the use of external dryer systems that are very expensive, have high power consumption, and require a large amount of maintenance. This document provides the details of this invention.

  6. Steam generator life management

    International Nuclear Information System (INIS)

    King, P.; McGillivray, R.; Reinhardt, W.; Millman, J.; King, B.; Schneider, W.

    2003-01-01

    'Full-Text:' Steam Generator Life Management responsibility embodies doing whatever is necessary to maintain the steam generation equipment of a nuclear plant in effective, reliable service. All comes together in that most critical deliverable, namely the submission of the documentation which wins approval for return to service after an outage program. Life management must address all aspects of SG reliability over the life of the plant. Nevertheless, the life management activities leading up to return to service approval is where all of it converges. Steam Generator Life Management activities entail four types of work, all equally important in supporting the objective of successful operation. These activities are i) engineering functions; including identification of inspection and maintenance requirements, outage planning and scope definition plus engineering assessment, design and analysis as necessary to support equipment operation, ii) fitness of service work; including the expert evaluation of degradation mechanisms, disposition of defects for return to service or not, and the fitness for service analysis as required to justify ongoing operation with acceptable defects, iii) inspection work; including large scale eddy current inspection of tubing, the definition of defect size and character, code inspections of pressure vessel integrity and visual inspections for integrity and iv) maintenance work; including repairs, retrofits, cleaning and modifications, all as necessary to implement the measures defined during activities i) through iii). The paper discusses the approach and execution of the program for the achievement of the above objectives and particularly of items i) and ii). (author)

  7. Performance Evaluation of a Printed Circuit Steam Generator for Integral Reactors: A Feasibility Test

    Energy Technology Data Exchange (ETDEWEB)

    Han, Hun Sik; Kang, Han-Ok; Yoon, Juhyeon; Kim, Young In; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of); Seo, Jang-won; Choi, Brain [Alfa Laval Korea Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    SMART (System-integrated Modular Advanced ReacTor) is a small-sized integral type pressurized water reactor. It adopts advanced design features such as structural safety improvement, system simplification, and component modularization to achieve highly enhanced safety and improved economics. The design issues related to further safety enhancement and cost reduction have received significant attention to increase its competitiveness in the global small reactor market. For the cost reduction, it is important to design the reactor vessel as small as possible. Thus, it is necessary to reduce the volume of main components such as a steam generator. Its manufacturing processes of the chemical etching and diffusion bonding provide high effectiveness, high compactness, and inherent structural safety under high temperatures and high pressures. Thus, it is expected to be an alternative to the conventional shell and tube type steam generator in SMART. In this paper, simple thermal-hydraulic performance measurement of a small-scale printed circuit steam generator (PCSG) is conducted to investigate the feasibility of applying it to SMART. The simple thermal-hydraulic performance of the PCSG has been experimentally evaluated. A small-scale PCHE is employed to investigate the feasibility of operating it as a steam generator. The performance assessment reveals that the PCSG stably produces superheated steam, and the increased degree of superheat is obtained at lower water flow rate. However, the flow instability is increased with the decrease of the water flow rate. Thus, it is required to apply the orifice design into the cold side plate to suppress the density-wave oscillations. The pressure drops and heat transfer rates increase with the water flow rate.

  8. An integrated automatic system for the eddy-current testing of the steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Hee Gon; Choi, Seong Su [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

    1995-12-31

    This research project was focused on automation of steam generator tubes inspection for nuclear power plants. ECT (Eddy Current Testing) inspection process in nuclear power plants is classified into 3 subprocesses such as signal acquisition process, signal evaluation process, and inspection planning and data management process. Having been automated individually, these processes were effectively integrated into an automatic inspection system, which was implemented in HP workstation with expert system developed (author). 25 refs., 80 figs.

  9. An integrated automatic system for the eddy-current testing of the steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Hee Gon; Choi, Seong Su [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

    1996-12-31

    This research project was focused on automation of steam generator tubes inspection for nuclear power plants. ECT (Eddy Current Testing) inspection process in nuclear power plants is classified into 3 subprocesses such as signal acquisition process, signal evaluation process, and inspection planning and data management process. Having been automated individually, these processes were effectively integrated into an automatic inspection system, which was implemented in HP workstation with expert system developed (author). 25 refs., 80 figs.

  10. EPRI steam generator programs

    International Nuclear Information System (INIS)

    Martel, L.J.; Passell, T.O.; Bryant, P.E.C.; Rentler, R.M.

    1977-01-01

    The paper describes the current overall EPRI steam generator program plan and some of the ongoing projects. Because of the recent occurrence of a corrosion phenomenon called ''denting,'' which has affected a number of operating utilities, an expanded program plan is being developed which addresses the broad and urgent needs required to achieve improved steam generator reliability. The goal of improved steam generator reliability will require advances in various technologies and also a management philosophy that encourages conscientious efforts to apply the improved technologies to the design, procurement, and operation of plant systems and components that affect the full life reliability of steam generators

  11. Comparison of Heat Transfer Fluid and Direct Steam Generation technologies for Integrated Solar Combined Cycles

    International Nuclear Information System (INIS)

    Rovira, Antonio; Montes, María José; Varela, Fernando; Gil, Mónica

    2013-01-01

    At present time and in the medium term, Solar Thermal Power Plants are going to share scenario with conventional energy generation technologies, like fossil and nuclear. In such a context, Integrated Solar Combined Cycles (ISCCs) may be an interesting choice since integrated designs may lead to a very efficient use of the solar and fossil resources. In this work, different ISCC configurations including a solar field based on parabolic trough collectors and working with the so-called Heat Transfer Fluid (HTF) and Direct Steam Generation (DSG) technologies are compared. For each technology, four layouts have been studied: one in which solar heat is used to evaporate part of the high pressure steam of a bottoming Rankine cycle with two pressure levels, another that incorporates a preheating section to the previous layout, the third one that includes superheating instead of preheating and the last one including both preheating and superheating in addition to the evaporation. The analysis is made with the aim of finding out which of the different layouts reaches the best performance. For that purpose, three types of comparisons have been performed. The first one assesses the benefits of including a solar steam production fixed at 50 MW th . The second one compares the configurations with a standardised solar field size instead of a fixed solar steam production. Finally, the last one consists on an even more homogeneous comparison considering the same steam generator size for all the configurations as well as standardised solar fields. The configurations are studied by mean of exergy analyses. Several figures of merit are used to correctly assess the configurations. Results reveal that the only-evaporative DSG configuration becomes the best choice, since it benefits of both low irreversibility at the heat recovery steam generator and high thermal efficiency in the solar field. Highlights: ► ISCC configurations with DSG and HTF technologies are compared. ► Four

  12. Proceedings of the CNRA/CSNI workshop on steam generator tube integrity in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R. [Argonne National Lab., IL (United States)

    1997-02-01

    The objective of the workshop was to provide a working forum for the exchange of information by contributing experts on current issues related to PWR steam generator tube integrity. One hundred persons from 15 countries attended the workshop, including 36 from regulatory and nuclear policy agencies, 28 from research and development laboratories, 18 from nuclear vendors and consulting firms, and 18 from electrical utilities. The workshop opened with a plenary session; the first part of the session covered international steam generator regulatory practices and issues, featuring speakers from regulatory bodies in Belgium, France, Japan, Spain, and the US. In Part 2 of the plenary session, comprehensive technical overviews on steam generator tubing degradation, inspection, and integrity were presented by authorities in these fields from the US, France, and Belgium. Parallel working sessions on the second and third days of the workshop then developed findings and recommendations in the areas of (1) tubing degradation, (2) tubing inspection, (3) tubing integrity, (4) preventative and corrective measures, and (5) operational aspects and risk analysis. On the final day of the workshop, the working-session facilitators presented summaries of their sessions to the workshop attendees. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  13. Proceedings of the CNRA/CSNI workshop on steam generator tube integrity in nuclear power plants

    International Nuclear Information System (INIS)

    Diercks, D.R.

    1997-02-01

    The objective of the workshop was to provide a working forum for the exchange of information by contributing experts on current issues related to PWR steam generator tube integrity. One hundred persons from 15 countries attended the workshop, including 36 from regulatory and nuclear policy agencies, 28 from research and development laboratories, 18 from nuclear vendors and consulting firms, and 18 from electrical utilities. The workshop opened with a plenary session; the first part of the session covered international steam generator regulatory practices and issues, featuring speakers from regulatory bodies in Belgium, France, Japan, Spain, and the US. In Part 2 of the plenary session, comprehensive technical overviews on steam generator tubing degradation, inspection, and integrity were presented by authorities in these fields from the US, France, and Belgium. Parallel working sessions on the second and third days of the workshop then developed findings and recommendations in the areas of (1) tubing degradation, (2) tubing inspection, (3) tubing integrity, (4) preventative and corrective measures, and (5) operational aspects and risk analysis. On the final day of the workshop, the working-session facilitators presented summaries of their sessions to the workshop attendees. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  14. Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

    International Nuclear Information System (INIS)

    Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun; Ueno, Shingo; Yoshino, Takehito

    1999-12-01

    In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the evaluation of merits

  15. Design of SMART steam generator cassette

    International Nuclear Information System (INIS)

    Kim, Y. W.; Kim, J. I.; Jang, M. H.

    2001-01-01

    Basic design development for the steam generator to be installed in the integral reactor SMART has been performed. Optimization of the steam generator shape, determination of the basic dimension and confirmation of the structural strength have been carried out. Individual steam generator cassette can be replaced in the optimized design concept of steam generator. Shape design of the steam generator cassette has been done on the computer based on 3-D CAE strategy. The structural integrity of the developed steam generator was investigated by performing the dynamic analysis for the steam generator cassette, flow induced vibration analysis for the tube bundle, and the thermo-mechanical analysis for the module header and tube. As for the manufacturing of steam generator, the numerical and the experimental simulation have been carried to control the amount of spring back and to eliminate residual stress. SMART steam generator cassette was developed by a sequential research of the aforementioned activities

  16. Safety provisions for steam generator in Mochovce nuclear power plant. BO CI 04 Integrity of primary collectors of VVER 440 steam generators

    International Nuclear Information System (INIS)

    Cikryt, F.; Bednarek, L.; Matocha, K.; Vejvoda, S.

    1997-01-01

    This paper dealt with the identification of possible damaging mechanism of the collector of the WWER 440 steam generator, cracking of primary collectors, corrosion damage of the protective coat of the primary collector circumferential weld, cracking of breathing space in the region of blinding effect by corrosion and strain, leaking of disassembling joint of the primary collector lid and with the integrity of heat exchanging tubes

  17. A new repair criterion for steam generator tubes with axial cracks based on probabilistic integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Su; Oh, Chang-Kyun [KEPCO Engineering and Construction Company, Inc., 269, Hyeoksin-ro, Gimcheon, Gyeongsangbuk-do 39660 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, College of Engineering, Kyung Hee University, 1732 Deokyoungdaero, Giheung, Yongin, Gyeonggi 446-701 (Korea, Republic of)

    2017-03-15

    Highlights: • Probabilistic assessment was performed for axially cracked steam generator tubes. • The threshold crack sizes were determined based on burst pressures of the tubes. • A new repair criterion was suggested as a function of operation time. - Abstract: Steam generator is one of the major components in a nuclear power plant, and it consists of thousands of thin-walled tubes. The operating record of the steam generators has indicated that a number of axial cracks due to stress corrosion have been frequently detected in the tubes. Since the tubes are closely related to the safety and also the efficiency of a nuclear power plant, an establishment of the appropriate repair criterion for the defected tubes and its applications are necessary. The objective of this paper is to develop an accurate repair criterion for the tubes with axial cracks. To do this, a thorough review is performed on the key parameters affecting the tube integrity, and then the probabilistic integrity assessment is carried out by considering the various uncertainties. In addition, the sizes of critical crack are determined by comparing the burst pressure of the cracked tube with the required performance criterion. Based on this result, the new repair criterion for the axially cracked tubes is defined from the reasonably conservative value such that the required performance criterion in terms of the burst pressure is able to be met during the next operating period.

  18. Steam generator tube integrity program: Annual report, August 1995--September 1996. Volume 2

    International Nuclear Information System (INIS)

    Diercks, D.R.; Bakhtiari, S.; Kasza, K.E.; Kupperman, D.S.; Majumdar, S.; Park, J.Y.; Shack, W.J.

    1998-02-01

    This report summarizes work performed by Argonne National Laboratory on the Steam Generator Tube Integrity Program from the inception of the program in August 1995 through September 1996. The program is divided into five tasks: (1) assessment of inspection reliability, (2) research on ISI (inservice-inspection) technology, (3) research on degradation modes and integrity, (4) tube removals from steam generators, and (5) program management. Under Task 1, progress is reported on the preparation of facilities and evaluation of nondestructive evaluation techniques for inspecting a mock-up steam generator for round-robin testing, the development of better ways to correlate failure pressure and leak rate with eddy current (EC) signals, the inspection of sleeved tubes, workshop and training activities, and the evaluation of emerging NDE technology. Results are reported in Task 2 on closed-form solutions and finite-element electromagnetic modeling of EC probe responses for various probe designs and flaw characteristics. In Task 3, facilities are being designed and built for the production of cracked tubes under aggressive and near-prototypical conditions and for the testing of flawed and unflawed tubes under normal operating, accident, and severe-accident conditions. Crack behavior and stability are also being modeled to provide guidance for test facility design, develop an improved understanding of the expected rupture behavior of tubes with circumferential cracks, and predict the behavior of flawed and unflawed tubes under severe accident conditions. Task 4 is concerned with the acquisition of tubes and tube sections from retired steam generators for use in the other research tasks. Progress on the acquisition of tubes from the Salem and McGuire 1 nuclear plants is reported

  19. Corrosion Product Measurements to ensure integrity of the Steam Generators in Beznau NPP

    International Nuclear Information System (INIS)

    Mailand, Irene; Franz, Patrick; Venz, Hartmut

    2012-09-01

    The Nuclear Power Plant Beznau comprises two identical 380 MWe PWR units with two loops each, commissioned in 1969 and 1971. Westinghouse was responsible for the primary part of the plant and BBC/ABB for the secondary circuit. The original materials used in the secondary systems were made of several copper-based alloys, such as for the Condensers, the Low Pressure Pre-heaters and the Moisture Separator Re-heater. The original Steam Generator Tubes were made of Inconel 600 MA. Regarding its age, the NPP Beznau has to be qualified as an old plant. However, in fact particularly in the last 20 years the plant has undergone an extensive modernisation programme in which about 1.5 billion Swiss Francs have been invested. Important measures were the replacements of the Steam Generators with tubes comprising Inconel 690 TT which was realized at unit 1 in 1993 and at unit 2 in 1999. Copper was completely banished from the secondary system and replaced by stainless and chromium steel. The Condensers were fitted with titanium tubes. The secondary water chemistry had to be changed by these replacements and moved step by step from Low-AVT with a pH of about 9.3 to High-AVT with a pH of 9.8 to 9.9, currently. To ensure the integrity of the new Steam Generators as well as of the whole Secondary System a corrosion product programme was introduced at the end of the Nineties. Several investigations which are performed periodically are represented by analyses of corrosion products, measurements of sludge mass and composition in the Steam Generators, Hide-Out-Return- and mass balance measurements of corrosion products in the whole circuit. Objectives of these investigations are assessments of the efficiency of the water chemistry and trend considerations regarding to the transport of corrosion products and pollutants into the Steam Generator, as well as of the potential danger of deposits and stored or absorbed pollutants. The main target of all measures is to avoid any chemical

  20. Steam generator with perfected dryers

    International Nuclear Information System (INIS)

    Fenet, J.C.

    1987-01-01

    This steam generator has vertically superposed array of steam dryers. These dryers return the steam flow of 180 0 . The return of the water is made by draining channels to the steam production zone [fr

  1. Research program plan: steam generators

    International Nuclear Information System (INIS)

    Muscara, J.; Serpan, C.Z. Jr.

    1985-07-01

    This document presents a plan for research in Steam Generators to be performed by the Materials Engineering Branch, MEBR, Division of Engineering Technology, (EDET), Office of Nuclear Regulatory Research. It is one of four plans describing the ongoing research in the corresponding areas of MEBR activity. In order to answer the questions posed, the Steam Generator Program has been organized with the three elements of non-destructive examination; mechanical integrity testing; and corrosion, cleaning and decontamination

  2. Steam generator tube failures

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Shah, V.N.; Ward, L.W.; Ellison, P.G.

    1996-04-01

    A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service

  3. Regulation of ageing steam generators

    International Nuclear Information System (INIS)

    Jarman, B.L.; Grant, I.M.; Garg, R.

    1998-01-01

    Recent years have seen leaks and shutdowns of Canadian CANDU plants due to steam generator tube degradation by mechanisms including stress corrosion cracking, fretting and pitting. Failure of a single steam generator tube, or even a few tubes, would not be a serious safety related event in a CANDU reactor. The leakage from a ruptured tube is within the makeup capacity of the primary heat transport system, so that as long as the operator takes the correct actions, the off-site consequences will be negligible. However, assurance that no tubes deteriorate to the point where their integrity could be seriously breached as result of potential accidents, and that any leakage caused by such an accident will be small enough to be inconsequential, can only be obtained through detailed monitoring and management of steam generator condition. This paper presents the AECB's current approach and future regulatory directions regarding ageing steam generators. (author)

  4. Structural integrity assessment of steam generator tubes deteriorated through primary water stress corrosion cracking in transition region of tube expansion

    International Nuclear Information System (INIS)

    Silveira, Helvecio Carlos Klinke da

    2002-01-01

    In PWR plants, steam generator tube degradation has been one of the most important economical concerns, besides causing operational safety problems. In this work, a survey of steam generator tube degradation modes is done. Degradation mechanisms and influence factors are introduced and discussed. The importance of stress corrosion cracking, especially in transition region of tube expansion zone, is underlined. The actual steam generator tube plugging criteria are conservative. Proposed alternative criteria are introduced and discussed. Distinction is done to structural integrity assessment of defective tubes. Real data of tube defect indications of axial cracks in expansion transition zone due to primary water stress corrosion cracking are used in analysis. Results allow discussing application aspects of deterministic and probabilistic criteria on structural integrity assessment of tubes with defect indications. Applied models are specifics, but the application of concept may be extended to other steam generator tube degradation modes. (author)

  5. French steam generator

    International Nuclear Information System (INIS)

    Remond, A.

    1986-01-01

    After recalling the potential damage mode of tubes of steam generator, the author recalls the safety criteria used in France. The improvements and the process of damage prejudice and reparation for tubular bundle are presented [fr

  6. Steam generator water lancing

    International Nuclear Information System (INIS)

    Kamler, F.; Schneider, W.

    1992-01-01

    The tubesheet and tube support plate deposits in CANDU steam generators are notable for their hardness. Also notable is the wide variety of steam generator access situations. Because of the sludge hardness and the difficulty of the access, traditional water lancing processes which directed jets from the central tube free lane or from the periphery of the bundle have proven unsuitable. This has led to the need for some very unique inter tube water lancing devices which could direct powerful water jets directly onto the deposits. This type of process was applied to the upper broached plates of the Bruce A steam generators, which had become severely blocked. It has since been applied to various other steam generator situations. This paper describes the flexlance equipment development, qualification, and performance in the various CANDU applications. 4 refs., 2 tabs., 7 figs

  7. Integration of steam injection and inlet air cooling for a gas turbine generation system

    International Nuclear Information System (INIS)

    Wang, F.J.; Chiou, J.S.

    2004-01-01

    The temperature of exhaust gases from simple cycle gas turbine generation sets (GENSETs) is usually very high (around 500 deg. C), and a heat recovery steam generator (HRSG) is often used to recover the energy from the exhaust gases and generate steam. The generated steams can be either used for many useful processes (heating, drying, separation etc.) or used back in the power generation system for enhancing power generation capacity and efficiency. Two well-proven techniques, namely steam injection gas turbine (STIG) and inlet air cooling (IAC) are very effective features that can use the generated steam to improve the power generation capacity and efficiency. Since the energy level of the generated steam needed for steam injection is different from that needed by an absorption chiller to cool the inlet air, a proper arrangement is required to implement both the STIG and the IAC features into the simple cycle GENSET. In this study, a computer code was developed to simulate a Tai power's Frame 7B simple cycle GENSET. Under the condition of local summer weather, the benefits obtained from the system implementing both STIG and IAC features are more than a 70% boost in power and 20.4% improvement in heat rate

  8. Liquid metal steam generator

    International Nuclear Information System (INIS)

    Wolowodiuk, W.

    1975-01-01

    A liquid metal heated steam generator is described which in the event of a tube failure quickly exhausts out of the steam generator the products of the reaction between the water and the liquid metal. The steam is generated in a plurality of bayonet tubes which are heated by liquid metal flowing over them between an inner cylinder and an outer cylinder. The inner cylinder extends above the level of liquid metal but below the main tube sheet. A central pipe extends down into the inner cylinder with a centrifugal separator between it and the inner cylinder at its lower end and an involute deflector plate above the separator so that the products of a reaction between the liquid metal and the water will be deflected downwardly by the deflector plate and through the separator so that the liquid metal will flow outwardly and away from the central pipe through which the steam and gaseous reaction products are exhausted. (U.S.)

  9. Consolidated nuclear steam generator

    International Nuclear Information System (INIS)

    Jabsen, F.S.; Schluderberg, D.C.; Paulson, A.E.

    1978-01-01

    An improved system of providing power has a unique generating means for nuclear reactors with a number of steam generators in the form of replaceable modular units of the expendable type to attain the optimum in effective and efficient vaporization of fluid during the generating power. The system is most adaptable to undrground power plants and marine usage

  10. Steam Generator Maintenance Measures as Part of an Integrated Management in PWRs

    International Nuclear Information System (INIS)

    Weiss, S.; Drexler, A.; Fandrich, J.

    2012-01-01

    The Steam generator condition is a key factor for plant performance, high plant availability, possible life time extension and plant safety. Its major safety function is to act as a barrier between the radioactive primary side and the non-radioactive secondary side of pressurized water reactors. Any degradation mechanism, which impairs this barrier function, is a significant safety concern. The main reason for SG tube failure is known to be the accumulation of deposits contributing to formation of local aggressive conditions. Furthermore deposits on primary as well as secondary side of SG tubes reduce the heat transfer performance. A SG cleanliness management program is therefore mandatory to ensure high plant performance regarding efficiency as well as component integrity. Cleaning measures of steam generator are essential parts of the cleanliness management program. Mechanical cleaning, e.g. tubesheet and inner bundle lancing or upper bundle flushing are efficient methods for removal of local loose deposits. But a chemical cleaning is the only method to remove deposits from the complete SG. AREVA is providing with its C 3 (customized chemical cleaning) concept a tool box of chemical cleaning methods, to adapt to plant specific needs and requirements. (author)

  11. Integration between direct steam generation in linear solar collectors and supercritical carbon dioxide Brayton power cycles

    OpenAIRE

    Coco Enríquez, Luis; Muñoz Antón, Javier; Martínez-Val Peñalosa, José María

    2015-01-01

    Direct Steam Generation in Parabolic Troughs or Linear Fresnel solar collectors is a technology under development since beginning of nineties (1990's) for replacing thermal oils and molten salts as heat transfer fluids in concentrated solar power plants, avoiding environmental impacts. In parallel to the direct steam generation technology development, supercritical Carbon Dioxide Brayton power cycles are maturing as an alternative to traditional Rankine cycles for increasing net plant efficie...

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  13. Nuclear steam generator tubesheet shield

    International Nuclear Information System (INIS)

    Nickerson, J.H.D.; Ruhe, A.

    1982-01-01

    The invention involves improvements to a nuclear steam generator of the type in which a plurality of U-shaped tubes are connected at opposite ends to a tubesheet and extend between inlet and outlet chambers, with the steam generator including an integral preheater zone adjacent to the downflow legs of the U-shaped tubes. The improvement is a thermal shield disposed adjacent to an upper face of the tubesheet within the preheater zone, the shield including ductile cladding material applied directly to the upper face of the tubesheet, with the downflow legs of the U-shaped tubes extending through the cladding into the tubesheet

  14. Structural integrity assessments of steam generator tubes using the FAD methodology

    Energy Technology Data Exchange (ETDEWEB)

    Bergant, Marcos A., E-mail: marcos.bergant@cab.cnea.gov.ar [Gerencia CAREM, Centro Atómico Bariloche (CNEA), Av. Bustillo 9500, San Carlos de Bariloche 8400 (Argentina); Yawny, Alejandro A., E-mail: yawny@cab.cnea.gov.ar [División Física de Metales, Centro Atómico Bariloche (CNEA)/CONICET, Av. Bustillo 9500, San Carlos de Bariloche 8400 (Argentina); Perez Ipiña, Juan E., E-mail: juan.perezipina@fain.uncoma.edu.ar [Grupo Mecánica de Fractura, Universidad Nacional del Comahue/CONICET, Buenos Aires 1400, Neuquén 8300 (Argentina)

    2015-12-15

    Highlights: • The Failure Assessment Diagram (FAD) is used to assess cracked steam generator tubes. • Typical loading conditions and reported tensile and fracture properties are used. • The FAD is capable to predict the failure mode for different cracks and loads. • The FAD can be used to reduce the conservatism of the current plugging criteria. • Appropriate tensile and fracture properties at operating conditions are required. - Abstract: Steam generator tubes (SGTs) represents up to 60% of the total primary pressure retaining boundary area of a nuclear power plant. They have been found susceptible to diverse degradation mechanisms during service. Due to the significance of a SGT failure on the plant safe operation, nuclear regulatory authorities have established tube plugging or repairing criteria which are based on the defect depth. The widespreadly used “40% criterion” proposed in the 70s is an example whose use is still recommended in the last editions of the ASME Boiler and Pressure Vessel Code. In the present work, an alternative, more realistic and less conservative methodology for SGT integrity evaluation is proposed. It is based on the Failure Assessment Diagram (FAD) and takes advantage of the recent developments in non-destructive techniques which allow a more comprehensive characterization of tube defects, i.e., depth, length, orientation and type. The proposed approach has been applied to: the study of the influence of primary and secondary stresses on tube integrity; the prediction of failure mode (i.e., ductile fracture or plastic collapse) of defective SGTs for varied crack geometries and loading conditions; the analysis of the sensibility of tensile and fracture properties with temperature. The potentiality of the FAD as a comprehensive methodology for predicting the failure loads and failure modes of flawed SGTs is highlighted.

  15. Steam generator tube integrity program. Semiannual report, August 1995--March 1996

    International Nuclear Information System (INIS)

    Diercks, D.R.; Bakhtiari, S.; Chopra, O.K.

    1997-04-01

    This report summarizes work performed by Argonne National Laboratory on the Steam Generator Tube Integrity Program from the inception of that program in August 1995 through March 1996. The program is divided into five tasks, namely (1) Assessment of Inspection Reliability, (2) Research on ISI (in-service-inspection) Technology, (3) Research on Degradation Modes and Integrity, (4) Development of Methodology and Technical Requirements for Current and Emerging Regulatory Issues, and (5) Program Management. Under Task 1, progress is reported on the preparation of and evaluation of nondestructive evaluation (NDE) techniques for inspecting a mock-up steam generator for round-robin testing, the development of better ways to correlate burst pressure and leak rate with eddy current (EC) signals, the inspection of sleeved tubes, workshop and training activities, and the evaluation of emerging NDE technology. Under Task 2, results are reported on closed-form solutions and finite element electromagnetic modeling of EC probe response for various probe designs and flaw characteristics. Under Task 3, facilities are being designed and built for the production of cracked tubes under aggressive and near-prototypical conditions and for the testing of flawed and unflawed tubes under normal operating, accident, and severe accident conditions. In addition, crack behavior and stability are being modeled to provide guidance on test facility design, to develop an improved understanding of the expected rupture behavior of tubes with circumferential cracks, and to predict the behavior of flawed and unflawed tubes under severe accident conditions. Task 4 is concerned with the cracking and failure of tubes that have been repaired by sleeving, and with a review of literature on this subject

  16. Structural integrity assessments of steam generator tubes using the FAD methodology

    International Nuclear Information System (INIS)

    Bergant, Marcos A.; Yawny, Alejandro A.; Perez Ipiña, Juan E.

    2015-01-01

    Highlights: • The Failure Assessment Diagram (FAD) is used to assess cracked steam generator tubes. • Typical loading conditions and reported tensile and fracture properties are used. • The FAD is capable to predict the failure mode for different cracks and loads. • The FAD can be used to reduce the conservatism of the current plugging criteria. • Appropriate tensile and fracture properties at operating conditions are required. - Abstract: Steam generator tubes (SGTs) represents up to 60% of the total primary pressure retaining boundary area of a nuclear power plant. They have been found susceptible to diverse degradation mechanisms during service. Due to the significance of a SGT failure on the plant safe operation, nuclear regulatory authorities have established tube plugging or repairing criteria which are based on the defect depth. The widespreadly used “40% criterion” proposed in the 70s is an example whose use is still recommended in the last editions of the ASME Boiler and Pressure Vessel Code. In the present work, an alternative, more realistic and less conservative methodology for SGT integrity evaluation is proposed. It is based on the Failure Assessment Diagram (FAD) and takes advantage of the recent developments in non-destructive techniques which allow a more comprehensive characterization of tube defects, i.e., depth, length, orientation and type. The proposed approach has been applied to: the study of the influence of primary and secondary stresses on tube integrity; the prediction of failure mode (i.e., ductile fracture or plastic collapse) of defective SGTs for varied crack geometries and loading conditions; the analysis of the sensibility of tensile and fracture properties with temperature. The potentiality of the FAD as a comprehensive methodology for predicting the failure loads and failure modes of flawed SGTs is highlighted.

  17. Steam generator collector integrity of WWER-1000 reactors. IAEA extrabudgetary programme on the safety of WWER NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Lin, C.; Strupczewski, A. [International Atomic Energy Agency, Vienna (Austria)

    1995-12-31

    At the Consultants` Meeting on `The Safety of WWER-1000 Model 320 Nuclear Power Plants` organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of WWER-1000 steam generator integrity was identified as an important issue of safety concern. Considering the safety importance of this issue, a Consultants` Meeting on `The Steam Generator Integrity of WWER-1000 Nuclear Power Plants` was convened in Vienna in May 1993, attended by 15 international experts in the area to compile information on the steam generator operating experience, deficiencies and corrective measures implemented and planned. In order to also include information from the main designer OKB Gidropress and to finalize the meeting report the IAEA convened a second meeting on the issue on 23-27 November 1993. The present paper summarizes the information and conclusions from those meetings.

  18. Steam generator collector integrity of WWER-1000 reactors. IAEA extrabudgetary programme on the safety of WWER NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Lin, C; Strupczewski, A [International Atomic Energy Agency, Vienna (Austria)

    1996-12-31

    At the Consultants` Meeting on `The Safety of WWER-1000 Model 320 Nuclear Power Plants` organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of WWER-1000 steam generator integrity was identified as an important issue of safety concern. Considering the safety importance of this issue, a Consultants` Meeting on `The Steam Generator Integrity of WWER-1000 Nuclear Power Plants` was convened in Vienna in May 1993, attended by 15 international experts in the area to compile information on the steam generator operating experience, deficiencies and corrective measures implemented and planned. In order to also include information from the main designer OKB Gidropress and to finalize the meeting report the IAEA convened a second meeting on the issue on 23-27 November 1993. The present paper summarizes the information and conclusions from those meetings.

  19. Steam generators and furnaces

    Energy Technology Data Exchange (ETDEWEB)

    Swoboda, E

    1978-04-01

    The documents published in 1977 in the field of steam generators for conventional thermal power plants are classified according to the following subjects: power industry and number of power plants, planning and operation, design and construction, furnaces, environmental effects, dirt accumulation and corrosion, conservation and scouring, control and automation, fundamental research, and materials.

  20. Steam generator arrangement

    International Nuclear Information System (INIS)

    Ssinegurski, E.

    1981-01-01

    A steam flow path arrangement for covering the walls of the rear gas pass of a steam generator is disclosed. The entire flow passes down the sidewalls with a minor portion then passing up through the rear wall to a superheater inlet header at an intermediate elevation. The major portion of the flow passes up the front wall and through hanger tubes to a roof header. From there the major portion passes across the roof and down the rear wall to the superheater inlet header at the intermediate elevation

  1. Development of integral type forgings for steam generator of nuclear power plant

    International Nuclear Information System (INIS)

    Suzuki, Komei; Sato, Ikuo; Murai, Etsuo

    1992-01-01

    The use of integral type steel forgings for the construction of pressure vessel enhances the structural integrity of components and makes the fabrication of components and the execution of in-service inspection (ISI) easier than those fabrication from plate and casting materials. Such steel forgings have been realized for steam generator (SG) for nuclear power plant as follows : (1) Forged shell ring : change from welding fabrication of formed plates to forging ; (2) Forged conical shell ring : ditto ; (3) Forged head integral with nozzles (s) : (i) Primary head : change from casting to forging ; (ii) Secondary head : change from welding fabrication of formed plates to forging. These steel forgings have been realized by recent development in manufacturing technologies, such as steel making, forging processes and heat treatment which are vital to the quality of steel forgings. Some examples of recent typical high quality steel forgings developed by the Japan Steel Works, Ltd. (JSW) are introduced, and the main points of the manufacturing technique and the quality attained are also described. (author)

  2. Assessment of the integrity of degraded steam generator tube by the use of heterogeneous finite element method

    International Nuclear Information System (INIS)

    Duan, X.; Kozluk, M.; Pagan, S.; Mills, B.

    2006-01-01

    Steam generator tubes at Ontario Power Generation (OPG) have been experiencing a variety of degradations such as pitting, fretting wear, erosion-corrosion, thinning and denting. To assist with steam generator life cycle management, OPG has developed Fitness-For-Service Guidelines (FFSG) for steam generator tubes. The FFSG are intended to provide standard acceptance criteria and evaluation procedures for assessing the condition of steam generator tubes for structural integrity, operational leak rate, and consequential leakage during an upset or abnormal event. Based on inspection results in conjunction with representative, postulated distributions of flaws in the un-inspected tubes, the FFSG provide an acceptable method of satisfying the intent of CSA-N285.4 and justifying the continued operation of degraded steam generator tubes. Some non-mandatory empirical axial and circumferential flaw models are also provided in the FFSG for structural integrity assessments. The test data from the OPG Steam Generator Tube Test Program (SGTTP) showed that the FFSG axial flaw model is conservative for a wide range of defect morphologies. A defect-specific axial flaw model was proposed for lattice-bar fret defects in I800 tubes by utilizing the SGTTP database of extensive test results. A defect-specific flaw model for outer diameter (OD) pitting and inner diameter (ID) intergranular attack in Monel 400 tubes was also developed using the SGTTP test data. More tests have been scheduled to support the development of defect specific models for axial flaws (OD cracks or ID laps) in Monel 400 and to supplement the database for Monel 400 pits. This paper explores the use of simulated testing for use in developing defect specific flaw models to reduce the amount of expensive tests. The Heterogeneous Finite Element Model (HFEM) has been developed and successfully applied to predict the failure behaviour of ductile metals under various deformation modes, i.e. plane stress, plane strain and

  3. Steam generator sludge removal apparatus

    International Nuclear Information System (INIS)

    Schafer, B.W.; Werner, C.E.; Klahn, F.C.

    1992-01-01

    The present invention relates to equipment for cleaning steam generators and in particular to a high pressure fluid lance for cleaning sludge off the steam generator tubes away from an open tube lane. 6 figs

  4. Steam generator operation and maintenance

    International Nuclear Information System (INIS)

    Lee, C.K.

    2004-01-01

    Corrosion of steam generator tube has resulted in the need for extensive repair and replacement of steam generators. Over the past two decades, steam generator problems in the United States were viewed to be one of the most significant contributor to lost generation in operating PWR plants. When the SGOG-I (Steam Generator Owners Groups) was formed in early 1977, denting was responsible for almost 90% of the tube plugging. By the end of 1982, this figure was reduced to less than 2%. During the existence of SGOG-II (from 1982 to 1986), IGA/SCC (lntergranular Attack/Stress Corrosion Cracking) in the tube sheet, primary side SCC, pitting, and fretting surfaced as the primary causes of tube degradation. Although significant process has been made with wastage and denting, the utilities experience shows that the percentage of reactors plugging tubes and the percentage of tubes being plugged each year has remained relatively constant. The diversity of the damage mechanisms means that no one solution is likely to resolve all problems. The task of maintaining steam generator integrity continues to be formidable and challenging. As the older problems were brought under control, many new problems emerged. SGOG-II (Steam Generator Owners Group program from 1982 to 1986) has focused on these problem areas such as tube stress corrosion cracking (SCC) and intergranular attack (IGA) in the open tube sheet crevice, primary side tube cracking, pitting in the lower span, and tube fretting in preheated section and anti-vibration bar (AVB) locations. Primary Water Stress Corrosion Cracking (PWSCC) in the tube to tubesheet roll transition has been a wide spread problem in the Recirculation Steam Generators (RSG) during this period. Although significant progress has been made in resolving this problem, considerable work still remains. One typical problem in the Once Through Steam Generator (OTSG) was the tube support plate broached hole fouling which affects the OTSG steam generating

  5. Vertical steam generator

    International Nuclear Information System (INIS)

    Cuda, F.; Kondr, M.; Kresta, M.; Kusak, V.; Manek, O.; Turon, S.

    1982-01-01

    A vertical steam generator for nuclear power plants and dual purpose power plants consists of a cylindrical vessel in which are placed heating tubes in the form upside-down U. The heating tubes lead to the jacket of the cylindrical collector placed in the lower part of the steam generator perpendicularly to its vertical axis. The cylindrical collector is divided by a longitudinal partition into the inlet and outlet primary water sections of the heating tubes. One ends of the heating tube leads to the jacket of the collector for primary water feeding and the second ends of the heating tubes into the jacket of the collector which feeds and offtakes primary water from the heating tubes. (B.S.)

  6. Performance analysis of an Integrated Solar Combined Cycle using Direct Steam Generation in parabolic trough collectors

    International Nuclear Information System (INIS)

    Montes, M.J.; Rovira, A.; Munoz, M.; Martinez-Val, J.M.

    2011-01-01

    Highlights: → Solar hybridization improves the performance of CCGT in a very hot and dry weather. → The scheme analyzed is a DSG parabolic trough field coupled to the Rankine cycle. → An annual simulation has been carried out for two locations: Almeria and Las Vegas. → Economical analysis shows that this scheme is a cheaper way to exploit solar energy. → For that, solar hybridization must be limited to a small fraction of the CCGT power. - Abstract: The contribution of solar thermal power to improve the performance of gas-fired combined cycles in very hot and dry environmental conditions is analyzed in this work, in order to assess the potential of this technique, and to feature Direct Steam Generation (DSG) as a well suited candidate for achieving very good results in this quest. The particular Integrated Solar Combined Cycle (ISCC) power plant proposed consists of a DSG parabolic trough field coupled to the bottoming steam cycle of a Combined Cycle Gas Turbine (CCGT) power plant. For this analysis, the solar thermal power plant performs in a solar dispatching mode: the gas turbine always operates at full load, only depending on ambient conditions, whereas the steam turbine is somewhat boosted to accommodate the thermal hybridization from the solar field. Although the analysis is aimed to studying such complementary effects in the widest perspective, two relevant examples are given, corresponding to two well-known sites: Almeria (Spain), with a mediterranean climate, and Las Vegas (USA), with a hot and dry climate. The annual simulations show that, although the conventional CCGT power plant works worse in Las Vegas, owing to the higher temperatures, the ISCC system operates better in Las Vegas than in Almeria, because of solar hybridization is especially well coupled to the CCGT power plant in the frequent days with great solar radiation and high temperatures in Las Vegas. The complementary effect will be clearly seen in these cases, because the thermal

  7. Third international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues.

  8. Third international seminar on horizontal steam generators

    International Nuclear Information System (INIS)

    1995-01-01

    The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues

  9. Third international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues.

  10. Safety Evaluation for IHTS Integrity due to the Steam Generator Sodium-Water Reaction Event in the PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang-Jun; Lee, Kwi Lim; Ha, Kwi-Seok; Lee, Seung Won; Jeong, Taekyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper, the integrity of the IHTS and SG by the SWR event are evaluated using the SWAAMII code. A sodium has a chemical characteristics to rigorously react the water or steam and produce the high pressure waves and high temperature reaction heat. It has an excellent characteristics as a reactor coolant. But, there is an event to be considered in the sodium cooled fast reactor design. The Sodium-Water Reaction (SWR) event can be occurred by the water or steam leaks due to the break of the steam generator tubes. The propagated high pressure waves threathen the structural integrity of the affected Intermediate Heat Transport System (IHTS) and steam generator. If the IHTS pipes are failed, the sodium of the IHTS can be released to the containment building. To the peak pressure point of view, it is performed to evaluate the integrity of the major components due to the SWR event in the SG. The generated peak pressures due to the five SG tubes simultaneous break event are within the range of the design pressure for the SG, IHX and IHTS including the related pipes.

  11. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

    The steam generator of the PWR operates according to the principle of natural circulation. It consists of a U-shaped tube bundle whose free ends are welded to a bottom plate. The tube bundle is surrounded by a cylinder jacket which has slots closely above the bottom or tube plate. The feed water mixed with boiling water enters the tube bundle through these slots. Because of its buoyancy, the steam-water mixture flows upwards. Below the tube plate there are chambers for distributing and collecting pressurized water separated by means of a partition wall. By omitting some tubes, a free alloy is created so that the tubes in the center get sufficient water, too. By asymmetrical arrangement of the partition wall it is further possible to limit the tube alloy only to the inlet side for pressurized water. The flow over the tube plate is thus improved on the inlet side. (DG) [de

  12. NRC integrated program for the resolution of Unresolved Safety Issues A-3, A-4 and A-5 regarding steam generator tube integrity: Final report

    International Nuclear Information System (INIS)

    1988-09-01

    This report presents the results of the NRC integrated program for the resolution of Unresolved Safety Issues (USIs) A-3, A-4, and A-5 regarding steam generator tube integrity. A generic risk assessment is provided and indicates that risk from steam generator tube rupture (SGTR) events is not a significant contributor to total risk at a given site, nor to the total risk to which the general public is routinely exposed. This finding is considered to be indicative of the effectiveness of licensee programs and regulatory requirements for ensuring steam generator tube integrity in accordance with 10 CFR 50, Appendices A and B. This report also identifies a number of staff-recommended actions that the staff finds can further improve the effectiveness of licensee programs in ensuring the integrity of steam generator tubes and in mitigating the consequences of an SGTR. As part of the integrated program, the staff issued Generic Letter 85-02 encouraging licensees of pressurized water reactors (PWRs) to upgrade their programs, as necessary, to meet the intent of the staff-recommended actions; however, such actions do not constitute NRC requirements. In addition, this report describes a number of ongoing staff actions and studies involving steam generator issues which are being pursued to provide added assurance that risk from SGTR events will continue to be small. 146 refs., 5 figs., 11 tabs

  13. Steam generator waterlancing at DNGS

    International Nuclear Information System (INIS)

    Seppala, D.; Malaugh, J.

    1995-01-01

    Darlington Nuclear Generating Station (DNGS) is a four 900 MW Unit nuclear station forming part of the Ontario Hydro East System. There are four identical steam generators(SGs) per reactor unit. The Darlington SGs are vertical heat exchangers with an inverted U-tube bundle in a cylindrical shell. The DNGS Nuclear Plant Life Assurance Group , a department of DNGS Engineering Services have taken a Proactive Approach to ensure long term SG integrity. Instead of waiting until the tubesheets are covered by a substantial and established hard deposit; DNGS plan to clean each steam generator's tubesheet, first half lattice tube support assembly and bottom of the thermal plate every four years. The ten year business plan provides for cleaning and inspection to be conducted on all four SGs in each unit during maintenance outages (currently scheduled for every four years)

  14. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1984-10-01

    A review of the performance of steam generator tubes in 116 water-cooled nuclear power reactors showed that tubes were plugged at 54 (46 percent) of the reactors. The number of tubes removed from service decreased from 4 692 (0.30 percent) in 1981 to 3 222 (0.20 percent) in 1982. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that have used only volatile treatment, with or without condensate demineralization

  15. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Tapping, R.L.; Stipan, L.

    1992-03-01

    A survey of steam generator operating experience for 1986 has been carried out for 184 pressurized water and pressurized heavy-water reactors, and 1 water-cooled, graphite-moderated reactor. Tubes were plugged at 75 of the reactors (40.5%). In 1986, 3737 tubes were plugged (0.14% of those in service) and 3148 tubes were repaired by sleeving. A small number of reactors accounted for the bulk of the plugged tubes, a phenomenon consistent with previous years. For 1986, the available tubesheet sludge data for 38 reactors has been compiled into tabular form, and sludge/deposit data will be incorporated into all future surveys

  16. Predictions of structural integrity of steam generator tubes under normal operating, accident, and severe accident conditions

    International Nuclear Information System (INIS)

    Majumdar, S.

    1996-09-01

    Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation is confirmed by further tests at high temperatures as well as by finite element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation is confirmed by finite element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure is developed and validated by tests under varying temperature and pressure loading expected during severe accidents

  17. PWR steam generators tube integrity: plugging criteria for PWSCC in roll transition zone

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Cruz, Julio R.B.

    1999-01-01

    One of the most important causes for tube plugging in PWR (Pressurized Water Reactor) steam generators is the degradation mechanism called Primary Water Stress Corrosion Cracking (PWSCC) in roll transition zone (RTZ) near the tubesheet, mainly for Alloy 600 tubes. To avoid an excessive tube plugging, alternative criteria have been developed based on an approach that consists in withdrawing from service any tube containing a defect for which there is a high probability of a critical size under accident conditions to be reached during next operation cycle. Predictions of the number of tubes to be plugged can be done aiming at preventive maintenance and tube repair, and even a steam generator replacement, without a large and non-planned plant outage. This work presents important aspects related to tube plugging criteria for PWSCC in RTZ based on the risk of break after a leak detection. Calculations of allowable crack length and allowable leak rate for a particular situation are also shown. (author)

  18. Spanish approach to research and development applied to steam generator tubes structural integrity and life management

    International Nuclear Information System (INIS)

    Lozano, J.; Bollini, G.J.

    1997-01-01

    The operating experience acquired from certain Spanish Nuclear Power Plant steam generators shows that the tubes, which constitute the second barrier to release of fission products, are susceptible to mechanical damage and corrosion as a result of a variety of mechanisms, among them wastage, pitting, intergranular attack (IGA), stress-corrosion cracking (SCC), fatigue-induced cracking, fretting, erosion/corrosion, support plate denting, etc. These problems, which are common in many plants throughout the world, have required numerous investments by the plants (water treatment plants, replacement of secondary side materials such as condensers and heaters, etc.), have meant costs (operation, inspection and maintenance) and have led to the unavailability of the affected units. In identifying and implementing all these preventive and corrective measures, the Spanish utilities have moved through three successive stages: in the initial stage, the main source of information and of proposals for solutions was the Plant Vendor, whose participation in this respect was based on his own Research and Development programs; subsequently, the Spanish utilities participated jointly in the EPRI Steam Generator Owners Group, collaborating in financing; finally, the Spanish utilities set up their own Steam Generator Research and Development program, while maintaining relations with EPRI programs and those of other countries through information interchange

  19. Steam generator materials

    International Nuclear Information System (INIS)

    Kim, Joung Soo; Han, J. H.; Kim, H. P.; Lim, Y. S.; Lee, D. H.; Suh, J. H.; Hwang, S. S.; Hur, D. H.; Kim, D. J.; Kim, Y. H.

    2002-05-01

    In order to keep the nuclear power plant(NPP)s safe and increase their operating efficiency, axial stress corrosion cracking(SCC)(IGA/IGSCC, PWSCC, PbSCC) test techniques were developed and SCC property data of the archive steam generator tubing materials having been used in nuclear power plants operating in Korea were produced. The data obtained in this study were data-based, which will be used to clarify the damage mechanisms, to operate the plants safely, and to increase the lifetime of the tubing. In addition, the basic technologies for the improvement of the SCC property of the tubing materials, for new SCC inhibition, for damaged tube repair, and for manufacturing processes of the tubing were developed. In the 1 phase of this long term research, basic SCC test data obtained from the archive steam generator tubing materials used in NPPs operating in Korea were established. These basic technologies developed in the 1 phase will be used in developing process optimization during the 2 phase in order to develop application technologies to the field nuclear power plants

  20. Operating experiences with 1 MW steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Sano, A; Kanamori, A; Tsuchiya, T

    1975-07-01

    1 MW steam generator, which was planned as the first stage of steam generator development in Power Reactor and Nuclear Fuel Corp. (PNC) in Japan, is a single-unit, once-through, integrated shell and tube type with multi-helical coil tubes. It was completed in Oarai Engineering Center of PNC in March of 1971, and the various performance tests were carried out up to April, 1972. After the dismantle of the steam generator for structural inspection and material test, it was restored with some improvements. In this second 1 MW steam generator, small leak occurred twice during normal operation. After repairing the failure, the same kind of performance tests as the first steam generator were conducted in order to verify the thermal insulation effect of argon gas in downcomer zone from March to June, 1974. In this paper the above operating experiences were presented including the outline of some performance test results. (author)

  1. Circumferential cracking of steam generator tubes

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    1997-04-01

    On April 28, 1995, the U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 95-03, open-quote Circumferential Cracking of Steam Generator Tubes.close-quote GL 95-03 was issued to obtain information needed to verify licensee compliance with existing regulatory requirements regarding the integrity of steam generator tubes in domestic pressurized-water reactors (PWRs). This report briefly describes the design and function of domestic steam generators and summarizes the staff's assessment of the responses to GL 95-03. The report concludes with several observations related to steam generator operating experience. This report is intended to be representative of significant operating experience pertaining to circumferential cracking of steam generator tubes from April 1995 through December 1996. Operating experience prior to April 1995 is discussed throughout the report, as necessary, for completeness

  2. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1982-04-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1980. Tube defects occurred at 38% of the 97 reactors surveyed. This is a marginal improvement over 1979 when defects occurred at 41% of the reactors. The number of failed tubes was also lower, 0.14% of the tubes in service in 1980 compared with 0.20% of those in service in 1979. Analysis of the causes of these failures indicates that stress corrosion cracking was the leading failure mechanism. Reactors that used all-volatile treatment of secondary water, with or without full-flow condensate demineralization since start-up showed the lowest incidence of corrosion-related defects

  3. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1983-08-01

    A review of the performance of steam generator tubes in 110 water-cooled nuclear power reactors showed that tubes were plugged at 46 (42 percent) of the reactors. The number of tubes removed from service increased from 1900 (0.14 percent) in 1980 to 4692 (0.30 percent) in 1981. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that used all-volatile treatment since start-up. At one reactor a large number of degraded tubes were repaired by sleeving which is expected to become an important method of tube repair in the future

  4. Steam generator reliability improvement project

    International Nuclear Information System (INIS)

    Blomgren, J.C.; Green, S.J.

    1987-01-01

    Upon successful completion of its research and development technology transfer program, the Electric Power Research Institute's Steam Generator Owners Group (SGOG II) will disband in December 1986 and be replaced in January 1987 by a successor project, the Steam Generator Reliability Project (SGRP). The new project, funded in the EPRI base program, will continue the emphasis on reliability and life extension that was carried forward by SGOG II. The objectives of SGOG II have been met. Causes and remedies have been identified for tubing corrosion problems, such as stress corrosion cracking and pitting, and steam generator technology has been improved in areas such as tube wear prediction and nondestructive evaluation (NDE). These actions have led to improved reliability of steam generators. Now the owners want to continue with a centrally managed program that builds on what has been learned. The goal is to continue to improve steam generator reliability and solve small problems before they become large problems

  5. Steam generator reliability improvement project

    International Nuclear Information System (INIS)

    Blomgren, J.C.; Green, S.J.

    1987-01-01

    Upon successful completion of its research and development technology transfer program, the Electric Power Research Institute's (EPRI's) Steam Generator Owners Group (SGOG II) will disband in December 1986, and be replaced in January 1987, by a successor project, the Steam Generator Reliability Project (SGRP). The new project, funded in the EPRI base program, will continue to emphasize reliability and life extension, which were carried forward by SGOG II. The objectives of SGOG II have been met. Causes and remedies have been identified for tubing corrosion problems such as stress corrosion cracking and pitting, and steam generator technology has been improved in areas such as tube wear prediction and nondestructive evaluation. These actions have led to improved reliability of steam generators. Now the owners want to continue with a centrally managed program that builds on what has been learned. The goal is to continue to improve steam generator reliability and to solve small problems before they become large problems

  6. Failure analysis of retired steam generator tubings

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hong Pyo; Kim, J. S.; Hwang, S. S. and others

    2005-04-15

    Degradation of steam generator leads to forced outage and extension of outage, which causes increase in repair cost, cost of purchasing replacement power and radiation exposure of workers. Steam generator tube rupture incident occurred in Uljin 4 in 2002, which made public sensitive to nuclear power plant. To keep nuclear energy as a main energy source, integrity of steam generator should be demonstrated. Quantitative relationship between ECT(eddy current test) signal and crack size is needed in assesment of integrity of steam generator in pressurized water reactor. However, it is not fully established for application in industry. Retired steam generator of Kori 1 has many kinds of crack such as circumferential and axial primary water stress corrosion crack and outer diameter stress corrosion crack(ODSCC). So, it can be used in qualifying and improving ECT technology and in condition monitoring assesment for crack detected in ISI(in service inspection). In addition, examination of pulled tube of Kori 1 retired steam generator will give information about effectiveness of non welded sleeving technology which was employed to repair defect tubes and remedial action which was applied to mitigate ODSCC. In this project, hardware such as semi hot lab. for pulled tube examination and modification transportation cask for pulled tube and software such as procedure of transportation of radioactive steam generator tube and non-destructive and destructive examination of pulled tube were established. Non-destructive and destructive examination of pulled tubes from Kori 1 retired steam generator were performed in semi hot lab. Remedial actions applied to Kori 1 retired steam generator, PWSCC trend and bulk water chemistry and crevice chemistry in Kori 1 were evaluated. Electrochemical decontamination technology for pulled tube was developed to reduce radiation exposure and enhance effectiveness of pulled tube examination. Multiparameter algorithm developed at ANL, USA was

  7. Intelligent information data base of flow boiling characteristics in once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Inasaka, Fujio; Nariai, Hideki

    1998-01-01

    Valuable experimental knowledge with flow boiling characteristics of the helical-coil type once-through steam generator was converted into an intelligent information data base program. The program was created as a windows application using the Visual Basic. Main functions of the program are as follows: (1) steady state flow boiling analysis of any helical-coil type once-through steam generator, (2) analysis and comparison with the experimental data, (3) reference and graph display of the steady state experimental data, (4) reference of the flow instability experimental data and display of the instability threshold correlated by each parameter, (5) summary of the experimental apparatus. (6) menu bar such as a help and print. In the steady state analysis, the region lengths of subcooled boiling, saturated boiling, and super-heating, and the temperature and pressure distributions etc. for secondary water calculated. Steady state analysis results agreed well with the experimental data, with the exception of the pressure drop at high mass velocity. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized water reactor with helical-coil type steam generator

  8. Steam generators, turbines, and condensers. Volume six

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Volume six covers steam generators (How steam is generated, steam generation in a PWR, vertical U-tube steam generators, once-through steam generators, how much steam do steam generators make?), turbines (basic turbine principles, impulse turbines, reaction turbines, turbine stages, turbine arrangements, turbine steam flow, steam admission to turbines, turbine seals and supports, turbine oil system, generators), and condensers (need for condensers, basic condenser principles, condenser arrangements, heat transfer in condensers, air removal from condensers, circulating water system, heat loss to the circulating water system, factors affecting condenser performance, condenser auxiliaries)

  9. Advanced technologies on steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Sakata, Kaoru; Nakamura, Yuuki [Mitsubishi Heavy Industry Co., Takasago (Japan); Nakamori, Nobuo; Mizutani, Toshiyuki; Uwagawa, Seiichi; Saito, Itaru [Mitsubishi Heavy Industry Co., Kobe (Japan); Matsuoka, Tsuyoshi [Mitsubishi Heavy Industry Co., Yokohama (Japan)

    1997-12-31

    The thermal-hydraulic tests for a horizontal steam generator of a next-generation PWR (New PWR-21) were performed. The purpose of these tests is to understand the thermal-hydraulic behavior in the secondary side of horizontal steam generator during the plant normal operation. A test was carried out with cross section slice model simulated the straight tube region. In this paper, the results of the test is reported, and the effect of the horizontal steam generator internals on the thermalhydraulic behavior of the secondary side and the circulation characteristics of the secondary side are discussed. (orig.). 3 refs.

  10. Advanced technologies on steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Sakata, Kaoru; Nakamura, Yuuki [Mitsubishi Heavy Industry Co., Takasago (Japan); Nakamori, Nobuo; Mizutani, Toshiyuki; Uwagawa, Seiichi; Saito, Itaru [Mitsubishi Heavy Industry Co., Kobe (Japan); Matsuoka, Tsuyoshi [Mitsubishi Heavy Industry Co., Yokohama (Japan)

    1998-12-31

    The thermal-hydraulic tests for a horizontal steam generator of a next-generation PWR (New PWR-21) were performed. The purpose of these tests is to understand the thermal-hydraulic behavior in the secondary side of horizontal steam generator during the plant normal operation. A test was carried out with cross section slice model simulated the straight tube region. In this paper, the results of the test is reported, and the effect of the horizontal steam generator internals on the thermalhydraulic behavior of the secondary side and the circulation characteristics of the secondary side are discussed. (orig.). 3 refs.

  11. CANDU steam generator life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

    1998-01-01

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. The research and development program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factory. (author)

  12. Exergy Steam Drying and Energy Integration

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Prem; Muenter, Claes (Exergy Engineering and Consulting, SE-417 55 Goeteborg (Sweden)). e-mail: verma@exergyse.com

    2008-10-15

    Exergy Steam Drying technology has existed for past 28 years and many new applications have been developed during this period. But during past few years the real benefits have been exploited in connection with bio-fuel production and energy integration. The steam dryer consists of a closed loop system, where the product is conveyed by superheated and pressurised carrier steam. The carrier steam is generated by the water vapours from the product being dried, and is indirectly superheated by another higher temperature energy source such as steam, flue gas, thermal oil etc. Besides the superior heat transfer advantages of using pressurised steam as a drying medium, the energy recovery is efficient and simple as the recovered energy (80-90%) is available in the form of steam. In some applications the product quality is significantly improved. Examples presented in this paper: Bio-Combine for pellets production: Through integration of the Exergy Steam Dryer for wood with a combined heat and power (CHP) plant, together with HP steam turbine, the excess carrier steam can be utilised for district heating and/or electrical power production in a condensing turbine. Bio-ethanol production: Both for first and second generation of ethanol can the Exergy process be integrated for treatment of raw material and by-products. Exergy Steam Dryer can dry the distillers dark grains and solubles (DDGS), wood, bagasse and lignin. Bio-diesel production: Oil containing seeds and fruits can be treated in order to improve both the quality of oil and animal feed protein, thus minimizing further oil processing costs and increasing the sales revenues. Sewage sludge as bio-mass: Municipal sewage sludge can be considered as a renewable bio-fuel. By drying and incineration, the combustion heat value of the sludge is sufficient for the drying process, generation of electrical energy and production of district heat. Keywords; Exergy, bio-fuel, bio-mass, pellets, bio-ethanol, biodiesel, bio

  13. Steam generator thermal sleeve reconstruction

    Energy Technology Data Exchange (ETDEWEB)

    Caton, E.; Askari, A.; Volder, P. [Babcock and Wilcox Canada Ltd., Cambridge, Ontario (Canada)]. E-mail: eecaton@babcock.com

    2003-07-01

    'Full text:' Successful implementation of a physically difficult repair program requires collaboration of the design and construction functions of an organization to ensure that goals are shared and rework or on-the-fly design changes are not required. Furthermore, in a nuclear facility this collaboration results in the optimal safety condition as dose uptake is minimized with a well planned job. The replacement of the degraded thermal sleeves in the Pickering A Steam Generator feedwater nozzles posed this type of problem. The project may be summarized as follows: i) problem analysis, ii) identification of design parameters and limitations, iii) integration of field engineering and design engineering solutions, iv) installation. Integration of the design engineering and field engineering design parameters ensured that the most effective solution was implemented. (author)

  14. STEAM GENERATOR TUBE INTEGRITY ANALYSIS OF A TOTAL LOSS OF ALL HEAT SINKS ACCIDENT FOR WOLSONG NPP UNIT 1

    Directory of Open Access Journals (Sweden)

    HEOK-SOON LIM

    2014-02-01

    Full Text Available A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS and the steam generator (SG secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

  15. Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo [Korea Htydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Kim, Seoungrae [Nuclear Engineering Service and Solution, Daejeon (Korea, Republic of)

    2014-02-15

    A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

  16. Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

    International Nuclear Information System (INIS)

    Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo; Kim, Seoungrae

    2014-01-01

    A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident

  17. Steam generator thermal-hydraulics

    International Nuclear Information System (INIS)

    Inch, W.W.; Scott, D.A.; Carver, M.B.

    1980-01-01

    This paper discusses a code for detailed numerical modelling of steam generator thermal-hydraulics, and describes related experimental programs designed to promote in-depth understanding of three-dimensional two-phase flow. (auth)

  18. French steam generator design developments

    International Nuclear Information System (INIS)

    Ginier, R.; Campan, J.L.; Pontier, M.; Leridon, A.; Remond, A.; Castello, G.; Holcblat, A.; Paurobally, H.

    1986-01-01

    From the outset of the French nuclear power program, a significant R and D effort has been invested in improvement of the design and operation of Pressurized Water Reactors including a special committment to improving steam generators. The steam generator enhancement program has spawned a wide variety of specific R and D resources, e.g., low temperature hydraulic models for investigation of areas with single-phase flow, and freon-filled models for simulation of areas of steam generators experiencing two-phase flow (tube bundles and moisture separators). For the moisture separators, a large scale research program using freon-filled models and highly sophisticated instrumentation was used. Tests at reactor sites during startup of both 900 MWe and 1300 MWe have been used to validate the assumptions made on the basis of loop tests. These tests also demonstrated the validity of using freon to simulate two-phase flow conditions. The wealth of knowledge accumulated by the steam generator R and D program has been used to develop a new design of steam generators for the N4 plants. The current R and D effort is aimed at qualifying the N4 steam generator model and developing more comprehensive models. One prong of the R and D effort is the Megeve program. Megeve is a 25 MW steam generator which simulates operating conditions of the N4 model. The other prong is Clotaire, a freon-filled steam generator model which will be used to qualify thermal/hydraulic design codes used for multidimensional calculations for design of tube bundles

  19. Corrosion Evaluation and Corrosion Control of Steam Generators

    International Nuclear Information System (INIS)

    Maeng, W. Y.; Kim, U. C.; Sung, K. W.; Na, J. W.; Lee, Y. H.; Lee, D. H.; Kim, K. M.

    2008-06-01

    Corrosion damage significantly influences the integrity and efficiency of steam generator. Corrosion problems of steam generator are unsolved issues until now even though much effort is made around world. Especially the stress corrosion cracking of heat exchange materials is the first issue to be solved. The corrosion protection method of steam generator is important and urgent for the guarantee of nuclear plant's integrity. The objectives of this study are 1) to evaluate the corrosion properties of steam generator materials, 2) to optimize the water chemistry of steam generator and 3) to develop the corrosion protection method of primary and secondary sides of steam generator. The results will be reflected to the water chemistry guideline for improving the integrity and efficiency of steam generator in domestic power plants

  20. Corrosion Evaluation and Corrosion Control of Steam Generators

    Energy Technology Data Exchange (ETDEWEB)

    Maeng, W. Y.; Kim, U. C.; Sung, K. W.; Na, J. W.; Lee, Y. H.; Lee, D. H.; Kim, K. M

    2008-06-15

    Corrosion damage significantly influences the integrity and efficiency of steam generator. Corrosion problems of steam generator are unsolved issues until now even though much effort is made around world. Especially the stress corrosion cracking of heat exchange materials is the first issue to be solved. The corrosion protection method of steam generator is important and urgent for the guarantee of nuclear plant's integrity. The objectives of this study are 1) to evaluate the corrosion properties of steam generator materials, 2) to optimize the water chemistry of steam generator and 3) to develop the corrosion protection method of primary and secondary sides of steam generator. The results will be reflected to the water chemistry guideline for improving the integrity and efficiency of steam generator in domestic power plants.

  1. Options for Steam Generator Decommissioning

    International Nuclear Information System (INIS)

    Krause, Gregor; Amcoff, Bjoern; Robinson, Joe

    2016-01-01

    Selecting the best option for decommissioning steam generators is a key consideration in preparing for decommissioning PWR nuclear power plants. Steam Generators represent a discrete waste stream of large, complex items that can lend themselves to a variety of options for handling, treatment, recycling and disposal. Studsvik has significant experience in processing full size Steam Generators at its metal recycling facility in Sweden, and this paper will introduce the Studsvik steam generator treatment concept and the results achieved to date across a number of projects. The paper will outline the important parameters needed at an early stage to assess options and to help consider the balance between off-site and on-site treatment solutions, and the role of prior decontamination techniques. The paper also outlines the use of feasibility studies and demonstration projects that have been used to help customers prepare for decommissioning. The paper discusses physical, radiological and operational history data, Pro and Contra factors for on- and off-site treatment, the role of chemical decontamination prior to treatment, planning for off-site shipments as well as Studsvik experience This paper has an original focus upon the coming challenges of steam generator decommissioning and potential external treatment capacity constraints in the medium term. It also focuses on the potential during operations or initial shut-down to develop robust plans for steam generator management. (authors)

  2. Vibration Analysis for Steam Dryer of APR1400 Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sung-heum; Ko, Doyoung [KHNP CRI, Daejeon (Korea, Republic of); Cho, Minki [Doosan Heavy Industry, Changwon (Korea, Republic of)

    2016-10-15

    This paper is related to comprehensive vibration assessment program for APR1400 steam generator internals. According to U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 (Rev.3, March 2007), we conducted vibration analysis for a steam dryer as the second steam separator of steam generator internals. The vibration analysis was performed at the 100 % power operating condition as the normal operation condition. The random hydraulic loads were calculated by the computational fluid dynamics and the structural responses were predicted by power spectral density analysis for the probabilistic method. In order to meet the recently revised U.S. NRC RG 1.20 Rev.3, the CVAP against the potential adverse flow effects in APR1400 SG internals should be performed. This study conducted the vibration response analysis for the SG steam dryer as the second moisture separator at the 100% power condition, and evaluated the structural integrity. The predicted alternating stress intensities were evaluated to have more than 17.78 times fatigue margin compared to the endurance limit.

  3. Benefits of integrating chemical and mechanical cleaning processes for steam generator sludge removal

    International Nuclear Information System (INIS)

    Varrin, R.D.; Ferriter, A.M.; Oliver, T.W.; Le Surf, J.E.

    1992-01-01

    This paper discusses the benefits of performing in-bundle tubesheet lancing in conjunction with chemical cleaning of PWR and PHWR steam generators in which a hard sludge pile is known to exist. The primary benefits of in-bundle lancing are to: (1) increase the exposed area of the sludge pile by cutting furrows in the surface thereby enhancing dissolution of sludge, (2) reduce the volume of solvents required since material removed by lancing does not have to be dissolved chemically, (3) improve rinsing and removal of residual solvent between iron and copper dissolution steps, and (4) allow for verification of process effectiveness by providing high quality in-bundle visual inspection. The reduction in solvent volumes can lead to a significant reduction in solvent costs and waste processing. A case study which includes an economic evaluation for a combined chemical and mechanical cleaning shows a potential cost saving of up to US$ 300,000 over use of chemical cleaning alone. 14 refs., 2 tabs., 2 figs

  4. Model of reverse steam generator

    International Nuclear Information System (INIS)

    Malasek, V.; Manek, O.; Masek, V.; Riman, J.

    1987-01-01

    The claim of Czechoslovak discovery no. 239272 is a model designed for the verification of the properties of a reverse steam generator during the penetration of water, steam-water mixture or steam into liquid metal flowing inside the heat exchange tubes. The design may primarily be used for steam generators with a built-in inter-tube structure. The model is provided with several injection devices configured in different heat exchange tubes, spaced at different distances along the model axis. The design consists in that between the pressure and the circumferential casings there are transverse partitions and that in one chamber consisting of the circumferential casings, pressure casing and two adjoining partitions there is only one passage of the injection device through the inter-tube space. (Z.M.). 1 fig

  5. Darlington steam generator life assurance program

    International Nuclear Information System (INIS)

    Jelinski, E.; Dymarski, M.; Maruska, C.; Cartar, E.

    1995-01-01

    The Darlington Nuclear Generating Station belonging to Ontario Hydro is one of the most modern and advanced nuclear generating stations in the world. Four reactor units each generate 881 net MW, enough to provide power to a major city, and representing approximately 20% of the Ontario grid. The nuclear generating capacity in Ontario represents approximately 60% of the grid. In order to look after this major asset, many proactive preventative and predictive maintenance programs are being put in place. The steam generators are a major component in any power plant. World wide experience shows that nuclear steam generators require specialized attention to ensure reliable operation over the station life. This paper describes the Darlington steam generator life assurance program in terms of degradation identification, monitoring and management. The requirements for chemistry control, surveillance of process parameters, surveillance of inspection parameters, and the integration of preventative and predictive maintenance programs such as water lancing, chemical cleaning, RIHT monitoring, and other diagnostics to enhance our understanding of life management issues are identified and discussed. We conclude that we have advanced proactive activities to avoid and to minimize many of the problems affecting other steam generators. An effective steam generator maintenance program must expand the knowledge horizon to understand life limiting processes and to analyze and synthesize observations with theory. (author)

  6. An interation of lifetime monitoring of steam generators in power control systems; Integration der Lebensdauerueberwachung von Dampferzeugern in die Kraftwerksleittechnik

    Energy Technology Data Exchange (ETDEWEB)

    Kunze, Ulrich; Pels Leusden, Christoph; Spinner, Ralf [Siemens AG, Erlangen (Germany). Energy Sector; Hackstein, Holger [Siemens AG, Offenbach am Main (Germany). Energy Sector; Walz, Horst [Siemens AG, Karlsruhe (Germany). Energy Sector

    2008-07-01

    The substantial cost-relevant requirements of the operation of power stations are a highly flexible operation, efficient maintenance, a high efficiency and a high availability. Computer-assisted procedures are indispensable for the continuous monitoring of lifetime consumption and for the condition-dependent maintenance of the boiler. The fatigue monitoring system (FMS) offers all possibilities of the control system. The authors of the contribution under consideration report on an integration of life time monitoring of steam generators into the power station control technology. The technical fundamentals for the computation of the boiler lifetime as well as the fundamentals of integration philosophy and their conversion are presented. Subsequently, a configuration exemplarily is presented, and its results are described.

  7. Steam generator auxiliary systems

    International Nuclear Information System (INIS)

    Heinz, A.

    1982-01-01

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

  8. Fourth international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Tuomisto, H. [ed.] [IVO Group, Vantaa (Finland); Purhonen, H. [ed.] [VTT, Espoo (Finland); Kouhia, V. [ed.] [Lappeenranta Univ. of Technology (Finland)

    1997-12-31

    The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries.

  9. Fourth international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Tuomisto, H [ed.; IVO Group, Vantaa (Finland); Purhonen, H [ed.; VTT, Espoo (Finland); Kouhia, V [ed.; Lappeenranta Univ. of Technology (Finland)

    1998-12-31

    The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries.

  10. Fourth international seminar on horizontal steam generators

    International Nuclear Information System (INIS)

    Tuomisto, H.; Purhonen, H.; Kouhia, V.

    1997-01-01

    The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries

  11. A study on integrity of LMFBR secondary cooling system to hypothetical tube failure propagation in the steam generator

    International Nuclear Information System (INIS)

    Yoshihisa Shindo; Kazuo Haga

    2005-01-01

    Full text of publication follows: A fundamental safety issue of liquid-metal-cooled fast breeder reactor (LMFBR) is to maintain the integrity of the secondary cooling system components against violent chemical sodium-water reaction caused by the water leak from the heat transfer tube of steam generators (SG). The produced sodium-water reaction jet would attack more severely surrounding tubes and would cause other tube failures (tube failure propagation), if it was assumed that the water leak was not detected by function-less detectors and proper operating actions to mitigate the tube failure propagation, such as isolations of the SG from the secondary cooling system and turbine water/steam system, and blowing water and steam inside tubes in the SG, were not taken. This study has been made focusing on the affection of large-scale water leak enlarged due to SG tube failure propagation to the structural integrity of the secondary cooling system because the generated pressure pulse caused by a large-scale sodium-water reaction might break heat transfer tubes of the intermediate heat exchanger (IHX). The present work has been made as one part of the study of probabilistic safety assessment (PSA) of LMFBR, because if the heat-transfer tubes of IHX were failed, the reactor core may be affected by the pressure pulse and/or by the sodium-water reaction products transported through the primary cooling system. As tools for PSA of the water leak incident of SG, we have developed QUARK-LP Version 4 code that mainly analyzes the high temperature rupture phenomena and estimates the number of failed tubes during the middle-scale water leak. The pressure pulse behavior generated by sodium-water reaction in the failure SG and the pressure propagation in the secondary cooling system are calculated by using the SWAAM-2 code developed by ANL. Furthermore, the quasi-steady state high pressure and temperature of the secondary cooling system in a long term is estimated by using the SWAAM

  12. IAEA activities on steam generator life management

    International Nuclear Information System (INIS)

    Gueorguiev, B.; Lyssakov, V.; Trampus, P.

    2002-01-01

    The International Atomic Energy Agency (IAEA) carries out a set of activities in the field of Nuclear Power Plant (NPP) life management. Main activities within this area are implemented through the Technical Working Group on Life Management of NPPs, and mostly concentrated on studies of understanding mechanisms of degradation and their monitoring, optimisation of maintenance management, economic aspects, proven practices of and approaches to plant life management including decommissioning. The paper covers two ongoing activities related to steam generator life management: the International Database on NPP Steam Generators and the Co-ordinated Research Project on Verification of WWER Steam Generator Tube Integrity (WWER is the Russian designed PWR). The lifetime assessment of main components relies on an ability to assess their condition and predict future degradation trends, which to a large extent is dependent on the availability of relevant data. Effective management of ageing and degradation processes requires a large amount of data. Several years ago the IAEA started to work on the International Database on NPP Life Management. This is a multi-module database consisting of modules such as reactor pressure vessels materials, piping, steam generators, and concrete structures. At present the work on pressure vessel materials, on piping as well as on steam generator is completed. The paper will present the concept and structure of the steam generator module of the database. In countries operating WWER NPPs, there are big differences in the eddy current inspection strategy and practice as well as in the approach to steam generator heat exchanger tube structural integrity assessment. Responding to the need for a co-ordinated research to compare eddy current testing results with destructive testing using pulled out tubes from WWER steam generators, the IAEA launched this project. The main objectives of the project are to summarise the operating experiences of WWER

  13. Water box for steam generator

    International Nuclear Information System (INIS)

    Lecomte, Robert; Viaud, Michel.

    1975-01-01

    This invention relates to a water box for connecting an assembly composed of a vertical steam generator and a vertical pump to the vessel of the nuclear reactor, the assembly forming the primary cooling system of a pressurised water reactor. This invention makes it easy to dismantle the pump on the water box without significant loss of water in the primary cooling system of the reactor and particularly without it being necessary to drain the water contained in the steam generator beforehand. It makes it possible to shorten the time required for dismantling the primary pump in order to service or repair it and makes dismantling safer in that the dismantling does not involve draining the steam generator and therefore the critical storage of a large amount of cooling water that has been in contact with the fuel assemblies of the nuclear reactor core [fr

  14. On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers, Volumes 1, 2

    Energy Technology Data Exchange (ETDEWEB)

    Upadhyaya, Belle R. [Univ. of Tennessee, Knoxville, TN (United States); Hines, J. Wesley [Univ. of Tennessee, Knoxville, TN (United States); Lu, Baofu [Univ. of Tennessee, Knoxville, TN (United States)

    2005-06-03

    The overall purpose of this Nuclear Engineering Education Research (NEER) project was to integrate new, innovative, and existing technologies to develop a fault diagnostics and characterization system for nuclear plant steam generators (SG) and heat exchangers (HX). Issues related to system level degradation of SG and HX tubing, including tube fouling, performance under reduced heat transfer area, and the damage caused by stress corrosion cracking, are the important factors that influence overall plant operation, maintenance, and economic viability of nuclear power systems. The research at The University of Tennessee focused on the development of techniques for monitoring process and structural integrity of steam generators and heat exchangers. The objectives of the project were accomplished by the completion of the following tasks. All the objectives were accomplished during the project period. This report summarizes the research and development activities, results, and accomplishments during June 2001 September 2004. Development and testing of a high-fidelity nodal model of a U-tube steam generator (UTSG) to simulate the effects of fouling and to generate a database representing normal and degraded process conditions. Application of the group method of data handling (GMDH) method for process variable prediction. Development of a laboratory test module to simulate particulate fouling of HX tubes and its effect on overall thermal resistance. Application of the GMDH technique to predict HX fluid temperatures, and to compare with the calculated thermal resistance.Development of a hybrid modeling technique for process diagnosis and its evaluation using laboratory heat exchanger test data. Development and testing of a sensor suite using piezo-electric devices for monitoring structural integrity of both flat plates (beams) and tubing. Experiments were performed in air, and in water with and without bubbly flow. Development of advanced signal processing methods using

  15. Double wall steam generator tubing

    International Nuclear Information System (INIS)

    Padden, T.R.; Uber, C.F.

    1983-01-01

    Double-walled steam generator tubing for the steam generators of a liquid metal cooled fast breeder reactor prevents sliding between the surfaces due to a mechanical interlock. Forces resulting from differential thermal expansion between the outer tube and the inner tube are insufficient in magnitude to cause shearing of base metal. The interlock is formed by jointly drawing the tubing, with the inside wall of the outer tube being already formed with grooves. The drawing causes the outer wall of the inner tube to form corrugations locking with the grooves. (author)

  16. Experimental evaluation of emergency operating procedures on multiple steam generator tube rupture in INER integral system test facility

    International Nuclear Information System (INIS)

    Liu, T.J.; Lin, Y.M.; Lee, C.H.; Chang, C.Y.; Hong, W.T.

    1997-01-01

    The multiple steam generator tube rupture (SGTR) scenario in Westinghouse type pressurized water reactor (PWR) has been investigated at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility. This reduced-height and reduced-pressure test facility was designed to simulate the main features of Maanshan nuclear power plant. The SGTR test scenario assumes the double-ended break of one-, two- and six- tubes without other failures. The major operator actions follow the related symptom-oriented Emergency Operating Procedure (EOP) on the reference plant. This study focuses on the investigation of thermal-hydraulics phenomena and the adequacy of associated EOP to limit primary-to-secondary leakage. Through this study, it is found that the adequacy of current EOP in minimizing the radioactivity release demands early substantial operator involvement, especially in the multi-tubes break events. Also, the detailed mechanism of the main thermal-hydraulic phenomena during the SGTR transient are explored. (author)

  17. Predicting steam generator crevice chemistry

    International Nuclear Information System (INIS)

    Burton, G.; Strati, G.

    2006-01-01

    'Full text:' Corrosion of steam cycle components produces insoluble material, mostly iron oxides, that are transported to the steam generator (SG) via the feedwater and deposited on internal surfaces such as the tubes, tube support plates and the tubesheet. The build up of these corrosion products over time can lead to regions of restricted flow with water chemistry that may be significantly different, and potentially more corrosive to SG tube material, than the bulk steam generator water chemistry. The aim of the present work is to predict SG crevice chemistry using experimentation and modelling as part of AECL's overall strategy for steam generator life management. Hideout-return experiments are performed under CANDU steam generator conditions to assess the accumulation of impurities in hideout, and return from, model crevices. The results are used to validate the ChemSolv model that predicts steam generator crevice impurity concentrations, and high temperature pH, based on process parameters (e.g., heat flux, primary side temperature) and blowdown water chemistry. The model has been incorporated into ChemAND, AECL's system health monitoring software for chemistry monitoring, analysis and diagnostics that has been installed at two domestic and one international CANDU station. ChemAND provides the station chemists with the only method to predict SG crevice chemistry. In one recent application, the software has been used to evaluate the crevice chemistry based on the elevated, but balanced, SG bulk water impurity concentrations present during reactor startup, in order to reduce hold times. The present paper will describe recent hideout-return experiments that are used for the validation of the ChemSolv model, station experience using the software, and improvements to predict the crevice electrochemical potential that will permit station staff to ensure that the SG tubes are in the 'safe operating zone' predicted by Lu (AECL). (author)

  18. Optimised heat recovery steam generators for integrated solar combined cycle plants

    Science.gov (United States)

    Peterseim, Jürgen H.; Huschka, Karsten

    2017-06-01

    The cost of concentrating solar power (CSP) plants is decreasing but, due to the cost differences and the currently limited value of energy storage, implementation of new facilities is still slow compared to photovoltaic systems. One recognized option to lower cost instantly is the hybridization of CSP with other energy sources, such as natural gas or biomass. Various references exist for the combination of CSP with natural gas in combined cycle plants, also known as Integrated Solar Combined Cycle (ISCC) plants. One problem with current ISCC concepts is the so called ISCC crisis, which occurs when CSP is not contributing and cycle efficiency falls below efficiency levels of solely natural gas only fired combined cycle plants. This paper analyses current ISCC concepts and compares them with two optimised designs. The comparison is based on a Kuraymat type ISCC plant and shows that cycle optimization enables a net capacity increase of 1.4% and additional daily generation of up to 7.9%. The specific investment of the optimised Integrated Solar Combined Cycle plant results in a 0.4% cost increase, which is below the additional net capacity and daily generation increase.

  19. Tube structural integrity evaluation of Palo Verde Unit 1 steam generators for axial upper-bundle cracking

    International Nuclear Information System (INIS)

    Woodman, B.W.; Begley, J.A.; Brown, S.D.; Sweeney, K.; Radspinner, M.; Melton, M.

    1995-01-01

    The analysis of the issue of upper bundle axial ODSCC as it apples to steam generator tube structural integrity in Unit 1 at the Palo Verde Nuclear generating Station is presented in this study. Based on past inspection results for Units 2 and 3 at Palo Verde, the detection of secondary side stress corrosion cracks in the upper bundle region of Unit 1 may occur at some future date. The following discussion provides a description and analysis of the probability of axial ODSCC in Unit 1 leading to the exceedance of Regulatory Guide 1.121 structural limits. The probabilities of structural limit exceedance are estimated as function of run time using a conservative approach. The chosen approach models the historical development of cracks, crack growth, detection of cracks and subsequent removal from service and the initiation and growth of new cracks during a given cycle of operation. Past performance of all Palo Verde Units as well as the historical performance of other steam generators was considered in the development of cracking statistics for application to Unit 1. Data in the literature and Unit 2 pulled tube examination results were used to construct probability of detection curves for the detection of axial IGSCC/IGA using an MRPC (multi-frequency rotating panake coil) eddy current probe. Crack growth rates were estimated from Unit 2 eddy current inspection data combined with pulled tube examination results and data in the literature. A Monte-Carlo probabilistic model is developed to provide an overall assessment of the risk of Regulatory Guide exceedance during plant operation

  20. Steam generator replacement at Surry Power Station

    International Nuclear Information System (INIS)

    McKay, H.S.

    1982-01-01

    The purposes of the steam generator repair program at Surry Power Station were to repair the tube degradation caused by corrosion-related phenomena and to restore the integrity of the steam generators to a level equivalent to new equipment. The repair program consisted of (1) replacing the existing lower-shell assemblies with new ones and (2) adding new moisture separation equipment to the upper-shell assemblies. These tasks required that several pieces of reactor coolant piping, feedwater piping, main steam piping, and the steam generator be cut and refurbished for reinstallation after the new lower shell was in place. The safety implications and other potential effects of the repair program both during the repair work and after the unit was returned to power were part of the design basis of the repair program. The repair program has been completed on Unit 2 without any adverse effects on the health and safety of the general public or to the personnel engaged in the repair work. Before the Unit 1 repair program began, a review of work procedures and field changes for the Unit 2 repair was conducted. Several major changes were made to avoid recurrence of problems and to streamline procedures. Steam generator replacements was completed on June 1, 1981, and the unit is presently in the startup phase of the outrage

  1. Investigation on two-phase flow instability in steam generator of integrated nuclear reactor

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    In the pressure range of 3-18MPa,high pressure steam-water two-phase flow density wave instability in vertical upward parallel pipes with inner diameter of 12mm is studied experimentally.The oscillation curves of two-phase flow instability and the effects of several parameters on the oscillation threshold of the system are obtained.Based on the small pertubation linearization method and the stability principles of automatic control system,a mathematical model is developed to predict the characteristics of density wave instability threshold.The predictions of the model are in good agreement with the experimental results.

  2. A performance analysis of integrated solid oxide fuel cell and heat recovery steam generator for IGFC system

    DEFF Research Database (Denmark)

    Rudra, Souman; Lee, Jinwook; Rosendahl, Lasse

    2010-01-01

    efficiencies can be achieved. The outputs from SOFC can be utilized by heat recovery steam generator (HRSG), which drives the steam turbine for electricity production. The SOFC stack model was developed using the process flow sheet simulator Aspen Plus, which is of the equilibrium type. Various ranges...... of syngas properties gathered from different literature were used for the simulation. The results indicate a trade-off efficiency and power with respect to a variety of SOFC inputs. The HRSG located after SOFC was included in the current simulation study with various operating parameters. This paper...... describes IGFC power plants, particularly the optimization of HRSG to improve the efficiency of the heat recovery from the SOFC exhaust gas and to maximize the power production in the steam cycle in the IGFC system. HRSG output from different pressure levels varies depending on the SOFC output. The steam...

  3. Steam generator life cycle management: Ontario Power Generation (OPG) experience

    International Nuclear Information System (INIS)

    Maruska, C.C.

    2002-01-01

    A systematic managed process for steam generators has been implemented at Ontario Power Generation (OPG) nuclear stations for the past several years. One of the key requirements of this managed process is to have in place long range Steam Generator Life Cycle Management (SG LCM) plans for each unit. The primary goal of these plans is to maximize the value of the nuclear facility through safe and reliable steam generator operation over the expected life of the units. The SG LCM plans integrate and schedule all steam generator actions such as inspection, operation, maintenance, modifications, repairs, assessments, R and D, performance monitoring and feedback. This paper discusses OPG steam generator life cycle management experience to date, including successes, failures and how lessons learned have been re-applied. The discussion includes relevant examples from each of the operating stations: Pickering B and Darlington. It also includes some of the experience and lessons learned from the activities carried out to refurbish the steam generators at Pickering A after several years in long term lay-up. The paper is structured along the various degradation modes that have been observed to date at these sites, including monitoring and mitigating actions taken and future plans. (author)

  4. WWER-1000 steam generator integrity. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1997-07-01

    Programme was initiated by IAEA in 1990 with the aim to assist the countries of Central and Eastern Europe and former Soviet Union in evaluating the safety of their first generation WWER-440/230 nuclear power plants. The main objectives were: to identify major design and operational safety issues; to establish international consensus on priorities for safety improvements; and to provide assistance in the review of the competence and and adequacy of safety improvement programs. The scope was extended in 1992 ro include RBMK, WWER-440/312 and WWER-1000 plants in operation and under construction. Based on the operational experience of more than 90 reactor years of WWER-1000 NPPs having 80 steam generators in operation or under construction the steam generator integrity was recognized as an important issue of high safety concern. The purpose of this report is to integrate available information on the issue of WWER-1000 steam generator integrity with the focus on the steam generator cold collector damage in particular. This information covers the status of stem generators at operating plants, cause analysis of collector cracking, the damage mechanisms involved, operational aspects and corrective measures developed and implemented. Consideration is given to material, design and fabrication related aspects, operational conditions, system solutions, and in-service inspection. Detailed conclusions and recommendations are provided for each of these aspects

  5. Evaluation of nondestructive evaluation size measurement for integrity assessment of axial outside diameter stress corrosion cracking in steam generator tubes

    International Nuclear Information System (INIS)

    Joo, Kyung Mun; Hong, Jun Hee

    2015-01-01

    Recently, the initiation of outside diameter stress corrosion cracking (ODSCC) at the tube support plate region of domestic steam generators (SG) with Alloy 600 HTMA tubes has been increasing. As a result, SGs with Alloy 600 HTMA tubes must be replaced early or are scheduled to be replaced prior to their designed lifetime. ODSCC is one of the biggest threats to the integrity of SG tubes. Therefore, the accurate evaluation of tube integrity to determine ODSCC is needed. Eddy current testing (ECT) is conducted periodically, and its results could be input as parameters for evaluating the integrity of SG tubes. The reliability of an ECT inspection system depends on the performance of the inspection technique and ability of the analyst. The detection probability and ECT sizing error of degradation are considered to be the performance indices of a nondestructive evaluation (NDE) system. This paper introduces an optimized evaluation method for ECT, as well as the sizing error, including the analyst performance. This study was based on the results of a round robin program in which 10 inspection analysts from 5 different companies participated. The analysis of ECT sizing results was performed using a linear regression model relating the true defect size data to the measured ECT size data.

  6. Evaluation of nondestructive evaluation size measurement for integrity assessment of axial outside diameter stress corrosion cracking in steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Kyung Mun [Korea Hydro and Nuclear Power Company Ltd., Central Research Institute, Daejeon (Korea, Republic of); Hong, Jun Hee [Dept. of mechanical Engineering, Chungnam National University, Daejeon (Korea, Republic of)

    2015-02-15

    Recently, the initiation of outside diameter stress corrosion cracking (ODSCC) at the tube support plate region of domestic steam generators (SG) with Alloy 600 HTMA tubes has been increasing. As a result, SGs with Alloy 600 HTMA tubes must be replaced early or are scheduled to be replaced prior to their designed lifetime. ODSCC is one of the biggest threats to the integrity of SG tubes. Therefore, the accurate evaluation of tube integrity to determine ODSCC is needed. Eddy current testing (ECT) is conducted periodically, and its results could be input as parameters for evaluating the integrity of SG tubes. The reliability of an ECT inspection system depends on the performance of the inspection technique and ability of the analyst. The detection probability and ECT sizing error of degradation are considered to be the performance indices of a nondestructive evaluation (NDE) system. This paper introduces an optimized evaluation method for ECT, as well as the sizing error, including the analyst performance. This study was based on the results of a round robin program in which 10 inspection analysts from 5 different companies participated. The analysis of ECT sizing results was performed using a linear regression model relating the true defect size data to the measured ECT size data.

  7. On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers

    International Nuclear Information System (INIS)

    Upadhyaya, Belle R.; Wesley Hines, J.

    2004-01-01

    The overall purpose of this Nuclear Engineering Education Research (NEER) project was to integrate new, innovative, and existing technologies to develop a fault diagnostics and characterization system for nuclear plant steam generators (SG) and heat exchangers (HX). Issues related to system level degradation of SG and HX tubing, including tube fouling, performance under reduced heat transfer area, and the damage caused by stress corrosion cracking, are the important factors that influence overall plant operation, maintenance, and economic viability of nuclear power systems. The research at The University of Tennessee focused on the development of techniques for monitoring process and structural integrity of steam generators and heat exchangers. The objectives of the project were accomplished by the completion of the following tasks. All the objectives were accomplished during the project period. This report summarizes the research and development activities, results, and accomplishments during June 2001-September 2004. (1) Development and testing of a high-fidelity nodal model of a U-tube steam generator (UTSG) to simulate the effects of fouling and to generate a database representing normal and degraded process conditions. Application of the group method of data handling (GMDH) method for process variable prediction. (2) Development of a laboratory test module to simulate particulate fouling of HX tubes and its effect on overall thermal resistance. Application of the GMDH technique to predict HX fluid temperatures, and to compare with the calculated thermal resistance. (3) Development of a hybrid modeling technique for process diagnosis and its evaluation using laboratory heat exchanger test data. (4) Development and testing of a sensor suite using piezo-electric devices for monitoring structural integrity of both flat plates (beams) and tubing. Experiments were performed in air, and in water with and without bubbly flow. (5) Development of advanced signal

  8. On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers.

    Energy Technology Data Exchange (ETDEWEB)

    Belle R. Upadhyaya; J. Wesley Hines

    2004-09-27

    The overall purpose of this Nuclear Engineering Education Research (NEER) project was to integrate new, innovative, and existing technologies to develop a fault diagnostics and characterization system for nuclear plant steam generators (SG) and heat exchangers (HX). Issues related to system level degradation of SG and HX tubing, including tube fouling, performance under reduced heat transfer area, and the damage caused by stress corrosion cracking, are the important factors that influence overall plant operation, maintenance, and economic viability of nuclear power systems. The research at The University of Tennessee focused on the development of techniques for monitoring process and structural integrity of steam generators and heat exchangers. The objectives of the project were accomplished by the completion of the following tasks. All the objectives were accomplished during the project period. This report summarizes the research and development activities, results, and accomplishments during June 2001-September 2004. (1) Development and testing of a high-fidelity nodal model of a U-tube steam generator (UTSG) to simulate the effects of fouling and to generate a database representing normal and degraded process conditions. Application of the group method of data handling (GMDH) method for process variable prediction. (2) Development of a laboratory test module to simulate particulate fouling of HX tubes and its effect on overall thermal resistance. Application of the GMDH technique to predict HX fluid temperatures, and to compare with the calculated thermal resistance. (3) Development of a hybrid modeling technique for process diagnosis and its evaluation using laboratory heat exchanger test data. (4) Development and testing of a sensor suite using piezo-electric devices for monitoring structural integrity of both flat plates (beams) and tubing. Experiments were performed in air, and in water with and without bubbly flow. (5) Development of advanced signal

  9. Surry steam generator - examination and evaluation

    International Nuclear Information System (INIS)

    Clark, R.A.; Doctor, P.G.; Ferris, R.H.

    1985-10-01

    This report summarizes research conducted during the fourth year of the five year Steam Generator Group Project. During this period the project conducted numerous nondestructive examination (NDE) round robin inspections of the original Surry 2A steam generator. They included data acquisition/analysis and analysis-only round robins using multifrequency bobbin coil eddy current tests. In addition, the generator was nondestructively examined by alternate or advanced techniques including ultrasonics, optical fiber, profilometry and special eddy current instrumentation. The round robin interpretation data were compared. To validate the NDE results and for tube integrity testing, a selection of tubing samples, determined to be representative of the generator, was designated for removal. Initial sample removals from the generator included three sections of tube sheet, two sections of support plate and encompassed tubes, and a number of straight and U-bend tubing sections. Metallographic examination of these sections was initiated. Details of significant results are presented in the following paper. 13 figs

  10. Surry steam generator - examination and evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Clark, R A; Doctor, P G; Ferris, R H

    1987-01-01

    This report summarizes research conducted during the fourth year of the five year Steam Generator Group Project. During this period the project conducted numerous nondestructive examination (NDE) round robin inspections of the original Surry 2A steam generator. They included data acquisition/analysis and analysis-only round robins using multifrequency bobbin coil eddy current tests. In addition, the generator was nondestructively examined by alternate or advanced techniques including ultrasonics, optical fiber, profilometry and special eddy current instrumentation. The round robin interpretation data were compared. To validate the NDE results and for tube integrity testing, a selection of tubing samples, determined to be representative of the generator, was designated for removal. Initial sample removals from the generator included three sections of tube sheet, two sections of support plate and encompassed tubes, and a number of straight and U-bend tubing sections. Metallographic examination of these sections was initiated. Details of significant results are presented in the following paper.

  11. Hybrid preheat/recirculating steam generator

    International Nuclear Information System (INIS)

    Lilly, G.P.

    1985-01-01

    The patent describes a hybrid preheat/recirculating steam generator for nuclear power plants. The steam generator utilizes recirculated liquid to preheat incoming liquid. In addition, the steam generator incorporates a divider so as to limit the amount of recirculating water mixed with the feedwater. (U.K.)

  12. Changing the simualtor's steam generator

    International Nuclear Information System (INIS)

    Ruiz Martin, J.A.; Ortega Pascual, F.

    2006-01-01

    Two Spanish nuclear power plants (two PWR units each one) have planned to change their Westinghouse D-3 steam generators (SGo henceforth) for a new model, 61W/D3 from Siemens/KWU (SGn henceforth), during 1995/1997. This is the reason why TECNATOM has developed during 1994's last term, a new software for the full scope simulator that incorporates the modifications related to the steam generator substiution programme. This allows an anticipated training on the procedures, not only for normal, but for emergency procedures. As it is a component which has not yet been included in these plants, there are not real references or operative experience data. Therefore, the design of the validation strategy was one of the key points in this work. (author)

  13. Steam generators: critical components in nuclear steam supply systems

    Energy Technology Data Exchange (ETDEWEB)

    Stevens-Guille, P D

    1974-02-28

    Steam generators are critical components in power reactors. Even small internal leaks result in costly shutdowns for repair. Surveys show that leaks have affected one half of all water-cooled reactors in the world with steam generators. CANDU reactors have demonstrated the highest reliability. However, AECL is actively evolving new technology in design, manufacture, inspection and operation to maintain reliability. (auth)

  14. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  15. Parametric study for horizontal steam generator modelling

    Energy Technology Data Exchange (ETDEWEB)

    Ovtcharova, I. [Energoproekt, Sofia (Bulgaria)

    1995-12-31

    In the presentation some of the calculated results of horizontal steam generator PGV - 440 modelling with RELAP5/Mod3 are described. Two nodalization schemes have been used with different components in the steam dome. A study of parameters variation on the steam generator work and calculated results is made in cases with separator and branch.

  16. Parametric study for horizontal steam generator modelling

    Energy Technology Data Exchange (ETDEWEB)

    Ovtcharova, I [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    In the presentation some of the calculated results of horizontal steam generator PGV - 440 modelling with RELAP5/Mod3 are described. Two nodalization schemes have been used with different components in the steam dome. A study of parameters variation on the steam generator work and calculated results is made in cases with separator and branch.

  17. Horizontal steam generator thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Ubra, O. [SKODA Praha Company, Prague (Czechoslovakia); Doubek, M. [Czech Technical Univ., Prague (Czechoslovakia)

    1995-09-01

    Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. The 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.

  18. Strategic management of steam generators

    International Nuclear Information System (INIS)

    Hernalsteen, P.; Berthe, J.

    1991-01-01

    This paper addresses the general approach followed in Belgium for managing any kind of generic defect affecting a Steam Generator tubebundle. This involves the successive steps of: problem detection, dedicated sample monitoring, implementation of preventive methods, development of specific plugging criteria, dedicated 100% inspection, implementation of repair methods, adjusted sample monitoring and repair versus replacement strategy. These steps are illustrated by the particular case of Primary Water Stress Corrosion Cracking in tube roll transitions, which is presently the main problem for two Belgian units Doele-3 and Tihange-2. (author)

  19. Steam generator for nuclear reactors

    International Nuclear Information System (INIS)

    Byerley, W.M.; Bennett, R.R.

    1978-01-01

    In the steam generator, the primary medium is led through a U-shaped tube bundle heating up a secondary medium (feedwater) which flows around the tube bundle via a preheating chamber. In order to optimize heat transfer inside the preheating chamber, the feedwater is separated into a counterflow and a parallel flow with regard to the primary medium by means of partitioning walls and deflectors. The ratio is 70/30%. This way, boiling in the preheater is avoided, i.e. the high LMTD (logaritmic mean temperature difference) is fully utilized. (DG) [de

  20. Fracture toughness determination in steam generator tubes

    International Nuclear Information System (INIS)

    Bergant M; Yawny, A; Perez Ipina, J

    2012-01-01

    The assessment of the structural integrity of steam generator tubes in nuclear power plants deserved increasing attention in the last years due to the negative impact related to their failures. In this context, elastic plastic fracture mechanics (EPFM) methodology appears as a potential tool for the analysis. The application of EPFM requires, necessarily, knowledge of two aspects, i.e., the driving force estimation in terms of an elastic plastic toughness parameter (e.g., J) and the experimental measurement of the fracture toughness of the material (e.g., the material J-resistance curve). The present work describes the development of a non standardized experimental technique aimed to determine J-resistance curves for steam generator tubes with circumferential through wall cracks. The tubes were made of Incoloy 800 (Ni: 30.0-35.0; Cr: 19.0-23.0; Fe: 35.5 min, % in weight). Due to its austenitic microstructure, this alloy shows very high toughness and is widely used in applications where a good corrosion resistance in aqueous environment or an excellent oxidation resistance in high temperature environment is required. Finally, a procedure for the structural integrity analysis of steam generator tubes with crack-like defects, based on a FAD diagram (Failure Assessment Diagram), is briefly described (author)

  1. Integrated Gasification SOFC Plant with a Steam Plant

    DEFF Research Database (Denmark)

    Rokni, Masoud; Pierobon, Leonardo

    2011-01-01

    A hybrid Solid Oxide Fuel Cell (SOFC) and Steam Turbine (ST) plant is integrated with a gasification plant. Wood chips are fed to the gasification plant to produce biogas and then this gas is fed into the anode side of a SOFC cycle to produce electricity and heat. The gases from the SOFC stacks...... enter into a burner to burn the rest of the fuel. The offgases after the burner are now used to generate steam in a Heat Recovery Steam Generator (HRSG). The generated steam is expanded in a ST to produce additional power. Thus a triple hybrid plant based on a gasification plant, a SOFC plant...... and a steam plant is presented and studied. The plant is called as IGSS (Integrated Gasification SOFC Steam plant). Different systems layouts are presented and investigated. Electrical efficiencies up to 56% are achieved which is considerably higher than the conventional integrated gasification combined...

  2. Steam generator tubing NDE performance

    International Nuclear Information System (INIS)

    Henry, G.; Welty, C.S. Jr.

    1997-01-01

    Steam generator (SG) non-destructive examination (NDE) is a fundamental element in the broader SG in-service inspection (ISI) process, a cornerstone in the management of PWR steam generators. Based on objective performance measures (tube leak forced outages and SG-related capacity factor loss), ISI performance has shown a continually improving trend over the years. Performance of the NDE element is a function of the fundamental capability of the technique, and the ability of the analysis portion of the process in field implementation of the technique. The technology continues to improve in several areas, e.g. system sensitivity, data collection rates, probe/coil design, and data analysis software. With these improvements comes the attendant requirement for qualification of the technique on the damage form(s) to which it will be applied, and for training and qualification of the data analysis element of the ISI process on the field implementation of the technique. The introduction of data transfer via fiber optic line allows for remote data acquisition and analysis, thus improving the efficiency of analysis for a limited pool of data analysts. This paper provides an overview of the current status of SG NDE, and identifies several important issues to be addressed

  3. Thermal-hydraulics in recirculating steam generators

    International Nuclear Information System (INIS)

    Carver, M.B.; Carlucci, L.N.; Inch, W.W.R.

    1981-04-01

    This manual describes the THIRST code and its use in computing three-dimensional two-phase flow and heat transfer in a steam generator under steady state operation. The manual is intended primarily to facilitate the application of the code to the analysis of steam generators typical of CANDU nuclear stations. Application to other steam generator designs is also discussed. Details of the assumptions used to formulate the model and to implement the numerical solution are also included

  4. Chemical cleaning an essential part of steam generator asset management

    International Nuclear Information System (INIS)

    Amman, Franz

    2008-01-01

    Chemical Cleaning an essential part of Steam Generator asset management accumulation of deposits is intrinsic for the operation of Steam Generators in PWRs. Such depositions often lead to reduction of thermal performance, loss of component integrity and, in some cases to power restrictions. Accordingly removal of such deposits is an essential part of the asset management of the Steam Generators in a Nuclear Power Plant. Every plant has its individual condition, history and constraints which need to be considered when planning and performing a chemical cleaning. Typical points are: - Sludge load amount and constitution of the deposits - Sludge distribution in the steam generator - Existing or expected corrosion problems - Amount and tendency of fouling for waste treatment Depending on this points the strategy for chemical cleaning shall be evolved. the range of treatment starts with very soft cleanings with a removal of approx 100 kg per steam generator and goes to a full scale cleaning which can remove up to several thousand kilograms of deposits from a steam generator. Depending on the goal to be achieved and the steam generator present an adequate cleaning method shall be selected. This requires flexible and 'customisable' cleaning methods that can be adapted to the individual needs of a plant. Such customizing of chemical cleaning methods is an essential factor for an optimized asset management of the steam generator in a nuclear power plant

  5. Embalse steam generators - status in 2009

    International Nuclear Information System (INIS)

    Luna, P.; Yetisir, M.; Roy, S.; MacEacheron, R.

    2009-01-01

    The Embalse Nuclear Generating Station (ENGS) is a CANDU 6, a pressurized heavy water plant, with a net capacity of 648 MW. The primary heat transport system at Embalse includes four Steam Generators (SGs) manufactured by Babcock and Wilcox Canada (B and W). These steam generators are vertical recirculating heat exchangers with Incoloy 800 inverted U-tubes and an integral preheater. Embalse SGs performed very well until the late 1990s, when an increase in tube fretting was noticed in the U-bend region. In-service inspection in 2002 and 2004 confirmed that the cause of the tube fretting was flow accelerated corrosion (FAC) damage of scallop bar supports in the U-bend region. The straight leg tube support plates (TSPs) have also been degrading. Degradation was worst at the top support plates, and it was in the form of material loss on the cold leg. The hot leg TSPs were heavily fouled with deposits and flow areas were blocked. Visual inspections and subsequent studies showed that the cause of the TSP degradation was also FAC. The Embalse SGs have carbon steel supports that make them susceptible to FAC. To mitigate the effects of degraded tube support structures, three additional sets of anti-vibration bars were installed in the U-bend regions of all four steam generators in 2004. In 2007, an improved secondary-side chemistry specification was implemented to reduce the FAC rate and the hot leg TSPs was waterlanced. A root cause analysis and condition assessment was performed for the tube supports in 2007. Fitness for Service (FFS) evaluation was completed using the Canadian Industry Guidelines for steam generator tubes. The steam generators were returned to service and the plan has operated without another forced outage to date. The FAC degradation of the carbon steel U-bend tube support systems has had the most significant impact on the plant operation causing a number of forced outages. The discovery of the extent of TSP degradation and difficulties to repair TSPs

  6. Design of PFBR steam generator

    International Nuclear Information System (INIS)

    Chetal, S.C.; Bhoje, S.B.; Mitra, T.K.; Paranjpe, S.R.; Vaidyanathan, G.

    1990-01-01

    Vertical straight tube with an expansion bend in sodium path is the design selected for the steam generators of 500 MWe Prototype Fast Breeder Reactor (PFBR). There are 4 secondary loops with each loop consisting of 3 modules. With sodium reheat incorporated each module comprises of one evaporator, superheater and reheater. Material of construction is 2.25Cr-1Mo for evaporator and 9Cr-1Mo for superheater and reheater. The tube to tubesheet weld is internal bore butt weld with tubesheet having raised spigot. Aim is to have reliable design with higher plant availability. Design considerations leading to the choice of design features selected are discussed in the paper and a ''reference'' design has been described. (author). 2 figs, 1 tab

  7. Method to detect steam generator tube leakage

    International Nuclear Information System (INIS)

    Watabe, Kiyomi

    1994-01-01

    It is important for plant operation to detect minor leakages from the steam generator tube at an early stage, thus, leakage detection has been performed using a condenser air ejector gas monitor and a steam generator blow down monitor, etc. In this study highly-sensitive main steam line monitors have been developed in order to identify leakages in the steam generator more quickly and accurately. The performance of the monitors was verified and the demonstration test at the actual plant was conducted for their intended application to the plants. (author)

  8. Clinch river breeder reactor plant steam generator water quality

    International Nuclear Information System (INIS)

    Van Hoesen, D.; Lowe, P.A.

    1975-01-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: 1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; 2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and 3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present

  9. Clinch river breeder reactor plant steam generator water quality

    Energy Technology Data Exchange (ETDEWEB)

    Van Hoesen, D; Lowe, P A

    1975-07-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: (1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; (2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and (3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present.

  10. Digital simulation for nuclear once-through steam generators

    International Nuclear Information System (INIS)

    Chen, A.T.

    1976-01-01

    Mathematical models for calculating the dynamic response of the Oconee type once through steam generator (OTSG) and the integral economizer once through steam generator (IEOTSG) was developed and presented in this dissertation. Linear and nonlinear models of both steam generator types were formulated using the state variable, lumped parameter approach. Transient and frequency responses of system parameters were calculated for various perturbations from both the primary coolant side and the secondary side. Transients of key parameters, including primary outlet temperature, superheated steam outlet temperature, boiling length/subcooled length and steam pressure, were generated, compared and discussed for both steam generator types. Frequency responses of delta P/sub s//deltaT/sub pin/ of the linear OTSG model were validated by using the dynamic testing results obtained at the Oconee I nuclear power station. A sensitivity analysis in both the time and the frequency domains was performed. It was concluded that the mathematical and computer models developed in this dissertation for both the OTSG and the IEOTSG are suitable for overall plant performance evaluation and steam generator related component/system design analysis for nuclear plants using either type of steam generator

  11. Steam generator tube rupture (SGTR) scenarios

    International Nuclear Information System (INIS)

    Auvinen, A.; Jokiniemi, J.K.; Laehde, A.; Routamo, T.; Lundstroem, P.; Tuomisto, H.; Dienstbier, J.; Guentay, S.; Suckow, D.; Dehbi, A.; Slootman, M.; Herranz, L.; Peyres, V.; Polo, J.

    2005-01-01

    The steam generator tube rupture (SGTR) scenarios project was carried out in the EU 5th framework programme in the field of nuclear safety during years 2000-2002. The first objective of the project was to generate a comprehensive database on fission product retention in a steam generator. The second objective was to verify and develop predictive models to support accident management interventions in steam generator tube rupture sequences, which either directly lead to severe accident conditions or are induced by other sequences leading to severe accidents. The models developed for fission product retention were to be included in severe accident codes. In addition, it was shown that existing models for turbulent deposition, which is the dominating deposition mechanism in dry conditions and at high flow rates, contain large uncertainties. The results of the project are applicable to various pressurised water reactors, including vertical steam generators (western PWR) and horizontal steam generators (VVER)

  12. Thermal performances of molten salt steam generator

    International Nuclear Information System (INIS)

    Yuan, Yibo; He, Canming; Lu, Jianfeng; Ding, Jing

    2016-01-01

    Highlights: • Thermal performances of molten salt steam generator were experimentally studied. • Overall heat transfer coefficient reached maximum with optimal molten salt flow rate. • Energy efficiency first rose and then decreased with salt flow rate and temperature. • Optimal molten salt flow rate and temperature existed for good thermal performance. • High inlet water temperature benefited steam generating rate and energy efficiency. - Abstract: Molten salt steam generator is the key technology for thermal energy conversion from high temperature molten salt to steam, and it is used in solar thermal power station and molten salt reactor. A shell and tube type molten salt steam generator was set up, and its thermal performance and heat transfer mechanism were studied. As a coupling heat transfer process, molten salt steam generation is mainly affected by molten salt convective heat transfer and boiling heat transfer, while its energy efficiency is also affected by the heat loss. As molten salt temperature increased, the energy efficiency first rose with the increase of heat flow absorbed by water/steam, and then slightly decreased for large heat loss as the absorbed heat flow still rising. At very high molten salt temperature, the absorbed heat flow decreased as boiling heat transfer coefficient dropping, and then the energy efficiency quickly dropped. As the inlet water temperature increased, the boiling region in the steam generator remarkably expanded, and then the steam generation rate and energy efficiency both rose with the overall heat transfer coefficient increasing. As the molten salt flow rate increased, the wall temperature rose and the boiling heat transfer coefficient first increased and then decreased according to the boiling curve, so the overall heat transfer coefficient first increased and then decreased, and then the steam generation rate and energy efficiency of steam generator both had maxima.

  13. Steam generator leak detection using acoustic method

    International Nuclear Information System (INIS)

    Goluchko, V.V.; Sokolov, B.M.; Bulanov, A.N.

    1982-05-01

    The main requirements to meet by a device for leak detection in sodium - water steam generators are determined. The potentialities of instrumentation designed based on the developed requirements have been tested using a model of a 550 kw steam generator [fr

  14. LMR steam generator blowdown with RETRAN

    International Nuclear Information System (INIS)

    Wei, T.Y.C.

    1985-01-01

    One of the transients being considered in the FSAR Chapter 15 analyses of anticipated LMR transients is the fast blowdown of a steam generator upon inadvertent actuation of the liquid metal/water reaction mitigation system. For the blowdown analysis, a stand-alone steam generator model for the IFR plant was constructed using RETRAN

  15. Future steam generator designs. Single wall designs

    Energy Technology Data Exchange (ETDEWEB)

    Hayden, O [Nuclear Power Company Ltd, Warrington, Cheshire (United Kingdom)

    1978-10-01

    to be cut off and only extremely limited and difficult access to the tube plate can be obtained. Thus, for the replacement PFR superheater and reheater now in production, design changes have been introduced which improve the facilities for inspection, improve the engineering integrity of the sodium/water boundary and will enable NPC to prove the new units under real power station operating conditions. The CDFR design is a scaled up replica of the replacement Units and we, therefore, hope to validate the CDFR design by the successful operation in PFR of the new prototype units some time during 1981. It is worthwhile to comment upon the potential advantages of the gas space' requirement in the PFR type of removable 'U' tube. With regard to sodium/water reaction events in the high risk area of the PFR Steam Generators has proved to be in the tube to tube sheet welds and it is fortuitous that the provision of a sodium side gas space enables the designer to incorporate a highly sensitive leak detection system, with a large 'window of protection' value for the monitoring of the critical weld feature. Thus, if the operator wishes, he can lower the level of the sodium during leak location operations and convert what would be a more onerous under-sodium leak into a gas space. leak. If any opening up operations are required, then this could be of some advantage. Design changes from PFR to CDFR as incorporated into the prototype replacement steam generators are shown.

  16. Future steam generator designs. Single wall designs

    International Nuclear Information System (INIS)

    Hayden, O.

    1978-01-01

    to be cut off and only extremely limited and difficult access to the tube plate can be obtained. Thus, for the replacement PFR superheater and reheater now in production, design changes have been introduced which improve the facilities for inspection, improve the engineering integrity of the sodium/water boundary and will enable NPC to prove the new units under real power station operating conditions. The CDFR design is a scaled up replica of the replacement Units and we, therefore, hope to validate the CDFR design by the successful operation in PFR of the new prototype units some time during 1981. It is worthwhile to comment upon the potential advantages of the gas space' requirement in the PFR type of removable 'U' tube. With regard to sodium/water reaction events in the high risk area of the PFR Steam Generators has proved to be in the tube to tube sheet welds and it is fortuitous that the provision of a sodium side gas space enables the designer to incorporate a highly sensitive leak detection system, with a large 'window of protection' value for the monitoring of the critical weld feature. Thus, if the operator wishes, he can lower the level of the sodium during leak location operations and convert what would be a more onerous under-sodium leak into a gas space. leak. If any opening up operations are required, then this could be of some advantage. Design changes from PFR to CDFR as incorporated into the prototype replacement steam generators are shown

  17. CAREM-25 Steam Generator Stability Analysis

    International Nuclear Information System (INIS)

    Rabiti, A.; Delmastro, D.

    2003-01-01

    In this work the stability of a once-through CAREM-25 steam generator is analyzed.A fix nodes numerical model, that allows the modelling of the liquid, two-phase and superheated steam zones, is implemented.This model was checked against a mobile finite elements model under saturated steam conditions at the channel exit and a good agreement was obtained.Finally the stability of a CAREM steam generator is studied and the range of in let restrictions that a assure the system stability is analyzed

  18. Strategic maintenance plan for Cernavoda steam generators

    International Nuclear Information System (INIS)

    Cicerone, T.; Dhar, D.; VandenBerg, J.P.

    2002-01-01

    Steam generators are among the most important pieces of equipment in a nuclear power plant. They are required full time during the plant operation and obviously no redundancy exists. Past experience has shown that those utilities which implemented comprehensive steam generator inspection and maintenance programs and strict water chemistry controls, have had good steam generator performance that supports good overall plant performance. The purpose of this paper is to discuss a strategic Life Management and Operational-monitoring program for the Cernavoda steam generators. The program is first of all to develop a base of expertise for the management of the steam generator condition; and that is to be supported by a program of actions to be accomplished over time to assess their condition, to take measures to avoid degradation and to provide for inspections, cleaning and modifications as necessary. (author)

  19. Maintenance and repair of LMFBR steam generators

    International Nuclear Information System (INIS)

    Verriere, P.; Alanche, J.; Minguet, J.L.

    1984-06-01

    After some general remarks on the French fast neutron system, this paper presents the state of the program for the construction of fast reactor in France. Then, the general design of Super Phenix 1 steam generator components is outlined and, the in-service monitoring systems and protective devices with which they are equiped are briefly described. The methods used, in the event of leakage, for leak location, steam generator inspection, steam generator repair and putting the affected loop back into service, are discussed. There are two main lines of research, relating respectively to the means of water leak detection in sodium and the inspection arrangements that will be used either periodically, or following a sodium-water reaction. Finally, after a brief description of the steam generator, this paper describes the four incidents (leaks) that occurred on the Phenix steam generator in the course of 1982 and 1983, and the subsequent repair operations

  20. Cleanliness criteria to improve steam generator performance

    International Nuclear Information System (INIS)

    Schwarz, T.; Bouecke, R.; Odar, S.

    2005-01-01

    High steam generator performance is a prerequisite for high plant availability and possible life time extension. The major opponent to that is corrosion and fouling of the heating tubes. Such steam generator degradation problems arise from the continuous ingress of non-volatile contaminants, i.e. corrosion products and salt impurities may accumulate in the steam generators. These impurities have their origin in the secondary side systems. The corrosion products generally accumulate in the steam generators and form deposits not only in the flow restricted areas, such as on top of tube sheet and tube support structure, but also build scales on the steam generator heating tubes. In addition, the tube scales in general affect the steam generator thermal performance, which ultimately causes a reduction of power output. The most effective ways of counteracting all these degradation problems, and thus of improving the steam generator performance is to keep them in clean conditions or, if judged necessary, to plan cleaning measures such as mechanical tube sheet lancing or chemical cleaning. This paper presents a methodology how to assess the cleanliness condition of a steam generator by bringing together all available operational and inspection data such as thermal performance and water chemistry data. By means of this all-inclusive approach the cleanliness condition is quantified in terms of a fouling index. The fouling index allows to monitor the condition of a specific steam generator, compare it to other plants and, finally, to serve as criterion for cleaning measures such as chemical cleaning. The application of the cleanliness criteria and the achieved field results with respect to improvements of steam generator performance will be presented. (author)

  1. Steam generator operating experience: Update for 1984-1986

    International Nuclear Information System (INIS)

    Frank, L.; Stokley, J.

    1988-06-01

    This report summarizes operational events and degradation mechanisms affecting pressurized water reactor steam generator integrity, provides updated inspection results reported in 1984, 1985, and 1986, and highlights both prevalent problem areas and advances in improved equipment test practices, preventive measures, repair techniques, and replacement procedures. It describes equipment design features of the three major suppliers and discusses 68 plants in detail. Steam generator degradation mechanisms include intergranular stress corrosion cracking, primary water stress corrosion cracking, pitting, intergranular attack, and vibration wear that effects tube integrity and causes leakage. Plugging, sleeving heat treatment, peening, chemical cleaning, and steam generator replacements are described and regulatory instruments and inspection guidelines for nondestructive evaluations and girth weld cracking are discusses. The report concludes that although degradation mechanisms are generally understood, the elimination of unscheduled plant shutdowns and costly repairs resulting from leaking tubes has not been achieved. Highlights of steam generator research and unresolved safety issues are discussed. 21 refs., 8 tabs

  2. Testing installation for a steam generator

    International Nuclear Information System (INIS)

    Dubourg, M.

    1985-01-01

    The invention proposes a testing installation for a steam generator associated to a boiler, comprising a testing exchanger connected to a feeding circuit in secondary fluid and to a circuit to release the steam produced, and comprising a heating-tube bundle connected to a closed circuit of circulation of a primary coolant at the same temperature and at the pressure than the primary fluid. The heating-tube bundle of the testing exchanger has the same height than the primary bundle of the steam generator and the testing exchanger is at the same level and near the steam generator and is fed by the same secondary fluid such as it is subject to the same operation phases during a long period. The in - vention applies, more particularly, to the steam generators of pressurized water nuclear power plants [fr

  3. Response of the steam generator VVER 1000 to a steam line break

    International Nuclear Information System (INIS)

    Novotny, J.; Novotny, J. Jr.

    2003-01-01

    Dynamic effects of a steam line break in the weld of the steam pipe and the steam collector on the steam generator system are analyzed. Modelling of a steam line break may concern two cases. The steam line without a restraint and the steam line protected by a whip restraint with viscous elements applied at the postulated break cross-section. The second case is considered. Programme SYSTUS offers a special element the stiffness and viscous damping coefficients of which may be defined as dependent on the relative displacement and velocity of its nodes respectively. A circumferential crack is simulated by a sudden decrease of longitudinal and lateral stiffness coefficients of these special SYSTUS elements to zero. The computation has shown that one can simulate the pipe to behave like completely broken during a time interval of 0,0001 s or less. These elements are used to model the whip restraint with viscous elements and viscous dampers of the GERB type as well. In the case of a whip restraint model the stiffness coefficient-displacement relation and damping coefficient - velocity relation are chosen to fit the given characteristics of the restraint. The special SYSTUS elements are used to constitute Maxwell elements modelling the elasto-plastic and viscous properties of the GERB dampers applied to the steam generator. It has been ascertained that a steam line break at the postulated weld crack between the steam pipe and the steam generator collector cannot endanger the integrity of the system even in a case of the absence of a whip restraint effect. (author)

  4. Steam Generator Inspection Planning Expert System

    International Nuclear Information System (INIS)

    Rzasa, P.

    1987-01-01

    Applying Artificial Intelligence technology to steam generator non-destructive examination (NDE) can help identify high risk locations in steam generators and can aid in preparing technical specification compliant eddy current test (ECT) programs. A steam Generator Inspection Planning Expert System has been developed which can assist NDE or utility personnel in planning ECT programs. This system represents and processes its information using an object oriented declarative knowledge base, heuristic rules, and symbolic information processing, three artificial intelligence based techniques incorporated in the design. The output of the system is an automated generation of ECT programs. Used in an outage inspection, this system significantly reduced planning time

  5. Chemical cleaning - essential for optimal steam generator asset management

    International Nuclear Information System (INIS)

    Ammann, Franz

    2009-01-01

    Accumulation of deposits in Steam Generator is intrinsic during the operation of Pressurized Water Reactors. Such depositions lead to reduction of thermal performance, loss of component integrity and, in some cases, to power restrictions. Accordingly, removal of such deposits is an essential part of the asset management program of Steam Generators. Every plant has specific conditions, history and constraints which must be considered when planning and performing a chemical cleaning. Typical points are: -Constitution of the deposits or sludge - Sludge load - Sludge distribution in the steam generator - Existing or expected corrosion problems - Amount and tendency of fouling for waste treatment The strategy for chemical cleaning is developed from these points. The range of chemical cleaning treatments starts with very soft cleanings which can remove approximately 100kg per steam generator and ends with full scale, i.e., hard, cleanings which can remove several thousand kilograms of deposits from a steam generator. Dependent upon the desired goal for the operating plant and the steam generator material condition, the correct cleaning method can be selected. This requires flexible cleaning methods that can be adapted to the individual needs of a plant. Such customizing of chemical cleaning methods is a crucial factor for an optimized asset management program of steam generators in a nuclear power plant

  6. Cheaper power generation from surplus steam generating capacities

    International Nuclear Information System (INIS)

    Gupta, K.

    1996-01-01

    Prior to independence most industries had their own captive power generation. Steam was generated in own medium/low pressure boilers and passed through extraction condensing turbines for power generation. Extraction steam was used for process. With cheaper power made available in Nehru era by undertaking large hydro power schemes, captive power generation in industries was almost abandoned except in sugar and large paper factories, which were high consumers of steam. (author)

  7. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

  8. International examples of steam generator replacement

    International Nuclear Information System (INIS)

    Wiechmann, K.

    1993-01-01

    Since 1979-1980 a total of twelve nuclear power plants world-wide have had their steam generators replaced. The replacement of the Combustion steam generators in the Millstone-2 plant in the United States was completed very recently. Steam generator replacement activities are going on at present in four plants. In North Anna, the steam generators have been under replacement since January 1990. In Japan, preparations have been started for Genkai-1. Since January 1992, the two projects in Beznau-1, Switzerland, and Doel-3, Belgium, have bee planned and executed in parallel. Why steam generator replacement? There are a number of defect mechanisms which give rise to the need for early steam generator replacement. One of the main reasons is the use of Inconel-600 as material for the heating tubes. Steam generator heating tubes made of Inconel-600 have been known to exhibit their first defects due to stress corrosion cracking after less than one year of operation. (orig.) [de

  9. An expert system for steam generator maintenance

    International Nuclear Information System (INIS)

    Remond, A.

    1988-01-01

    The tube bundles in PWR steam generators are, by far, the major source of problems whether they are due to primary and secondary side corrosion mechanisms or to tube vibration-induced wear at tube support locations. Because of differences in SG operating, materials, and fabrication processes, the damage may differ from steam generator to steam generator. MPGV, an expert system for steam generator maintenance uses all steam generator data containing data on materials, fabrication processes, inservice inspection, and water chemistry. It has access to operational data for individual steam generators and contains models of possible degradation mechanisms. The objectives of the system are: · Diagnosing the most probable degradation mechanism or mechanisms by reviewing the entire steam generator history. · Identifying the tubes most exposed to future damage and evaluating the urgency of repair by simulating the probable development of the problem in time. · Establishing the appropriate preventive actions such as repair, inspection or other measures and establishing an action schedule. The system is intended for utilities either for individual plants before each inspection outage or any time an incident occurs or for a set of plants through a central MPGV center. (author)

  10. Preliminary design study of removable integral steam generator units of the multiple helically wound tube type for a 1250 MW(th) H.T.G.C. reactor

    International Nuclear Information System (INIS)

    Gilli, P.V.; Fritz, K.; Lippitsch, J.; Sandri, A.H.; Weiss, B.

    1965-11-01

    The possibilities of designing a multiple steam generator for a 1250 MW(th) High Temperature Gas-Cooled Reactor, consisting of 18 units which are able to pass through 5 ft diam. holes in the integral prestressed concrete pressure vessel are investigated. A lay-out and design with bundles of multi-start helical tubes is evolved, particular attention being paid to the questions of tube blanking and removal of the unit, and of selection of materials for superheater and reheater tubes. Thermal and stress calculations have been carried out, using the Waagner-Biro Computer Code ADURHELIX. (author)

  11. Procedure for generating steam and steam generator for operating said procedure

    International Nuclear Information System (INIS)

    Chlique, Bernard.

    1975-01-01

    This invention concerns the generation of steam by bringing the water to be vaporised into indirect thermal exchange relation with the heating steam which condenses when passing in series, along alternate routes, through bundles of tubes immersed in a vaporising chamber. A number of steam generators working on this principle already exist. The purpose of the invention is to modify the operating method of these steam generators by means of a special disposition making it possible to build a compact unit including an additional bundle of tubes heated by the condensates collected at the outlet of each bundle through which the heating steam passes [fr

  12. Control system for fluid heated steam generator

    Science.gov (United States)

    Boland, J.F.; Koenig, J.F.

    1984-05-29

    A control system for controlling the location of the nucleate-boiling region in a fluid heated steam generator comprises means for measuring the temperature gradient (change in temperature per unit length) of the heating fluid along the steam generator; means for determining a control variable in accordance with a predetermined function of temperature gradients and for generating a control signal in response thereto; and means for adjusting the feedwater flow rate in accordance with the control signal.

  13. Dynamic modelling of nuclear steam generators

    International Nuclear Information System (INIS)

    Kerlin, T.W.; Katz, E.M.; Freels, J.; Thakkar, J.

    1980-01-01

    Moving boundary, nodal models with dynamic energy balances, dynamic mass balances, quasi-static momentum balances, and an equivalent single channel approach have been developed for steam generators used in nuclear power plants. The model for the U-tube recirculation type steam generator is described and comparisons are made of responses from models of different complexity; non-linear versus linear, high-order versus low order, detailed modeling of the control system versus a simple control assumption. The results of dynamic tests on nuclear power systems show that when this steam generator model is included in a system simulation there is good agreement with actual plant performance. (author)

  14. The casebook of technical presentation on a steam generator

    International Nuclear Information System (INIS)

    1986-05-01

    This casebook consists of seven presentations, which are measures and experience of maintenance of water quality in PWR generator, corrosion in steam generator, safely evaluation by management and closing in steam generator, testing of eddy current in steam generator, unsettled problems of safety in steam generator and maintenance of water quality in PWR generator.

  15. Steam generator design requirements for ACR-1000

    International Nuclear Information System (INIS)

    Subash, S.; Hau, K.

    2006-01-01

    Atomic Energy of Canada Limited (AECL) has developed the ACR-1000 (Advanced CANDU Reactor-1000 ) to meet market expectations for enhanced safety of plant operation, high capacity factor, low operating cost, increased operating life, simple component replacement, reduced capital cost, and shorter construction schedule. The ACR-1000 design is based on the use of horizontal fuel channels surrounded by a heavy water moderator, the same feature as in all CANDU reactors. The major innovation in the ACR-1000 is the use of low enriched uranium fuel, and light water as the coolant, which circulates in the fuel channels. This results in a compact reactor core design and a reduction of heavy water inventory, both contributing to a significant decrease in capital cost per MWe produced. The ACR-1000 plant is a two-unit, integrated plant with each unit having a nominal gross output of about 1165 MWe with a net output of approximately 1085 MWe. The plant design is adaptable to a single unit configuration, if required. This paper focuses on the technical considerations that went into developing some of the important design requirements for the steam generators for the ACR-1000 plant and how these requirements are specified in the Technical Specification, which is the governing document for the steam generator (SG) detail design. Layout of these SGs in the plant is briefly described and their impacts on the SG design. (author)

  16. Wavelet network controller for nuclear steam generators

    International Nuclear Information System (INIS)

    Habibiyan, H; Sayadian, A; Ghafoori-Fard, H

    2005-01-01

    Poor control of steam generator water level is the main cause of unexpected shutdowns in nuclear power plants. Particularly at low powers, it is a difficult task due to shrink and swell phenomena and flow measurement errors. In addition, the steam generator is a highly complex, nonlinear and time-varying system and its parameters vary with operating conditions. Therefore, it seems that design of a suitable controller is a necessary step to enhance plant availability factor. The purpose of this paper is to design, analyze and evaluate a water level controller for U-tube steam generators using wavelet neural networks. Computer simulations show that the proposed controller improves transient response of steam generator water level and demonstrate its superiority to existing controllers

  17. In service inspection for steam generator tubes

    International Nuclear Information System (INIS)

    Comby, R.; Eyrolles, Ph.

    1988-01-01

    In this paper the authors show the means putting in place for examination of steam generators tubes. These means (eddy current probes, ultrasonic testing) associated with a knowledge on degradation phenomena allow mapping controlled tubes and limiting undesirable obturations [fr

  18. Steam generators for nuclear power plants

    International Nuclear Information System (INIS)

    Tillequin, Jean

    1975-01-01

    The role and the general characteristics of steam generators in nuclear power plants are indicated, and particular types are described according to the coolant nature (carbon dioxide, helium, light water, heavy water, sodium) [fr

  19. US PWR steam generator management: An overview

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.

    1997-01-01

    This paper provides an overview on the status of steam generator management activities in US PWRs, and includes: (1) an overview of the impact of steam generator problems; (2) a brief discussion of historical damage trends and the current damage mechanism of most concern; (3) a discussion of the elements of open-quotes steam generator managementclose quotes; and (4) a description of the approach being followed to implement a degradation-specific protocol for tubing inspection and repair. This paper was prepared in conjunction with another paper presented during the Plenary Session of this Conference, open-quotes Steam Generator Degradation: Current Mitigation Strategies for Controlling Corrosionclose quotes, and is provided as a supplement to that material

  20. Forming a cohesive steam generator maintenance strategy

    International Nuclear Information System (INIS)

    Poudroux, G.

    1991-01-01

    In older nuclear plants, steam generator tube bundles are the most fragile part of the reactor coolant system. Steam generator tubes are subject to numerous types of loading, which can lead to severe degradation (corrosion and wear phenomena). Preventive actions, such as reactor coolant temperature reduction or increasing the plugging limit and their associated analyses, can increase steam generator service life. Beyond these preventive actions, the number of affected tubes and the different locations of the degradations that occur often make repair campaigns necessary. Framatome has developed and qualified a wide range of treatment and repair processes. They enable careful management of the repair campaigns, to avoid reaching the maximum steam generator tube plugging limit, while optimizing the costs. Most of the available repair techniques allow a large number of affected tubes to be treated. Here we look only at those techniques that should be taken into account when defining a maintenance strategy. (author)

  1. Steam generator in the SNR-project

    International Nuclear Information System (INIS)

    van Westenbrugge, J.K.

    1979-01-01

    The design philosophy of steam generators for 1300 MWe LMFBR's is presented. The basis for this philosophy is the present experience with the licensing of the SNR-300. This experience is reported. The approach for the steam generators for the 1300 MWe LMFBR is elaborated on, both for accident prevention and damage limitation, for the component itself as well as for the system design. Both Design Base Accident and Hypothetical Accidents are discussed. 8 refs

  2. New steam generators slated for nuclear units

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This article is a brief discussion of Duke Power's plans to replace steam generators at its McGuire and Catawba nuclear units. A letter of intent to purchase (from Babcock and Wilcox) the 12 Westinghouse steam generators has been signed, but no constructor has been selected at this time. This action is brought about by the failures of more than 3000 tubes in these units

  3. Development of a steam generator lancing system

    International Nuclear Information System (INIS)

    Jeong, Woo-Tae; Kim, Seok-Tae; Hong, Sung-Yull

    2006-01-01

    It is recommended to clean steam generators of nuclear power plants during plant outages. Under normal operations, sludge is created and constantly accumulates in the steam generators. The constituents of this sludge are different depending on each power plant characteristics. The sludge of the Kori Unit 1 steam generator, for example, was found to be composed of 93% ferrous oxide, 3% carbon and 1% of silica oxide and nickel oxide each. The research to develop a lancing system that would remove sludge deposits from the tubesheet of a steam generator was started in 1998 by the Korea Electric Power Research Institute (KEPRI) of the Korea Electric Power Corporation (KEPCO). The first commercial domestic lancing system in Korea, and KALANS-I Lancing System, was completed in 2000 for Kori Unit 1 for cleaning the tubesheet of its Westinghouse Delta-60 steam generator. Thereafter, the success of the development and site implementation of the KALANS-I lancing system for YGN Units 1 and 2 and Ulchin Units 3 and 4 was also realized in 2004 for sludge removal at those sites. The upper bundle cleaning system for Westinghouse model F steam generators is now under development

  4. Steam generator tube fitness-for-service guidelines

    International Nuclear Information System (INIS)

    Gorman, J.A.; Harris, J.E.; Lowenstein, D.B.

    1995-07-01

    The objectives of this project were to characterize defect mechanisms which could affect the integrity of steam generator tubes, to review and critique state-of-the-art Canadian and international steam generator tube fitness-for-service criteria and guidelines, and to obtain recommendations for criteria that could be used to assess fitness-for service guidelines for steam generator tubes containing defects in Canadian power plant service. Degradation mechanisms, that could affect CANDU steam generator tubes in Canada, have been characterized. The design standards and safety criteria that apply to steam generator tubing in nuclear power plant service in Canada and in Belgium, France, Japan, Spain, Sweden, and the USA have been reviewed and described. The fitness-for-service guidelines used for a variety of specific defect types in Canada and internationally have been evaluated and described in detail in order to highlight the considerations involved in developing such defect specific guidelines. Existing procedures for defect assessment and disposition have been identified, including inspection and examination practices. The approaches used in Canada and in Belgium, France, Japan, Spain, Sweden, and the USA for fitness-for-service guidelines were compared and contrasted for a variety of defect mechanisms. The strengths and weaknesses of the various approaches have been assessed. The report presents recommendations on approaches that may be adopted in the development of fitness-for-service guidelines for use in the dispositioning of steam generator tubing defects in Canada. (author). 175 refs., 2 tabs., 28 figs

  5. Steam generator tube rupture simulation using extended finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurin; Natesan, Ken

    2016-08-15

    Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

  6. Steam generator tube rupture simulation using extended finite element method

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurin; Natesan, Ken

    2016-01-01

    Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

  7. CRBRP steam-generator design evolution

    International Nuclear Information System (INIS)

    Geiger, W.R.; Gillett, J.E.; Lagally, H.O.

    1983-01-01

    The overall design of the CRBRP Steam Generator is briefly discussed. Two areas of particular concern are highlighted and considerations leading to the final design are detailed. Differential thermal expansion between the shell and the steam tubes is accommodated by the tubes flexing in the curved section of the shell. Support of the tubes by the internals structure is essential to permit free movement and minimize tube wear. Special spacer plate attachment and tube hole geometry promote unimpeded axial movement of the tubes by allowing individual tubes to rotate laterally and by providing lateral movement of the spacer plates relative to the adjacent support structure. The water/steam heads of the CRBRP Steam Generator are spherical heads welded to the lower and upper tubesheets. They were chosen principally because they provide a positively sealed system and result in more favorable stresses in the tubesheets when compared to mechanically attached steamheads

  8. 49 CFR 229.105 - Steam generator number.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Steam generator number. 229.105 Section 229.105..., DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Safety Requirements Steam Generators § 229.105 Steam generator number. An identification number shall be marked on the steam generator's...

  9. Steam generation at Rihand STPP Stage 1

    Energy Technology Data Exchange (ETDEWEB)

    1987-12-01

    The steam generation plant at Rihand in India has two 500 MW boilers. The boilers are of the balanced draught, single cell, radiant furnace type, and are controlled automatically. Cochran Thermax shell type auxillary steam boilers are used for preheating air to the main boilers and for heating fuel oil during storage and pumping. Electrostatic precipitators and ash handling plants are provided to keep dust and ash within limits. 2 figs.

  10. Chemical cleaning of nuclear (PWR) steam generators

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Mundis, J.A.

    1982-01-01

    This paper reports on a significant research program sponsored by a group of utilities (the Steam Generator Owners Group), which was undertaken to develop a process to chemically remove corrosion product deposits from the secondary side of pressurized water reactor (PWR) power plant steam generators. Results of this work have defined a process (solvent system and application methods) that is capable of removing sludge and tube-to-tube support plate crevice corrosion products generated during operation with all-volatile treatment (AVT) water chemistry. Considers a plant-specific test program that includes all materials in the steam generator to be cleaned and accounts for the physical locations (proximity and contact) of those materials. Points out that prior to applying the process in an operational unit, the utility, with the participation of the NSSR vendor, must define allowable total corrosion to the materials of construction of the unit

  11. Non-polluting steam generators with fluidized-bed furnaces

    Energy Technology Data Exchange (ETDEWEB)

    Brandes, H [Deutsche Babcock A.G., Oberhausen (Germany, F.R.)

    1979-07-01

    The author reports on a 35 MW steam generator with hard coal fluidized-bed furnace a planned 35 MW steam generator with flotation-dirt fluidized-bed furnace, and on planned steam generators for fluidized-bed firing of hard coal up to a steam power of about 200 MW.

  12. Program for parameter studies of steam generators

    International Nuclear Information System (INIS)

    Mathisen, R.P.

    1982-11-01

    R2-GEN is a computer code for stationary thermal parameter studies of steam generators. The geometry and data are valid for Ringhals-2 generators. Subroutines and relevant calculations are included. The program is based on a heterogeneous flow model and some applications on tubes with varying contamination are presented. (G.B.)

  13. Development of steam generator manufacturing technology

    International Nuclear Information System (INIS)

    Grant, J.A.

    1979-01-01

    In 1968 Babcock and Wilcox (Operations) Ltd., received an order from the CEGB to design, manufacture, install and commission 16 Steam Generators for 2 x 660 Mw (e) Advanced Gas Cooled Reactor Power Station at Hartlepool. This order was followed in 1970 by a similar order for the Heysham Power Station. The design and manufacture of the Steam Generators represented a major advance in technology and the paper discusses the methods by which a manufacturing facility was developed, by the Production Division of Babcock, to produce components to a quality, complexity and accuracy unique in the U.K. commercial boilermaking industry. The discussion includes a brief design background, a description of the Steam Generators and a view of the Production Division background. This is followed by a description of the organisation of the technological development and a consideration of the results. (author)

  14. Corrosion problems in PWR steam generators

    International Nuclear Information System (INIS)

    Weber, J.; Suery, P.

    1976-01-01

    Examinations on pulled steam generator tubes from the Swiss nuclear power plants Beznau I and II, together with some laboratory tests, may be summarized as follows: Corrosion problems in vertical U-tube steam generators with Alloy 600 as tube material are localized towards relatively narrow regions above the tube sheet where thermohydraulic conditions and, as a consequence thereof, chemical conditions are uncontrolled. Within these zones Alloy 600 is not sufficienthy resistent to caustic or phosphate attack (caustic stress corrosion cracking and general corrosion, resp.). The mechanisms of several corrosion phenomena are not fully understood. (orig.) [de

  15. Acoustic leak detector in Monju steam generator

    International Nuclear Information System (INIS)

    Wachi, E.; Inoue, T.

    1990-01-01

    Acoustic leak detectors are equipped with the Monju steam generators for one of the R and D activities, which are the same type of the detectors developed in the PNC 50MW Steam Generator Test Facility. Although they are an additional leak detection system to the regular one in Monju SG, they would also detect the intermediate or large leaks of the SG tube failures. The extrapolation method of a background noise analysis is expected to be verified by Monju SG data. (author). 4 figs

  16. PWR steam generator tubing sample library

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    In order to compile the tubing sample library, two approaches were employed: (a) tubing sample replication by either chemical or mechanical means, based on field tube data and metallography reports for tubes already destructively examined; and (b) acquisition of field tubes removed from operating or retired steam generators. In addition, a unique mercury modeling concept is in use to guide the selection of replica samples. A compendium was compiled that summarizes field observations and morphologies of steam generator tube degradation types based on available NDE, destructive examinations, and field reports. This compendium was used in selecting candidate degradation types that were manufactured for inclusion in the tube library

  17. Stability study in one step steam generators

    International Nuclear Information System (INIS)

    Abbate, P.

    1990-01-01

    The TWO program is presented developed for the behaviour limit calculation stable in one step steam generators for the case of Density Waves phenomenom. The program is based on a nodal model which, using Laplace transformation equations, allows to study the system's transfer functions and foresee the beginning of the unstable behaviour. This program has been satisfactorily validated against channels data uniformly heated in the range from 4.0 to 6.0 Mpa. Results on the CAREM reactor's steam generator analysis are presented. (Author) [es

  18. Steam generator tubesheet waterlancing at Bruce B

    Energy Technology Data Exchange (ETDEWEB)

    Persad, R. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada); Eybergen, D. [Bruce Power, Tiverton, Ontario (Canada)

    2006-07-01

    High pressure water cleaning of steam generator secondary side tubesheet surfaces is an important and effective strategy for reducing or eliminating under-deposit chemical attack of the tubing. At the Bruce B station, reaching the interior of the tube bundle with a high-pressure water lance is particularly challenging due to the requirement to setup on-boiler equipment within the containment bellows. This paper presents how these and other design constraints were solved with new equipment. Also discussed is the application of new high-resolution inter-tube video probe capability to the Bruce B steam generator tubesheets. (author)

  19. Expandable antivibration bar for a steam generator

    International Nuclear Information System (INIS)

    Lagally, H.O.

    1986-01-01

    A steam generator tube support structure comprises expandable antivibration bars positioned between rows of tubes in the steam generator and attached to retaining rings surrounding the bundle of tubes. The antivibration bars have adjacent bar sections with mating surfaces formed as inclined planes which upon relative longitudinal motion between the upper and lower bars provides a means to increase the overall thickness across the structure to the required value. The bar section is retained against longitudinal movement in take-up assembly whereas the bar section is movable longitudinally by rotation of a nut. (author)

  20. Steam generators regulatory practices and issues in Spain

    International Nuclear Information System (INIS)

    Mendoza, C.; Castelao, C.; Ruiz-Colino, J.; Figueras, J.M.

    1997-01-01

    This paper presents the actual status of Spanish Steam Generator tubes, actions developed by PWR plant owners and submitted to CSN, and regulatory activities related to tube degradation mechanisms analysis; NDT tube inspection techniques; tube, tubesheet and TSPs integrity studies; tube plugging/repair criteria; preventive and corrective measures including whole SGs replacement; tube leak measurement methods and other operational aspects

  1. GRUVAL for ET inspection of the steam generator tubes

    International Nuclear Information System (INIS)

    Garcia Bueno, A.; Francia, L.; Jimenez Garcia, J. J.; Garcia, R.; Castelinou, M.; Torrens, J.

    2013-01-01

    The steam generators of the nuclear power plants, PWR type are one of the most important components from the point of view of safety and plant availability. Thousands of tubes that form, approximately 1 mm of thickness, required to be inspected in accordance with codes and standards, to ensure the integrity of the component during the operation of the plant.

  2. Conceptual design of once-through helical steam generator for SMART

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Wan; Kim, J. I.; Kim, J. H. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    Conceptual design of once-through helical steam generator for the integral reactor SMART is developed. The once-through helical steam generator requires quite different design concepts from the steam generators used in loop type commercial reactors. In this study the design requirements satisfying the operating conditions of the steam generator are derived, and the arrangements and the dimensions of the major parts are determined. By describing the design procedure, the cost of redesign and the costs of developments of similar new steam generators are minimized. The three dimensional models developed make it possible to preview the interferences of the steam generator components and to minimize the possibility of significant design changes in the next design stage by the preliminary strength analysis of the major parts. A methodology for evaluation of flow induced vibration of steam generator tubes has been developed and a preliminary flow induced vibration analysis has been performed. 24 refs., 54 figs., 9 tabs. (Author)

  3. Slurry steam generator program and baseline eddy current examination

    International Nuclear Information System (INIS)

    Clark, R.A.; Doctor, P.G.

    1985-01-01

    The Steam Generator Group Project was initiated in January 1982 with formation of consortium including NRC, EPRI, Japanese, French, and Italian participants. The project utilizes a retired-from-service nuclear steam generator established in a specially designed facility which houses the unit in its normal vertical operating position. The most important objectives deal with validation of nondestructive examination (NDE) techniques used to characterize steam generators during service. This research generator offers the first opportunity to characterize a statistically significant number of service-induced defects nondestructively followed by destructive metallographic confirmation. The project seeks to establish the reliability of defect detection and the accuracy of sizing defects via current state-of-the-art NDE. Other service degraded tubes will be burst tested to establish remaining service integrity. The integrity information and NDE reliability results will serve as inputs to establish a model for steam generator in-service inspections, and provide a data base for evaluation of tube plugging criteria. In addition to NDE validation goals, the project will use the service degraded generator as a specimen for demonstration/proof testing of repair and maintenance techniques, including chemical cleaning/decontamination technologies. In addition to the efforts associated with NDE, a multitude of other project tasks have continued through 1984, and results are presented

  4. Surry steam generator program and baseline eddy current examination

    International Nuclear Information System (INIS)

    Clark, R.A.; Doctor, P.G.

    1985-01-01

    The Steam Generator Group Project was initiated in January 1982 with formation of consortium including NRC, EPRI, Japanese, French, and Italian participants. The project utilizes a retired-from-service nuclear steam generator established in a specially designed facility which houses the unit in its normal vertical operating position. The most important objectives deal with validation of nondestructive examination (NDE) techniques used to characterize steam generators during service. This research generator offers the first opportunity to characterize a statistically significant number of service-induced defects nondestructively followed by destructive metallographic confirmation. The project seeks to establish the reliability of defect detection and the accuracy of sizing defects via current state-of-the-art NDE. Other service degraded tubes will be burst tested to establish remaining service integrity. The integrity information and NDE reliability results will serve as inputs to establish a model for steam generator in-service inspections, and provide a data base for evaluation of tube plugging criteria. In addition to NDE validation goals, the project will use the service degraded generator as a specimen for demonstration/proof testing of repair and maintenance techniques, including chemical cleaning/decontamination technologies. In addition to the efforts associated with NDE, a multitude of other project tasks have continued through 1984, and results are presented

  5. Passive system with steam-water injector for emergency supply of NPP steam generators

    International Nuclear Information System (INIS)

    Il'chenko, A.G.; Strakhov, A.N.; Magnitskij, D.N.

    2009-01-01

    The calculation results of reliability indicators of emergency power supply system and emergency feed-water supply system of serial WWER-1000 unit are presented. To ensure safe water supply to steam generators during station blackout it was suggested using additional passive emergency feed-water system with a steam-water injector working on steam generators dump steam. Calculated analysis of steam-water injector operating capacity was conducted at variable parameters of steam at the entrance to injector, corresponding to various moments of time from the beginning of steam-and-water damping [ru

  6. LPGC, Levelized Steam Electric Power Generator Cost

    International Nuclear Information System (INIS)

    Coen, J.J.; Delene, J.G.

    1994-01-01

    1 - Description of program or function: LPGC is a set of nine microcomputer programs for estimating power generation costs for large steam-electric power plants. These programs permit rapid evaluation using various sets of economic and technical ground rules. The levelized power generation costs calculated may be used to compare the relative economics of nuclear and coal-fired plants based on life-cycle costs. Cost calculations include capital investment cost, operation and maintenance cost, fuel cycle cost, decommissioning cost, and total levelized power generation cost. These programs can be used for quick analyses of power generation costs using alternative economic parameters, such as interest rate, escalation rate, inflation rate, plant lead times, capacity factor, fuel prices, etc. The two major types of electric generating plants considered are pressurized-water reactor (PWR) and pulverized coal-fired plants. Data are also provided for the Large Scale Prototype Breeder (LSPB) type liquid metal reactor. Costs for plant having either one or two units may be obtained. 2 - Method of solution: LPGC consists of nine individual menu-driven programs controlled by a driver program, MAINPWR. The individual programs are PLANTCAP, for calculating capital investment costs; NUCLOM, for determining operation and maintenance (O and M) costs for nuclear plants; COALOM, for computing O and M costs for coal-fired plants; NFUEL, for calculating levelized fuel costs for nuclear plants; COALCOST, for determining levelized fuel costs for coal-fired plants; FCRATE, for computing the fixed charge rate on the capital investment; LEVEL, for calculating levelized power generation costs; CAPITAL, for determining capitalized cost from overnight cost; and MASSGEN, for generating, deleting, or changing fuel cycle mass balance data for use with NFUEL. LPGC has three modes of operation. In the first, each individual code can be executed independently to determine one aspect of the total

  7. Eddy current testing of steam generator tubes

    International Nuclear Information System (INIS)

    Neumaier, P.

    1981-01-01

    A rotating probe is described for improving the inspection of tubes and end plate in steam generators. The method allows a representation of the whole defect, consequently the observer is able to determine directly the type of defect, signal processing in-line or off-line is possible [fr

  8. Operating conditions of steam generators for LMFBR's

    International Nuclear Information System (INIS)

    Ratzel, W.

    1975-01-01

    Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

  9. Operating conditions of steam generators for LMFBR's

    Energy Technology Data Exchange (ETDEWEB)

    Ratzel, W

    1975-07-01

    Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

  10. Open channel steam generator feedwater system

    International Nuclear Information System (INIS)

    Kim, R.F.; Min-Hsiung Hu.

    1985-01-01

    A steam generator which utilizes a primary fluid to vaporize a secondary fluid is provided with an open flow channel and elevated discharge nozzle for the introduction of secondary fluid. The discharge nozzle is positioned above a portion of the inlet line such that the secondary fluid passes through a vertical section of inlet line prior to its discharge into the open channel. (author)

  11. KNK steam generator leakage, evaluations and improvements

    International Nuclear Information System (INIS)

    Dumm, K.; Ratzel, W.

    1975-01-01

    On September 23, 1972 the KNK reactor was shut down due to a sodium-water reaction. Detection, localisation and possible leak development are described and discussed. Improvements on the KNK steam generator system due to this leak experience are explained. (author)

  12. KNK steam generator leakage, evaluations and improvements

    Energy Technology Data Exchange (ETDEWEB)

    Dumm, K; Ratzel, W

    1975-07-01

    On September 23, 1972 the KNK reactor was shut down due to a sodium-water reaction. Detection, localisation and possible leak development are described and discussed. Improvements on the KNK steam generator system due to this leak experience are explained. (author)

  13. Method for servicing a steam generator

    International Nuclear Information System (INIS)

    Cooper, J.W. Jr.; Castner, R.P.

    1982-01-01

    The servicing of a steam generator is made easier by mapping the tubesheet with a remotely controlled probe to locate precisely each hole in the sheet. The locations are stored and used to maneuver various tools into position to perform operations on each tube hole

  14. Thermohydraulic verification during THTR steam generator commissioning

    International Nuclear Information System (INIS)

    Henry, C.; Elter, C.

    1988-01-01

    In one of the six THTR 300 steam generators thermocouples are installed inside the heat transfer tube bundles for measuring the gas and steam temperatures. Fluid temperature distribution measurements along and across the helix bundle have been recorded in its first months of operation over a load range of 40% up to 100% for steady state and transient conditions. Using these measurements as well as the rest of the operating instrumentation. the computer programs for the design of heat exchanger heat transfer areas are verified. The temperature measurements for steady state conditions are compared with predictions obtained in the design stage. In these codes. the heat transferred from the outside helium gas to the water/steam inside the tubes is determined in discrete steps along the heating surface by one- and two-phase heat transfer correlations. The degree of conformity between prediction and measurement is discussed and compared with more recent correlations. (author)

  15. Stress analysis of HTR-10 steam generator heat exchanging tubes

    International Nuclear Information System (INIS)

    Dong Jianling; Zhang Xiaohang; Yin Dejian; Fu Jiyang

    2001-01-01

    Steam Generator (SG) heat exchanging tubes of 10 MW High Temperature Gas Cooled Reactor (HTR-10) are protective screens between the primary loop of helium with radioactivity and the secondary loop of feeding water and steam without radioactivity. Water and steam will enter into the primary loop when rupture of the heat exchanging tubes occurs, which lead to increase of the primary loop pressure and discharge of radioactive materials. Therefore it is important to guarantee the integrity of the tubes. The tube structure is spiral tube with small bending radius, which make it impossible to test with volumetric in-service detection. For such kind of spiral tube, using LBB concept to guarantee the integrity of the tubes is an important option. The author conducts stress analysis and calculation of HTR-10 SG heat exchanging tubes using the FEM code of piping stress analysis, PIPESTRESS. The maximum stress and the dangerous positions are obtained

  16. Corrosion problems of PWR steam generators

    International Nuclear Information System (INIS)

    Urbancik, L.; Kostal, M.

    Literature data are assessed on corrosion failures of steam generator tubes made of INCONEL 600 or INCOLOY 800. It was found that both alloys with high nickel content showed good stability in a corrosion environment while being sensitive to carbide formation on grain boundaries. The gradual depletion of chromium results from the material and corrosion resistance deteriorates. INCOLOY 800 whose chromium carbide precipitation on grain boundaries in pure water and steam is negligible up to 75O degC and which is not subject to corrosion attacks in the above media and in an oxidizing environment at a temperature to about 700 degC shows the best corrosion resistance. Its favourable properties were tested in long-term operation in the Peach Bottom 1 nuclear power plant where no failures due to corrosion of this material have been recorded since 1967. In view of oxygenic-acid surface corrosion, it is necessary to work in a neutral or slightly basic environment should any one of the two alloys be used for steam generator construction. The results are summed up of an analysis conducted for the Beznau I NOK reactor. Water treatment with ash-free amines can be used as prevention against chemical corrosion mechanisms, although the treatment itself does not ensure corrosion resistance of steam generator key components. (J.B.)

  17. Dynamic simulation of steam generator failures

    Energy Technology Data Exchange (ETDEWEB)

    Meister, G [Institut fuer Nukleare Sicherheitsforschung, Kernforschungsanlage Juelich GmbH, Juelich (Germany)

    1988-07-01

    A computer program will be described which is capable to simulate severe transients in a gas heated steam generator. Such transients may arise in the safety analysis of accidents resulting from failures in the heat removal system of an HTGR power plant. Important failure modes which have to be considered are ruptures of one or more steam generator tubes leading to water or steam ejection into the primary system or anomalous operating conditions which my cause damage due to excessive thermal stress. Examples are the complete dryout as a consequence of feedwater interrupt in connection with continuing gas heating and the reflooding of the secondary channel with cold feedwater after dryout. The steam generator program which is capable to simulate accidents of this type is written as a module which can be implemented into a program system fur the simulation of the total heat rejection system. It based on an advanced mathematical model for the two phase flow taking deviations from thermal equilibrium into account. Mass, energy and momentum balances for the primary and secondary fluid and the heat diffusion equations for the heat exchanging wall form a system of coupled differential equations which is solved numerically by an algorithm which is stiffly stable and suppresses effectively oscillations of numerical origin. Results of the simulation of transients of the type mentioned above will be presented and discussed. (author)

  18. Leakage experiences with 1 MW steam generator

    International Nuclear Information System (INIS)

    Kanamori, A.; Kawara, M.; Sano, A.

    1975-01-01

    An 1 MW steam generator was tested from October, 1971 and completed with the first series of experiments by May, 1972 after 3600 hours of operation. During these tests, unextraordinary heat absorption was experienced in the downcomer region, which led to shortage of heat transfer area to attain the rated steam temperature and to one of the reasons of flow instabilities. The steam generator was disassembled to get test pieces for structure as well as material examinations and then it was reassembled to proceed the second series of tests. Before it was done, a modification was provided to insulate the downcomer region by putting a gas space around the downcomer tube. The gas space was provided by a dual tube and spacers were welded on the inner tube and an end plate was welded on upper parts between the two to seal the gap by means of fillet welding. After the modified steam generator was put into operation, water happened to leak into a sodium side two times through these additional welding spots for the gas insulation. This paper presents operating conditions and behaviors of monitors at the time of the leakages, identifications of leaked spots, an evaluation of causes and a treatment or a precaution for them

  19. Dynamic simulation of steam generator failures

    International Nuclear Information System (INIS)

    Meister, G.

    1988-01-01

    A computer program will be described which is capable to simulate severe transients in a gas heated steam generator. Such transients may arise in the safety analysis of accidents resulting from failures in the heat removal system of an HTGR power plant. Important failure modes which have to be considered are ruptures of one or more steam generator tubes leading to water or steam ejection into the primary system or anomalous operating conditions which my cause damage due to excessive thermal stress. Examples are the complete dryout as a consequence of feedwater interrupt in connection with continuing gas heating and the reflooding of the secondary channel with cold feedwater after dryout. The steam generator program which is capable to simulate accidents of this type is written as a module which can be implemented into a program system fur the simulation of the total heat rejection system. It based on an advanced mathematical model for the two phase flow taking deviations from thermal equilibrium into account. Mass, energy and momentum balances for the primary and secondary fluid and the heat diffusion equations for the heat exchanging wall form a system of coupled differential equations which is solved numerically by an algorithm which is stiffly stable and suppresses effectively oscillations of numerical origin. Results of the simulation of transients of the type mentioned above will be presented and discussed. (author)

  20. Monitoring method for steam generator operation

    International Nuclear Information System (INIS)

    Tamaoki, Tetsuo

    1991-01-01

    In an LMFBR plant having an once-through steam generator, reduction of life of a heat transfer pipe caused by heat cycle fatigue is monitored by early finding for the occurrence of abnormality in the inside of the steam generator and by continuous monitoring for the position of departure from nucleate boiling (DNB), which are difficult with existent static characteristic analysis codes. That is, RMS values of fluctuations in temperature signals sent from thermocouples for measuring the fluid temperature in the vicinity of heat transfer pipe disposed along a primary channel of the once-through type steam generator. The abnormality in heat transfer performance is monitored by the distribution change of the RMS values. Subsequently, DNB point on the side of water and steam is determined by the distribution of the RMS value. Then, accumulated values of the product between the time in which the starting point stays in the DNB region and a life consumption amount per unit time given in accordance with the operation condition are monitored. Accordingly, thermal fatigue failure of the heat transfer pipe due to temperature fluctuation in the DNB region is monitored. (I.S.)

  1. Solar energy for steam generation in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    De Carvalho, Jr, A V; Orlando, A DeF; Magnoli, D

    1979-05-01

    Steam generation is a solar energy application that has not been frequently studied in Brazil, even though for example, about 10% of the national primary energy demand is utilized for processing heat generation in the range of 100 to 125/sup 0/C. On the other hand, substitution of automotive gasoline by ethanol, for instance, has received much greater attention even though primary energy demand for process heat generation in the range of 100 to 125/sup 0/C is of the same order of magnitude than for total automotive gasoline production. Generation of low-temperature steam is analyzed in this article using distributed systems of solar collectors. Main results of daily performance simulation of single flat-plate collectors and concentrating collectors are presented for 20/sup 0/S latitude, equinox, in clear days. Flat plate collectors considered are of the aluminum roll-bond absorber type, selective surface single or double glazing. Considering feedwater at 20/sup 0/C, saturated steam at 120/sup 0/C and an annual solar utilization factor of 50%, a total collector area of about 3,000 m/sup 2/ is necessary for the 10 ton/day plant, without energy storage. A fuel-oil back-up system is employed to complement the solar steam production, when necessary. Preliminary economic evaluation indicates that, although the case-study shows today a long payback period relative to subsidized fuel oil in the domestic market (over 20 years in the city of Rio de Janeiro), solar steam systems may be feasible in the medium term due to projected increase of fuel oil price in Brazil.

  2. Steam generation from solar energy

    International Nuclear Information System (INIS)

    Gozzi, M.

    2001-01-01

    The vapor for thermoelectric use is one of the most promoted methods for electric power generation from solar energy. The new plants are becoming more and more safe, and anyway in some cases the natural gas makes easy the production of electricity [it

  3. CATHENA simulations of steam generator tube rupture

    International Nuclear Information System (INIS)

    Abdul-Razzak, A.; Lin, M.R.; Wright, A.C.D.

    1997-01-01

    The CATHENA thermalhydraulic computer code was used to simulate various scenarios following a CANDU 9 steam generator tube rupture (SGTR) event. The analysis included cases with class IV power and emergency core cooling system (ECCS) available and other cases with subsequent loss of class IV power (LCIVP) or impairment of ECCS injection. Two main approaches were followed in the analysis of each case. In the first approach, D 2 O feed was credited to provide conservative data for input to radionuclide release and dose calculations. Also operator actions are credited. The other approach is designed to give conservative predictions with respect to the acceptance criteria of fuel and fuel channel integrity and to prove that in case of such event, the operator will have enough time to mitigate the consequences. This is done by not crediting makeup for the inventory loss and relying on the automatic operation of safety systems. The analysis of the cases of the first approach provided the required data for radionuclide release and dose calculations and gave a good insight into the required sequence of operator timely actions to mitigate the consequences of such event. On the other hand, the cases of the second approach confirmed compliance with regulatory requirements for pressure tube and fuel integrity. The runs with ECCS available, showed the ECCS injection is effective in filling and cooling the core and that regulatory requirement's for fuel and channel integrity are met. In the event of ECCS impairment, the earliest indication of late fuel heat-up is late enough to provide the operator with an adequate time to act in mitigating the consequences of this event. (author)

  4. Steam generator tube failures: experience with water-cooled nuclear power reactors during 1976

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1978-02-01

    A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author)

  5. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mendler, O J; Takeuchi, K; Young, M Y

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

  6. Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

    International Nuclear Information System (INIS)

    Mendler, O.J.; Takeuchi, K.; Young, M.Y.

    1986-10-01

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results

  7. Electrosleeve process for in-situ nuclear steam generator repair

    International Nuclear Information System (INIS)

    Renaud, E.; Brennenstuhl, A.M.; Stewart, D.R.; Gonzalez, F.

    2000-01-01

    Degradation of steam generator tubing by localized corrosion is a widespread problem in the nuclear industry that can lead to costly forced outages, unit derating, steam generator replacement or even the permanent shutdown of a reactor. In response to the onset of steam generator degradation at Ontario Power Generation's Pickering Nuclear Generating Station (PNGS) Unit 5, and the determined unsuitability of conventional repair methods (mechanically expanded or welded sleeves) for Alloy 400, an alternative repair technology was developed. Electrosleeve is a non-intrusive, low-temperature process that involves the electrodeposition of a nanocrystalline nickel microalloy forming a continuously bonded, structural layer over the internal diameter of the degraded region. This technology is designed to provide a long-term pressure boundary repair, fully restoring the structural integrity of the damaged region to its original state. This paper describes the Electrosleeve process for steam generator tubing repair and the unique properties of the advanced sleeve material. The successful installation of fourteen Electrosleeves that have been in service for more than six years in Alloy 400 tubing at the Pickering-S CANDU unit, and the more recent (Nov. 99) extension of the technology to Alloy 600 by the installation of 57 sleeves in a U.S. pressurized water reactor (PWR) at Callaway, is presented. The Electrosleeve process has been granted a conditional license by the U.S. Nuclear Regulatory Commission (NRC). In Canada, the process of licensing Electrosleeve with the CNSC / TSSA has begun. (author)

  8. On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers

    International Nuclear Information System (INIS)

    Upadhyaya, Belle R.; Hines, J. Wesley

    2004-01-01

    Integrity monitoring and flaw diagnostics of flat beams and tubular structures was investigated in this research task using guided acoustic signals. A piezo-sensor suite was deployed to activate and collect Lamb wave signals that propagate along metallic specimens. The dispersion curves of Lamb waves along plate and tubular structures are generated through numerical analysis. Several advanced techniques were explored to extract representative features from acoustic time series. Among them, the Hilbert-Huang transform (HHT) is a recently developed technique for the analysis of non-linear and transient signals. A moving window method was introduced to generate the local peak characters from acoustic time series, and a zooming window technique was developed to localize the structural flaws. The time-frequency analysis and pattern recognition techniques were combined for classifying structural defects in brass tubes. Several types of flaws in brass tubes were tested, both in the air and in water. The techniques also proved to be effective under background/process noise. A detailed theoretical analysis of Lamb wave propagation was performed and simulations were carried out using the finite element software system ABAQUS. This analytical study confirmed the behavior of the acoustic signals acquired from the experimental studies. The report presents the background the analysis of acoustic signals acquired from piezo-electric transducers for structural defect monitoring. A comparison of the use of time-frequency techniques, including the Hilbert-Huang transform, is presented. The report presents the theoretical study of Lamb wave propagation in flat beams and tubular structures, and the need for mode separation in order to effectively perform defect diagnosis. The results of an extensive experimental study of detection, location, and isolation of structural defects in flat aluminum beams and brass tubes are presented. The results of this research show the feasibility of on

  9. On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Belle R. Upadhyaya; J. Wesley Hines

    2004-09-27

    Integrity monitoring and flaw diagnostics of flat beams and tubular structures was investigated in this research task using guided acoustic signals. A piezo-sensor suite was deployed to activate and collect Lamb wave signals that propagate along metallic specimens. The dispersion curves of Lamb waves along plate and tubular structures are generated through numerical analysis. Several advanced techniques were explored to extract representative features from acoustic time series. Among them, the Hilbert-Huang transform (HHT) is a recently developed technique for the analysis of non-linear and transient signals. A moving window method was introduced to generate the local peak characters from acoustic time series, and a zooming window technique was developed to localize the structural flaws. The time-frequency analysis and pattern recognition techniques were combined for classifying structural defects in brass tubes. Several types of flaws in brass tubes were tested, both in the air and in water. The techniques also proved to be effective under background/process noise. A detailed theoretical analysis of Lamb wave propagation was performed and simulations were carried out using the finite element software system ABAQUS. This analytical study confirmed the behavior of the acoustic signals acquired from the experimental studies. The report presents the background the analysis of acoustic signals acquired from piezo-electric transducers for structural defect monitoring. A comparison of the use of time-frequency techniques, including the Hilbert-Huang transform, is presented. The report presents the theoretical study of Lamb wave propagation in flat beams and tubular structures, and the need for mode separation in order to effectively perform defect diagnosis. The results of an extensive experimental study of detection, location, and isolation of structural defects in flat aluminum beams and brass tubes are presented. The results of this research show the feasibility of on

  10. Steam-generator replacement sets new marks

    International Nuclear Information System (INIS)

    Beck, R.L.

    1995-01-01

    This article describes how, in one of the most successful steam-generator replacement experiences at PWRs worldwide, the V C Summer retrofit exceeded plant goals for critical-path duration, radiation, exposure, and radwaste generation. Intensive planning and teamwork, combined with the firm support of station management and the use of mockups to prepare the work crews for activity in a radiological environment, were key factors in the record performance achieved by South Carolina Electric and Gas Co (SCE and G) in replacing three steam generators at V C Summer nuclear station. The 97-day, two-hour breaker-to-breaker replacement outage -- including an eight-day delay for repair of leak in a small-bore seal-injection line of a reactor coolant pump (unrelated to the replacement activities) -- surpassed the project goal by over one day. Moreover, the outage was only 13 hours shy of the world record held by Virginia Power Co's North Anna Unit 1

  11. Steam Generator Group Project. Annual report, 1982

    International Nuclear Information System (INIS)

    Clark, R.A.; Lewis, M.

    1984-02-01

    The Steam Generator Group Project (SGGP) is an NRC program joined by additional sponsors. The SGGP utilizes a steam generator removed from service at a nuclear plant (Surry 2) as a vehicle for research on a variety of safety and reliability issues. This report is an annual summary of progress of the program for 1982. Information is presented on the Steam Generator Examination Facility (SGEF), especially designed and constructed for this research. Loading of the generator into the SGEF is then discussed. The report then presents radiological field mapping results and personnel exposure monitoring. This is followed by information on field reduction achieved by channel head decontaminations. The report then presents results of a secondary side examination through shell penetrations placed prior to transport, confirming no change in generator condition due to transport. Decontamination of the channel head is discussed followed by plans for eddy current testing and removal of the plugs placed during service. Results of a preliminary profilometry examination are then provided

  12. Future aspects for liquid metal heated steam generators

    International Nuclear Information System (INIS)

    Jansing, W.; Ratzel, W.; Vinzens, K.

    1975-01-01

    The present status of steam generators is shown. The experience gained until now is expressed in form of basic points. The most important design criteria for steam generator systems are outlined. On the basis of these design criteria, two possible steam generator concepts are shown. Costs in relationship to the repair concepts of two modular steam generators (thermal output 156 and 625 MW) and a pool design of 625 MW are compared. (author)

  13. Future aspects for liquid metal heated steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Jansing, W; Ratzel, W; Vinzens, K

    1975-07-01

    The present status of steam generators is shown. The experience gained until now is expressed in form of basic points. The most important design criteria for steam generator systems are outlined. On the basis of these design criteria, two possible steam generator concepts are shown. Costs in relationship to the repair concepts of two modular steam generators (thermal output 156 and 625 MW) and a pool design of 625 MW are compared. (author)

  14. CASTOR - Advanced System for VVER Steam Generator Inspection

    International Nuclear Information System (INIS)

    Mateljak, Petar

    2014-01-01

    From the safety point of view, steam generator is a very important component of a nuclear power plant. Only a thin tube wall prevents leakage of radioactive material from the primary side into the environment. Therefore, it is very important to perform inspections in order to detect pipe damage and apply appropriate corrective actions during outage. Application of the nondestructive examination (NDE) technique, that can locate degradation and measure its size and orientation, is an integral part of nuclear power plant maintenance. The steam generator inspection system is consisted of remotely controlled manipulator, testing instrument and software for data acquisition and analysis. Recently, the inspection systems have evolved to a much higher level of automation, efficiency and reliability resulting in a lower cost and shorter outage time. Electronic components have become smaller and deal with more complex algorithms. These systems are very fast, precise, reliable and easy to handle. The whole inspection, from the planning, examination, data analysis and final report, is now a highly automated process, which makes inspection much easier and more reliable. This paper presents the new generation of INETEC's VVER steam generator inspection system as ultimate solution for steam generator inspection and repair. (author)

  15. Material choices for the commercial fast reactor steam generators

    International Nuclear Information System (INIS)

    Willby, C.; Walters, J.

    1978-01-01

    Experience with fast reactor steam generators has shown them to be critical components in achieving a high availability. This paper presents the designers views on the use of ferritic materials for steam generators and describes the proposed design of the steam generators for the Commercial Fast Reactor (CFR), prototype of which are to be inserted in the Prototype Fast Reactor at Dounreay. (author)

  16. Steam generator development in France for the Super Phenix project

    International Nuclear Information System (INIS)

    Robin, M.G.

    1975-01-01

    'Steam Generator Development for Super Phenix Project'. The development program of steam generators studied by Fives-Cail Babcock and Stein Industrie Companies, jointly with CEA end EDF, for the Super Phenix 1200 MWe Fast Breeder Power Plant, is presented. The main characteristics of both sodium heated steam generators are emphasized and experimental studies related to their key features are reported. (author)

  17. Theoretical and experimental work on steam generator integrity and reliability with particular reference to leak development and detection. United Kingdom status report. October 1983

    International Nuclear Information System (INIS)

    Smedley, J.A.; Edge, D.M.

    1984-01-01

    This paper reviews the experimental and theoretical work in the UK on the characteristics of sodium-water reactions and describes work on the development of leak detection systems. A review of the operating experience with the PFR steam generators and the protection philosophy used on PFR is also given and the design studies for the Commercial Demonstration Fast Reactor (CDFR) are described

  18. Steam generator for pressurized-water reactors

    International Nuclear Information System (INIS)

    Michel, E.

    1971-01-01

    In the steam generator for a PWR the central fall space of a U-tube bundel heat exchanger is used as a preliminary cyclon separator. The steam escaping upwards, which is largely free of water, can flow through the residual heating surface, i.e. the U-tube turns. In this way substantial drying and less superheating by the heat still added becomes possible. In its upper part the central fall space for the water separated in the preliminary separator, enclosed by a cylindrical guide wall and the U-tube bundle, is provided with tangential inlet slots. Through these, the water-steam mixture steams out of the section of the vertical legs of the U-tube bundle into the fall space. Above the inlet slots the rising space is closed by means of a turn-round plate. At the lower end of the guide wall outlet, slots are provided for the water flowing downwards and radially outwards into the unfilled space. (DG/PB) [de

  19. Modelling the horizontal steam generator with APROS

    Energy Technology Data Exchange (ETDEWEB)

    Ylijoki, J. [VTT Energy, Espoo (Finland); Palsinajaervi, C.; Porkholm, K. [IVO International Ltd, Vantaa (Finland)

    1995-12-31

    In this paper the capability of the five- and six-equation models of the simulation code APROS to simulate the behaviour of the horizontal steam generator is discussed. Different nodalizations are used in the modelling and the results of the stationary state runs are compared. Exactly the same nodalizations have been created for the five- and six-equation models. The main simulation results studied in this paper are void fraction and mass flow distributions in the secondary side of the steam generator. It was found that quite a large number of simulation volumes is required to simulate the distributions with a reasonable accuracy. The simulation results of the different models are presented and their validity is discussed. (orig.). 4 refs.

  20. Three Steam Generator Replacement Projects in 1995

    International Nuclear Information System (INIS)

    Holz, R.; Clavier, G.

    1996-01-01

    Since the companies Siemens AG and Framatome S. A. joined their experience and efforts in the field of steam generator replacements and formed a consortium in 1991, the following projects were performed in 1995: Ringhals 3, Tihange 3 and Asco 1. Further projects will follow in 1996, i. e., Doel 4 and Asco 2. Currently, this European consortium is bidding for the contract to replace the steam generators at the Krsko NPP and hopes to be awarded in 1996. An overview of the way the Consortium Siemens and Framatome approaches SG replacement projects is given based on the projects performed in 1995. Various aspects of project planning, management, licensing, personnel qualification and techniques used on site will be discussed. (author)

  1. The steam generator programme of PISC III

    International Nuclear Information System (INIS)

    Birac, C.; Herkenrath, H.

    1990-12-01

    The PISC III Actions are intended to extend the results and methodologies of the previous PISC excercises, i.e. the validation of the capabilities of the various examination techniques when used on real defects in real components under real conditions of inspection. Being aware of the important safety role that steam generator tubes play as barrier between primary and secondary cooling system and of the industrial problems that the degradation of these tubes can create, the PISC III Management Board agreed to include in the PISC III Programme a special Action on Steam Generator Tubes Testing (SGT). It was decided to organize the programme in three phases, including Round Robin Tests (RRT): - capability tests on loose tubes, - capability tests on transportable mock-ups, - reliability tests on fixed mock-ups including some interesting SURRY tubes

  2. Modelling the horizontal steam generator with APROS

    Energy Technology Data Exchange (ETDEWEB)

    Ylijoki, J [VTT Energy, Espoo (Finland); Palsinajaervi, C; Porkholm, K [IVO International Ltd, Vantaa (Finland)

    1996-12-31

    In this paper the capability of the five- and six-equation models of the simulation code APROS to simulate the behaviour of the horizontal steam generator is discussed. Different nodalizations are used in the modelling and the results of the stationary state runs are compared. Exactly the same nodalizations have been created for the five- and six-equation models. The main simulation results studied in this paper are void fraction and mass flow distributions in the secondary side of the steam generator. It was found that quite a large number of simulation volumes is required to simulate the distributions with a reasonable accuracy. The simulation results of the different models are presented and their validity is discussed. (orig.). 4 refs.

  3. Acoustic leak detection of LMFBR steam generator

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Yoshida, Kazuo

    1993-01-01

    The development of a water leak detector with short response time for LMFBR steam generators is required to prevent the failure propagation caused by the sodium-water reaction and to maintain structural safety in steam generators. The development of an acoustic leak detector assuring short response time has attracted. The purpose of this paper is to confirm the basic detection feasibility of the active acoustic leak detector, and to investigate the leak detection method by erasing the background noise by spectrum analysis of the passive acoustic leak detector. From a comparison of the leak detection sensitivity of the active and the passive method, the active method is not influenced remarkably by the background noise, and it has possibility to detect microleakage with short response time. We anticipate a practical application of the active method in the future. (author)

  4. Update of operations with Westinghouse steam generators

    International Nuclear Information System (INIS)

    Malinowski, D.D.; Fletcher, W.D.

    1978-01-01

    Westinghouse commercial steam generators in operation now number 112, of which 98 are tubed with Inconel 600, the remainder with stainless steel. The implementation of all volatile treatment (AVT) was reported. It was noted that several plants had exhibited some tube corrosion during their initial periods using AVT; this observation indicated that the transition from phosphate chemistry control to AVT may have been subject to certain residual effects due to incomplete removal of phosphated deposits. As inspection results from steam generators operated on AVT became more generally available with the passage of time, a pattern of results emerged that seemed to correlate with the operating experience with phosphate chemistry control. Specifically, all the plants that experienced corrosion problems had from 1 to 8 yr of operational history using phosphates, while those with less than a year's experience using phosphates tended to be less affected by corrosion problems

  5. Advances in steam generator service technology

    International Nuclear Information System (INIS)

    Perez, Ric

    1998-01-01

    The most recent advances in pressurized water reactor steam generator service technology are discussed in this article. Focus is on new developments in robotics, including the Remotely Operated Service Arm (ROSA III); repair and maintenance services on the SG secondary side; and the newest advances in SG inspection. These products and services save utility costs, shorten outage durations, enhance plant performance and safety, and reduce radiation exposure. (author)

  6. Advances in steam generator service technology

    International Nuclear Information System (INIS)

    Nair, B. R.; Bastin, J. J.

    1997-01-01

    This paper will discuss the most recent and innovative advances in the areas of pressurized water reactor (PWR) steam generator service technology. The paper will include detail of new products such as the Remotely Operated Service Arm (ROSA-III), laser welded sleeving, and laser welded Direct Tube Repair (DTR) - products and services that save utility costs, shorten outage durations, enhance plant performance and safety, and reduce radiation exposure. (author)

  7. Materials engineering issues, LMFBR steam generators

    International Nuclear Information System (INIS)

    Spalaris, C.N.; Challenger, K.D.; Day, R.A.; Dutina, D.; Ring, P.J.

    1976-01-01

    Selection of 2-1/4 Cr-1 Mo as the reference construction material for LMFBR steam generators assumed a balance between its known intrinsic properties and our ability to accommodate certain of its deficiencies through design allowance. A comprehensive development program was undertaken to define base data needed, confirm assumptions made relative to desired performance, minimize defects by optimization of melting, fabrication and heat treatment processes, and prepare specifications for purchasing reactor components

  8. Fifth CNS international steam generator conference

    International Nuclear Information System (INIS)

    2006-01-01

    The Fifth CNS International Steam Generator Conference was held on November 26-29, 2006 in Toronto, Ontario, Canada. In contrast with other conferences which focus on specific aspects, this conference provided a wide ranging forum on nuclear steam generator technology from life-cycle management to inspection and maintenance, functional and structural performance characteristics to design architecture. The 5th conference has adopted the theme: 'Management of Real-Life Equipment Conditions and Solutions for the Future'. This theme is appropriate at a time of transition in the industry when plants are looking to optimize the performance of existing assets, prevent costly degradation and unavailability, while looking ahead for new steam generator investments in life-extension, replacements and new-build. More than 50 technical papers were presented in sessions that gave an insight to the scope: life management strategies; fouling, cleaning and chemistry; replacement strategies and new build design; materials degradation; condition assessment/fitness for service; inspection advancements and experience; and thermal hydraulic performance

  9. Data analysis for steam generator tubing samples

    International Nuclear Information System (INIS)

    Dodd, C.V.

    1996-07-01

    The objective of the Improved Eddy-Current ISI for Steam Generators program is to upgrade and validate eddy-current inspections, including probes, instrumentation, and data processing techniques for inservice inspection of new, used, and repaired steam generator tubes; to improve defect detection, classification and characterization as affected by diameter and thickness variations, denting, probe wobble, tube sheet, tube supports, copper and sludge deposits, even when defect types and other variables occur in combination; to transfer this advanced technology to NRC's mobile NDE laboratory and staff. This report provides a description of the application of advanced eddy-current neural network analysis methods for the detection and evaluation of common steam generator tubing flaws including axial and circumferential outer-diameter stress-corrosion cracking and intergranular attack. The report describes the training of the neural networks on tubing samples with known defects and the subsequent evaluation results for unknown samples. Evaluations were done in the presence of artifacts. Computer programs are given in the appendix

  10. Operating experience of steam generator test facility

    International Nuclear Information System (INIS)

    Sureshkumar, V.A.; Madhusoodhanan, G.; Noushad, I.B.; Ellappan, T.R.; Nashine, B.K.; Sylvia, J.I.; Rajan, K.K.; Kalyanasundaram, P.; Vaidyanathan, G.

    2006-01-01

    Steam Generator (SG) is the vital component of a Fast Reactor. It houses both water at high pressure and sodium at low pressure separated by a tube wall. Any damage to this barrier initiates sodium water reaction that could badly affect the plant availability. Steam Generator Test Facility (SGTF) has been set up in Indira Gandhi Centre for Atomic Research (IGCAR) to test sodium heated once through steam generator of 19 tubes similar to the PFBR SG dimension and operating conditions. The facility is also planned as a test bed to assess improved designs of the auxiliary equipments used in Fast Breeder Reactors (FBR). The maximum power of the facility is 5.7 MWt. This rating is arrived at based on techno economic consideration. This paper covers the performance of various equipments in the system such as Electro magnetic pumps, Centrifugal sodium pump, in-sodium hydrogen meters, immersion heaters, and instrumentation and control systems. Experience in the system operation, minor modifications, overall safety performance, and highlights of the experiments carried out etc. are also brought out. (author)

  11. Materials performance in operating PWR steam generators

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    The Inconel-600 tubing in operating PWR steam generators has developed leaks due to intergranular stress corrosion cracking or a general wastage attack, originating from the secondary side of the tubing. Corrosion has been limited to those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate. Wastage or pitting attack has always been associated with local concentration of sodium hydrogen phosphates, whereas stress corrosion has been associated with local concentration of sodium or potassium hydroxides. The only instance of stress corrosion originating from the primary side occurred on cold-worked tubing when hydrogen was not added to getter oxygen, and LiOH was not added to raise the pH of the primary coolant. All PWR manufacturers are now recommending that the phosphate treatment of the secondary coolant be abandoned in favor of an all-volatile treatment. Experience in operating plants has shown, however, that removal of phosphate-rich sludge deposits is difficult, and that further wastage and/or intergranular stress corrosion may develop; the residual sodium phosphates gradually convert by reaction with corrosion product hydroxides to sodium hydroxide, which remains concentrated in the limited flow areas. Improvements in circulation patterns have been achieved by inserting flow baffles in some PWR steam generators. Inservice monitoring by eddy current techniques is useful for detecting corrosion-induced defects in the tubing, but irreproducibility in field examinations can lead to uncertainties interpreting the results. (U.S.)

  12. Development of the visual inspection system for the top of the tube sheet in steam generators

    International Nuclear Information System (INIS)

    Kim, Gyung Sub; Choi, Sang Hoon; Kim, Ki Chul

    2008-01-01

    Steam Generators at Nuclear Power plants have a important function to isolate Radioactivity between the primary side radioactive fluid running through tubes and the secondary side with non-radioactive fluid through out of a tube bundle, in addition to a function of steam generation. Therefore, To obtain integrity of Steam Generator is really important for safety in the nuclear power plant. At the same time, sludge and foreign objects in steam generators are known as major sources causing the damage of SG tubes. But there is no way to prevent those coming to steam generators until now. Therefore, a periodic inspection and removal of those in steam generators is the only way for those Generally, Most of the Nuclear Power Plants have been inspecting visually every outage for the top of the tube sheet in which sludge and foreign objects lead to the buildup to know how these are

  13. Corrosion and Rupture of Steam Generator Tubings in PWRs

    International Nuclear Information System (INIS)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-01

    This report is intended to provide corrosion engineers in the filed of nuclear energy with information on the corrosion and rupture behavior of steam generator tubing in PWRs. Various types of corrosion in PWR steam generator tubing have been reported all around the world, and countermeasures such as the addition of corrosion inhibitors, a water chemistry control, a tube plugging and sleeving have been applied. Steam generators equipped with alloy 600 tubing, which are not so resistant to a stress corrosion cracking (SCC), have generally been replaced with new steam generators made of alloy 690 TT (Thermally treated). Pull tube examination results which were performed of KAERI are summarized. The tubes were affected by a pitting, SCC, and a denting. Nondestructive examination method for the tubes and repair techniques are also reviewed. In addition, the regulatory guidance of some countries are reviewed. As a part of a tube integrity project in Korea, some results on a tube rupture and leak behaviors for axial cracks are also mentioned

  14. Corrosion and Rupture of Steam Generator Tubings in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-15

    This report is intended to provide corrosion engineers in the filed of nuclear energy with information on the corrosion and rupture behavior of steam generator tubing in PWRs. Various types of corrosion in PWR steam generator tubing have been reported all around the world, and countermeasures such as the addition of corrosion inhibitors, a water chemistry control, a tube plugging and sleeving have been applied. Steam generators equipped with alloy 600 tubing, which are not so resistant to a stress corrosion cracking (SCC), have generally been replaced with new steam generators made of alloy 690 TT (Thermally treated). Pull tube examination results which were performed of KAERI are summarized. The tubes were affected by a pitting, SCC, and a denting. Nondestructive examination method for the tubes and repair techniques are also reviewed. In addition, the regulatory guidance of some countries are reviewed. As a part of a tube integrity project in Korea, some results on a tube rupture and leak behaviors for axial cracks are also mentioned.

  15. Minimize corrosion degradation of steam generator tube materials

    International Nuclear Information System (INIS)

    Lu, Y.

    2006-01-01

    As part of a coordinated program, AECL is developing a set of tools to aid with the prediction and management of steam generator performance. Although stress corrosion cracking (of Alloy 800) has not been detected in any operating steam generator, for life management it is necessary to develop mechanistic models to predict the conditions under which stress corrosion cracking is plausible. Experimental data suggest that all steam generator tube materials are susceptible to corrosion degradation under some specific off-specification conditions. The tolerance to the chemistry upset for each steam generator tube alloy is different. Electrochemical corrosion behaviors of major steam generator tube alloys were studied under the plausible aggressive crevice chemistry conditions. The potential hazardous conditions leading to steam generator tube degradation and the conditions, which can minimize steam generator tube degradation have been determined. Recommended electrochemical corrosion potential/pH zones were defined for all major steam generator tube materials, including Alloys 600, 800, 690 and 400, under CANDU steam generator operating and startup conditions. Stress corrosion cracking tests and accelerated corrosion tests were carried out to verify and revise the recommended electrochemical corrosion potential/pH zones. Based on this information, utilities can prevent steam generator material degradation surprises by appropriate steam generator water chemistry management and increase the reliability of nuclear power generating stations. (author)

  16. Friction pressure drop and heat transfer coefficient of two-phase flow in helically coiled tube once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Nariai, Hideki; Kobayashi, Michiyuki; Matsuoka, Takeshi.

    1982-01-01

    Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report. (author)

  17. 1984 NRC Branch annual report: Steam Generator Group Project

    International Nuclear Information System (INIS)

    Clark, R.A.; Bickford, R.L.; Birks, A.S.

    1985-02-01

    Program objectives are to develop validated models, based on experimental data, for prediction of margins-to-failure under burst and collapse pressures of steam generator tubing found to be service-degraded by eddy current inservice inspection. This required the research contractor, Pacific Northwest Laboratory, to obtain, house and prepare an out-of-service degraded steam generator for a confirmatory research program. This research includes: (1) validation of current and developmental nondestructive examination techniques; (2) optimization of inservice inspection procedures, sampling plan and inspection period; (3) validation of tube integrity predictive models; (4) validation of stress corrosion cracking predictive models; (5) development/optimization of tube plugging criteria; and (6) evaluation of proposed chemical cleaning and decontamination processes/procedures with respect to near-term integrity and long-term effects on corrosion, degradation and safety

  18. Tube sheet design for PFBR steam generator

    International Nuclear Information System (INIS)

    Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

    1991-01-01

    Top and bottom tube sheets of PFBR Steam Generators have been analysed with 3D and axisymmetric models using CASTEM Programs. Analysis indicates that the effects of piping reactions at the inlet/outlet nozzles on the primary stresses in the tube sheets are negligible and the asymmetricity of the deformation pattern introduced in the tube sheet by the presence of inlet/outlet and manhole nozzles is insignificant. The minimum tube sheet thicknesses for evaporator and reheater are 135 mm and 75 mm respectively. Further analysis has indicated the minimum fillet radius at the junction of tube sheet and dished end should be 20 mm. Simplified methodology has been developed to arrive at the number of thermal baffles required to protect the tube sheet against fatigue damage due to thermal transient. This method has been applied to PFBR steam generators to determine the required number of thermal baffles. For protecting the bottom tube sheet of evaporator against the thermal shock due to feed water and secondary pump trip, one thermal shield is found to be sufficient. Further analysis is required to decide upon the actual number to take care of the severe thermal transient, following the event of sudden dumping of water/steam, immediately after the sodium-water reaction. (author)

  19. Steam generator of FBR type reactor

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1992-01-01

    Liquid metal (for example, mercury) which is scarcely reactive with metal sodium is contained and cover gases which are scarcely reactive with the liquid metal are filled in a steam generator of an FBR type reactor and it is closed. The heat of primary sodium is transferred to the liquid metal, which is not reactive with sodium, in a primary thermal conduction portion. Since the temperature of the primary thermal conduction portion is high, the density is extremely low. On the other hand, since a second thermal conduction portion is kept at a single phase and the temperature is lower compared with that of the first thermal conduction portion, the density is kept high. since the density difference and gas jetting speed generate a great circulating force to liquid metal passing the opening of a partition plate, heat can be conducted on the side of water without disposing pumps. The steam concentration in the liquid metal is low being in a single phase of steams, corrosion caused from the outside of pipes of the primary thermal conduction pipe is scarcely promoted. Even if sodium leaks should be caused, since the sodium concentration in the liquid metal is extremely low and the reactivity is low, the temperature of the liquid metal is not elevated. (N.H.)

  20. Ferritic steels for French LMFBR steam generators

    International Nuclear Information System (INIS)

    Aubert, M.; Mathieu, B.; Petrequin, P.

    1983-06-01

    Austenitic stainless steels have been widely used in many components of the French LMFBR. Up to now, ferritic steels have not been considered for these components, mainly due to their relatively low creep properties. Some ferritic steels are usable when the maximum temperatures in service do not exceed about 530 0 C. It is the case of the steam generators of the Phenix plant, where the exchange tubes of the evaporator are made of 2,25% Cr-1% Mo steel, stabilized or not by addition of niobium. These ferritic alloys have worked successfully since the first steam production in October 1973. For the SuperPhenix power plant, an ''all austenitic stainless alloy'' apparatus has been chosen. However, for the future, ferritic alloys offer potential for use as alternative materials in the evaporators: low alloys steels type 2,25% Cr-1% Mo (exchange tubes, tube-sheets, shells), or at higher chromium content type 9% Cr-2% Mo NbV (exchange tubes) or 12M Cr-1% Mo-V (tube-sheets). Most of these steels have already an industrial background, and are widely used in similar applications. The various potential applications of these steels are reviewed with regards to the French LMFBR steam generators, indicating that some points need an effort of clarification, for instance the properties of the heterogeneous ferritic/austenitic weldments

  1. Diagnostic system of steam generator, especially molten metal heated steam generator

    International Nuclear Information System (INIS)

    Matal, O.; Martoch, J.

    1986-01-01

    A diagnostic system is described and graphically represented consisting of a leak detector, a medium analyzer and sensors placed on the piping connected to the indication sections of both tube plates. The advantage of the designed system consists in the possibility of detecting tube failure immediately on leak formation, especially in generators with duplex tubes. This shortens the period of steam generator shutdown for repair and reduces power losses. The design also allows to make periodical leak tests during planned steam generator shutdowns. (A.K.)

  2. LMFBR steam generator development: duplex bayonet tube steam generator. Volume II

    International Nuclear Information System (INIS)

    DeFur, D.D.

    1975-04-01

    This report represents the culmination of work performed in fulfillment of ERDA Contract AT(11-1)-2426, Task Agreement 2, in which alternate steam generator designs were developed and studied. The basic bayonet tube generator design previously developed by C-E under AEC Contract AT(11-1)-3031 was expanded by incorporating duplex heat transfer tubes to enhance the unit's overall safety and reliability. The effort consisted of providing and evaluating conceptual designs of the evaporator, superheater and reheater components for a large plant LMFBR steam generator (950 MWt per heat transport loop)

  3. Analysis of fast reactor steam generator performance

    International Nuclear Information System (INIS)

    Hulme, G.; Curzon, A.F.

    1992-01-01

    A computer model for the prediction of flow and temperature fields within a fast reactor steam generator unit is described. The model combines a commercially available computational fluid dynamics (CFD) solver (PHOENICS) with a steam-tube calculation and provides solutions for the fully coupled flow and temperature fields on both the shell side and the tube side. The model includes the inlet and outlet headers and the bottom end stagnant zone. It also accounts for the effects of support grids and edge-gaps. Two and three dimensional and transient calculations have been performed for both straight tube and J-tube units. Examples of the application of the model are presented. (7 figures) (Author)

  4. Expandable antivibration bar for a steam generator

    International Nuclear Information System (INIS)

    Lagally, H.O.

    1987-01-01

    This patent describes a steam generator for a nuclear power plant comprising a shell, a plurality of tubes having a U-shaped configuration arranged in successive columns within the shell. The tubes are adapted to heat feedwater flowing around the outside of the tubes by the flow of hot reactor coolant within the tubes, and antivibration bars any vibrations of the tubes as a result of steam between the columns of tubes. The improvement described here comprises means for varying the thickness of the antivibration bars to fit substantially the actual space between the columns of tubes comprising first and second bars, with at least one bar being movable, and with at least one mating inclined surface between the first and second bars

  5. Integration of torrefaction with steam power plant

    Energy Technology Data Exchange (ETDEWEB)

    Zakri, B.; Saari, J.; Sermyagina, E.; Vakkilainen, E.

    2013-09-01

    Torrefaction is one of the pretreatment technologies to enhance the fuel characteristics of biomass. The efficient and continuous operation of a torrefaction reactor, in the commercial scale, demands a secure biomass supply, in addition to adequate source of heat. Biorefinery plants or biomass-fuelled steam power plants have the potential to integrate with the torrefaction reactor to exchange heat and mass, using available infrastructure and energy sources. The technical feasibility of this integration is examined in this study. A new model for the torrefaction process is introduced and verified by the available experimental data. The torrefaction model is then integrated in different steam power plants to simulate possible mass and energy exchange between the reactor and the plants. The performance of the integrated plant is investigated for different configurations and the results are compared. (orig.)

  6. Chemical control and design considerations for CANDU-PHW steam generators

    International Nuclear Information System (INIS)

    Frost, C.R.; Churchill, B.R.

    1978-01-01

    Ontario Hydro presently operates eight nuclear power units with a total capacitiy of about 4000 MW(e) net. Operating experience has been with Monel-400 and with Inconel-600 tubed steam generators using sodium phosphate or all volatile control of the boiler steam and water system. With a heavy water Heat Transport System, steam generator tube integrity is an essential ingredient of economical power production. Only three steam generator tube failures have occurred so far in about 40 unit-years operation. None was attributable to corrosion. Factors in the good reliability are, careful engineering design, good quality control at all stages of tubing and steam generator manufacture and close chemical control. The continuing evolution of our steam generator design means that future requirements will be more stringent. (author)

  7. Report on US-Japan 1983 meetings on steam generators

    International Nuclear Information System (INIS)

    1984-04-01

    This is a report on a trip to Japan by personnel of the US Nuclear Regulatory Commission in 1983 to exchange information on steam generators of nuclear power plants. Steam generators of Japanese pressurized water reactors have experienced nearly all of the forms of degradation that have been experienced in US recirculating-type steam generators, except for denting and pitting. More tubes have been plugged per year of reactor operation in Japanese than in US steam generators, but much of the Japanese tube plugging is preventative rather than the result of leaks experienced. The number of leaks per reactor year is much smaller for Japanese than for US steam generators. No steam generators have been replaced in Japan while several have been replaced in the US. The Japanese experience may be related to their very stringent inspection and maintenance programs for steam generators

  8. Structural integrity analysis of a steam turbine

    International Nuclear Information System (INIS)

    Villagarcia, Maria P.

    1997-01-01

    One of the most critical components of a power utility is the rotor of the steam turbine. Catastrophic failures of the last decades have promoted the development of life assessment procedures for rotors. The present study requires the knowledge of operating conditions, component geometry, the properties of materials, history of the component, size, location and nature of the existing flaws. The aim of the present work is the obtention of a structural integrity analysis procedure for a steam turbine rotor, taking into account the above-mentioned parameters. In this procedure, a stress thermal analysis by finite elements is performed initially, in order to obtain the temperature and stress distribution for a subsequent analysis by fracture mechanics. The risk of a fast fracture due to flaws in the central zone of the rotor is analyzed. The procedure is applied to an operating turbine: the main steam turbine of the Atucha I nuclear power utility. (author)

  9. Moisture separator for steam generator level measurement system

    International Nuclear Information System (INIS)

    Cantineau, B.J.

    1987-01-01

    A steam generator level measurement system having a reference leg which is kept full of water by a condensation pot, has a liquid/steam separator in the connecting line between the condensation pot and the steam phase in the steam generator to remove excess liquid from the steam externally of the steam generator. This ensures that the connecting line does not become blocked. The separator pot has an expansion chamber which slows down the velocity of the steam/liquid mixture to aid in separation, and a baffle, to avoid liquid flow into the line connected to the condensate pot. Liquid separated is returned to the steam generator below the water level through a drain line. (author)

  10. 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

  11. Structural considerations in steam generator replacement

    International Nuclear Information System (INIS)

    Bertheau, S.R.; Gazda, P.A.

    1991-01-01

    Corrosion of the tubes and tube-support structures inside pressurized water reactor (PWR) steam generators has led many utilities to consider a replacement of the generators. Such a project is a major undertaking for a utility and must be well planned to ensure an efficient and cost-effective effort. This paper discusses various structural aspects of replacement options, such as total or partial generator replacement, along with their associated pipe cuts; major structural aspects associated with removal paths through the equipment hatch or through an opening in the containment wall, along with the related removal processes; onsite movement and storage of the generators; and the advantages and disadvantages of the removal alternatives. This paper addresses the major structural considerations associated with a steam generator replacement project. Other important considerations (e.g., licensing, radiological concerns, electrical requirements, facilities for management and onsite administrative activities, storage and fabrication activities, and offsite transportation) are not discussed in this paper, but should be carefully considered when undertaking a replacement project

  12. Design features of Advanced Power Reactor (APR) 1400 steam generator

    International Nuclear Information System (INIS)

    Park, Tae-Jung; Park, Jun-Soo; Kim, Moo-Yong

    2004-01-01

    Advanced Power Reactor 1400 (APR 1400) which is to achieve the improvement of the safety and economical efficiency has been developed by Korea Hydro and Nuclear Power Co., Ltd. (KHNP) with the support from industries and research institutes. The steam generator for APR 1400 is an evolutionary type from System 80 + , which is the recirculating U-tube heat exchanger with integral economizer. Compared to the System 80 + steam generator, it is focused on the improved design features, operating and design conditions of APR 1400 steam generator. Especially, from the operation experience of Korean Standard Nuclear Power Plant (KSNP) steam generator, the lessons-learned measures are incorporated to prevent the tube wear caused by flow-induced vibration (FIV). The concepts for the preventive design features against FIV are categorized to two fields; flow distribution and dynamic response characteristics. From the standpoint of flow distribution characteristics, the egg-crate flow distribution plate (EFDP) is installed to prevent the local excessive flow loaded on the most susceptible tube to wear. The parametric study is performed to select the optimum design with the efficient mitigation of local excessive flow. ATHOS3 Mod-01 is used and partly modified to analyze the flow field of the APR 1400 steam generator. In addition, the upper tube bundle support is designed to eliminate the presence of tube with a low natural frequency. Based on the improved upper tube bundle support, the modal analysis is performed and compared with that of System 80 + . Using the results of flow distribution and modal analysis, the two mechanisms of flow-induced vibration are investigated; fluid-elastic instability (FEI) and random turbulence excitation (RTE). (authors)

  13. Emergency makeup of nuclear steam generators in blackout conditions

    International Nuclear Information System (INIS)

    Korolev, A.V.; Derevyanko, O.V.

    2014-01-01

    The paper describes an original solution for using steam energy to organize makeup of NPP steam generators in blackout conditions. The proposed solution combines a disk friction turbine and an axial turbine in a single housing to provide a high overall technical effect enabling the replenishment of nuclear steam generators with steam using the pump turbine drive assembly. The application of the design is analyzed and its efficiency and feasibility are shown

  14. LMFBR steam generators in the United Kingdom

    International Nuclear Information System (INIS)

    Anderson, R.; Hayden, O.

    2002-01-01

    Experience has been gained in the UK on the operation of LMFBR Steam Generator Units (SGU) over a period of 20 years from the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR). The DFR steam generator featured a double barrier and therefore did not represent a commercial design. PFR, however, faced the challenge of a single wall design and it is experience from this which is most valuable. The PFR reactor went critical in March 1974 and the plant operating history since then has been dominated by experience with leaks in the tube to tube plate welds of the high performance U-tubes SGU's. Operation at high power using the full complement of three secondary sodium circuits was delayed until July 1976 by the occurrence of leaks in the tube to tube plate welds of the superheater and reheater units which are fabricated in stainless steel. Repairs were carried out to the two superheaters and they were returned to service. The reheater tube bundle was removed from circuit after sodium was found to have entered the steam side. When the sodium had been removed and inspection carried out it was decided not to recover the unit. Since 1976 the remaining five stainless steel units have operated satisfactorily. This year a replacement reheater unit has been installed. This is of a new design in 9-Cr-Mo ferritic steel using a sleeve through which the steam tube passes to eliminate the tube to tube plate weld. Despite a few early leaks in evaporator tube to tube plate welds up to 1979, these failures did not initially present a major problem. However, in 1980 the rate of evaporator weld failures increased and despite the successful application of a shot peening process to eliminate stress corrosion failures from the water side of the weld, failures traced to the sodium side continued. A sleeving process was developed for application to complete evaporator units on a production basis with the objective of bypassing the welds at each end of the 500 tubes. The decision

  15. Steam generator for PWR type reactor

    International Nuclear Information System (INIS)

    Baba, Iwao; Hiyama, Nobuyuki.

    1994-01-01

    A steam generator of the present invention comprises a primary coolant chamber having primary coolants circulating therein, a secondary coolants chamber having secondary coolants and steams circulating therein, which are isolated from each other by a partition wall, and heat pipes disposed being passed through the partition wall. The heat pipes are disposed having an evaporation portion in the primary coolants chamber, a condensation portion in the secondary coolants chamber, and an intermediate heat insulating portion in the partition wall. Since the primary coolants containing radioactivity and the secondary coolants not containing radioactivity does not transfer heat directly by a heat transfer wall, a leakage accident of radioactivity to the secondary coolants can be prevented. Moreover, since the heat pipes are used, a great amount of heat can be transferred by a slight temperature difference by using steams of the heat transfer medium itself, latent heat due to coagulation, and capillary phenomenon. Since neither transferring power nor pumps are required, heat of the primary coolants can effectively be transferred to the secondary coolants. (N.H.)

  16. Integrity assessment of the ferritic / austenitic dissimilar weld joint between intermediate heat exchanger and steam generator in fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayakumar, T.; Laha, K.; Chandravathi, K. S.; Parameswaran, P.; Goyal, S.; Kumar, J. G.; Mathew, M. D. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam- 603 102 (India)

    2012-07-01

    Integrity of the modified 9Cr-1Mo / alloy 800 dissimilar joint welded with Inconel 182 electrodes has been assessed under creep condition based on the detailed analysis of microstructure and stress distribution across the joint by finite element analysis. A hardness peak at the ferritic / austenitic weld interface and a hardness trough at the inter-critical heat affected zone (HAZ) in ferritic base metal developed. Un-tempered martensite was found at the ferritic / austenitic weld interface to impart high hardness in it; whereas annealing of martensitic structure of modified 9Cr-1Mo steel by inter-critical heating during welding thermal cycle resulted in hardness tough in the inter-critical HAZ. Creep tests were carried out on the joint and ferritic steel base metal at 823 K over a stress range of 160-320 MPa. The joint possessed lower creep rupture strength than its ferritic steel base metal. Failure of the joint at relatively lower stresses occurred at the ferritic / austenitic weld interface; whereas it occurred at inter-critical region of HAZ at moderate stresses. Cavity nucleation associated with the weld interface particles led to premature failure of the joint. Finite element analysis of stress distribution across the weld joint considering the micro-mechanical strength inhomogeneity across it revealed higher von-Mises and principal stresses at the weld interface. These stresses induced preferential creep cavitation at the weld interface. Role of precipitate in enhancing creep cavitation at the weld interface has been elucidated based on the FE analysis of stress distribution across it. (authors)

  17. Effects of Chemistry Parameters of Primary Water affecting Leakage of Steam Generator Tube Cracks

    Energy Technology Data Exchange (ETDEWEB)

    Shin, D. M.; Cho, N. C.; Kang, Y. S.; Lee, K. H. [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    Degradation of steam generator (SG) tubes can affect pressure boundary tightness. As a defense-in-depth measure, primary to secondary leak monitoring program for steam generators is implemented, and operation is allowed under leakage limits in nuclear power plants. Chemistry parameters that affect steam generator tube leakage due to primary water stress corrosion cracking (PWSCC) are investigated in this study. Tube sleeves were installed to inhibit leakage and improve tube integrity as a part of maintenance methods. Steam generators occurred small leak during operation have been replaced with new steam generators according to plant maintenance strategies. The correlations between steam generator leakage and chemistry parameters are presented. Effects of primary water chemistry parameters on leakage from tube cracks were investigated for the steam generators experiencing small leak. Unit A experienced small leakage from steam generator tubes in the end of operation cycle. It was concluded that increased solubility of oxides due to high pHT could make leakage paths, and low boron concentration lead to less blockage in cracks. Increased dissolved hydrogen may retard crack propagations, but it did not reduce leak rate of the leaking steam generator. In order to inhibit and reduce leakage, pH{sub T} was controlled by servicing cation bed operation. The test results of decreasing pHT indicate low pHT can reduce leak rate of PWSCC cracks in the end of cycle.

  18. Effects of Chemistry Parameters of Primary Water affecting Leakage of Steam Generator Tube Cracks

    International Nuclear Information System (INIS)

    Shin, D. M.; Cho, N. C.; Kang, Y. S.; Lee, K. H.

    2016-01-01

    Degradation of steam generator (SG) tubes can affect pressure boundary tightness. As a defense-in-depth measure, primary to secondary leak monitoring program for steam generators is implemented, and operation is allowed under leakage limits in nuclear power plants. Chemistry parameters that affect steam generator tube leakage due to primary water stress corrosion cracking (PWSCC) are investigated in this study. Tube sleeves were installed to inhibit leakage and improve tube integrity as a part of maintenance methods. Steam generators occurred small leak during operation have been replaced with new steam generators according to plant maintenance strategies. The correlations between steam generator leakage and chemistry parameters are presented. Effects of primary water chemistry parameters on leakage from tube cracks were investigated for the steam generators experiencing small leak. Unit A experienced small leakage from steam generator tubes in the end of operation cycle. It was concluded that increased solubility of oxides due to high pHT could make leakage paths, and low boron concentration lead to less blockage in cracks. Increased dissolved hydrogen may retard crack propagations, but it did not reduce leak rate of the leaking steam generator. In order to inhibit and reduce leakage, pH_T was controlled by servicing cation bed operation. The test results of decreasing pHT indicate low pHT can reduce leak rate of PWSCC cracks in the end of cycle

  19. Steam generator replacement from ALARA aspects

    International Nuclear Information System (INIS)

    Terry, I.; Breznik, B.

    2003-01-01

    This paper is going to consider radiological related parameters important for steam generator replacement (SGR) implementation. These parameters are identified as ALARA related parameters, owner-contractor relationship, planning, health physics with logistic services, and time required for the replacement. ALARA related parameters such as source or initial dose rate and plant system configuration define the initial conditions for the planning. There is room to optimise work planning. managerial procedures and also the staff during the implementation phase. The overview of these general considerations is based on the following background: using internationally available data and the experience of one of the vendors, i.e. Siemens-Framatome, and management experience of SG replacement which took place at Krsko NPP in the spring of 2000. Generally plant decisions on maintenance or repair procedures under radiation conditions take into account ALARA considerations. But in the main it is difficult to adjudge the results of an ALARA study, usually in the form of a collective dose estimate, because a comparison standard is missing. That is, very often the planned work is of a one-off nature so comparisons are not possible or the scopes are not the same. In such a case the collective doses for other types of work are looked at and a qualitative evaluation is made. In the case of steam generator replacement this is not the case. Over years of steam generator replacements world-wide a standard has been developed gradually. The first part of the following displays an overview of SGR and sets the Krsko SGR in perspective by applying dose analysis. The second part concentrates on the Krsko SGR itself and its ALARA aspects. (authors)

  20. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  1. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  2. Fatigue cracking on a steam generator tube

    International Nuclear Information System (INIS)

    Boccanfuso, M.; Lothios, J.; Thebault, Y.; Bruyere, B.; Duisabeau, L.; Herms, E.

    2015-01-01

    A circumferential fatigue crack was observed on a steam generator tube of the unit 2 of the Fessenheim plant. The results of destructive testing and the examination of the fracture surface show that the circumferential crack is linked to a large number of cycles with a very low stress intensity factor. Other aggravating factors like inter-granular corrosion have played a role in the initiating phase of fatigue cracking. The damage has been exacerbated by the lack of support of the tube at the level of the anti-vibration bars. (A.C.)

  3. Steam generator replacement project in 2000

    International Nuclear Information System (INIS)

    Cerjak, J.; Holz, R.; Haus, J.; Gloaguen, C.

    1999-01-01

    NE Krsko has awarded the contract for the Steam Generator Replacement Project, which is one of the modernization projects in Krsko, to the Consortium of Siemens / Framatome in February 1998. This paper deals with the various aspects of the project: scope planning, engineering, preparation of modification packages for licensing, management, major techniques used, etc., showing also the status of the activities for the project which are scheduled to be performed in April through June 2000. The project is being performed on a turnkey basis, that means the Consortium is performing all engineering, preparation of the modification packages and site activities; NE Krsko is dealing with the licensing of the project.(author)

  4. Steam generator of the forced circulation type

    International Nuclear Information System (INIS)

    Forestier, Jean; Leblanc, Bernard; Monteil, Marcel; Monteil, Pierre

    1977-01-01

    The steam generator described is of the forced circulation single passage type comprising an outer casing including a vertical generally cylindrical side ring, an internal skirt coaxial with the outer casing, the bottom of this skirt having a free edge separated from a bottom end closing the outer casing, a central tube plate extending horizontally near a top end, in opposition to the bottom end, a peripheral tube plate, parallel to the central plate and located in the annular space under this central plate, a bundle of J shaped tubes [fr

  5. Perspective of the Westinghouse steam generator secondary side maintenance approach

    Energy Technology Data Exchange (ETDEWEB)

    Ramaley, D. [Westinghouse Electric Company LLC, Cranberry Township, Pennsylvania (United States)

    2012-07-01

    Historically, Westinghouse had developed a set of steam generator secondary maintenance guidelines focused around performing recurring activities each outage without direct regards to the age, deposit loading, operational status, or corrosion status of the steam generator. Through the evolution of steam generator design and steam generator condition data, Westinghouse now uses a proactive assessment and planning approach for utilities. Westinghouse works with utilities to develop steam generator secondary maintenance plans for long term steam generator viability. Westinghouse has developed a portfolio of products to allow utilities to optimize steam generator operability and develop programs aimed at maintaining the steam generator secondary side in a favorable condition for successful long term operation. Judicious use of the means available for program development should allow for corrosion free operation, long term full power operation at optimum thermal efficiency, and leveling of outage expenditures over a long period of time. This paper will review the following required elements for an effective steam generator secondary side strategy: • Assessment: In order to develop an appropriate maintenance strategy, actions must be taken to obtain an accurate picture of the SG secondary side condition. • Forecasting: Using available data predictions are developed for future steam generator conditions and required maintenance actions. • Action: Cost effective engineering and maintenance actions must be completed at the appropriate time as designated by the plan. • Evaluation of Results: Following execution of maintenance tactics, it is necessary to revise strategy and develop technology enhancements as appropriate. (author)

  6. Safety significance of steam generator tube degradation mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Roussel, G; Mignot, P [AIB-Vincotte Nuclear - AVN, Brussels (Belgium)

    1991-07-01

    Steam generator (SG) tube bundle is a part of the Reactor Coolant Pressure Boundary (RCPB): this means that its integrity must be maintained. However, operating experience shows various types of tube degradation to occur in the SG tubing, which may lead to SG tube leaks or SG tube ruptures and create a loss of primary system coolant through the SG, therefore providing a direct path to the environment outside the primary containment structure. In this paper, the major types of known SG tube degradations are described and analyzed in order to assess their safety significance with regard to SG tube integrity. In conclusion: The operational reliability and the safety of the PWR steam generator s requires a sufficient knowledge of the degradation mechanisms to determine the amount of degradation that a tube can withstand and the time that it may remain in operation. They also require the availability of inspection techniques to accurately detect and characterize the various degradations. The status of understanding of the major types of degradation summarized in this paper shows and justifies why efforts are being performed to improve the management of the steam generator tube defects.

  7. Hideout in steam generator tube deposits

    International Nuclear Information System (INIS)

    Balakrishnan, P.V.; Franklin, K.J.; Turner, C.W.

    1998-05-01

    Hideout in deposits on steam generator tubes was studied using tubes coated with magnetite. Hideout from sodium chloride solutions at 279 degrees C was followed using an on-line high-temperature conductivity probe, as well as by chemical analysis of solution samples from the autoclave in which the studies were done. Significant hideout was observed only at a heat flux greater than 200 kW/m 2 , corresponding to a temperature drop greater than 2 degrees C across the deposits. The concentration factor resulting from the hideout increased highly non-linearly with the heat flux (varying as high as the fourth power of the heat flux). The decrease in the apparent concentration factor with increasing deposit thickness suggested that the pores in the deposit were occupied by a mixture of steam and water, which is consistent with the conclusion from the thermal conductivity measurements on deposits in a separate study. Analyses of the deposits after the hideout tests showed no evidence of any hidden-out solute species, probably due to the concentrations being very near the detection limits and to their escape from the deposit as the tests were being ended. This study showed that hideout in deposits may concentrate solutes in the steam generator bulk water by a factor as high as 2 x 10 3 . Corrosion was evident under the deposit in some tests, with some chromium enrichment on the surface of the tube. Chromium enrichment usually indicates an acidic environment, but the mobility required of chromium to become incorporated into the thick magnetite deposit may indicate corrosion under an alkaline environment. An alkaline environment could result from preferential accumulation of sodium in the solution in the deposit during the hideout process. (author)

  8. Mathematical modeling of control system for the experimental steam generator

    Science.gov (United States)

    Podlasek, Szymon; Lalik, Krzysztof; Filipowicz, Mariusz; Sornek, Krzysztof; Kupski, Robert; Raś, Anita

    2016-03-01

    A steam generator is an essential unit of each cogeneration system using steam machines. Currently one of the cheapest ways of the steam generation can be application of old steam generators came from army surplus store. They have relatively simple construction and in case of not so exploited units - quite good general conditions, and functionality of mechanical components. By contrast, electrical components and control systems (mostly based on relay automatics) are definitely obsolete. It is not possible to use such units with cooperation of steam bus or with steam engines. In particular, there is no possibility for automatically adjustment of the pressure and the temperature of the generated steam supplying steam engines. Such adjustment is necessary in case of variation of a generator load. The paper is devoted to description of improvement of an exemplary unit together with construction of the measurement-control system based on a PLC. The aim was to enable for communication between the steam generator and controllers of the steam bus and steam engines in order to construction of a complete, fully autonomic and maintenance-free microcogeneration system.

  9. Mathematical modeling of control system for the experimental steam generator

    Directory of Open Access Journals (Sweden)

    Podlasek Szymon

    2016-01-01

    Full Text Available A steam generator is an essential unit of each cogeneration system using steam machines. Currently one of the cheapest ways of the steam generation can be application of old steam generators came from army surplus store. They have relatively simple construction and in case of not so exploited units – quite good general conditions, and functionality of mechanical components. By contrast, electrical components and control systems (mostly based on relay automatics are definitely obsolete. It is not possible to use such units with cooperation of steam bus or with steam engines. In particular, there is no possibility for automatically adjustment of the pressure and the temperature of the generated steam supplying steam engines. Such adjustment is necessary in case of variation of a generator load. The paper is devoted to description of improvement of an exemplary unit together with construction of the measurement-control system based on a PLC. The aim was to enable for communication between the steam generator and controllers of the steam bus and steam engines in order to construction of a complete, fully autonomic and maintenance-free microcogeneration system.

  10. Repair technology for steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Ho; Jung, Hyun Kyu; Jung, Seung Ho; Kim, Chang Hoi; Jung, Young Moo; Seo, Yong Chil; Kim, Jung Su; Seo, Moo Hong

    2001-02-01

    The most commonly used sleeving materials are thermally treated Alloy 600 and thermally treated Alloy 690 Alloy. Currently, thermally treated Alloy 690 and Alloy 800 are being offered although Alloy 800 has not been licensed in the US. To install sleeve, joint strength, leak tightness, PWSCC resistance, evaluation on process parameter range and the effect of equipments and procedures on repair plan and radiation damage have to be investigated before sleeving. ABB CE provides three type of leak tight Alloy 690 TIG welded and PLUSS sleeve. Currently, Direct Tube Repair technique using Nd:YAG laser has been developed by ABB CE and Westinghouse. FTI has brazed and kinetic sleeve designs for recirculating steam generator and hydraulic and rolled sleeve designs for one-through steam generators. Westinghouse provides HEJ, brazed and laser welded sleeve design. When sleeve is installed in order to repair the damaged S/G tubes, it is certain that defects can be occurred due to the plastic induced stress and thermal stress. Therefore it is important to minimize the residual stress. FTI provides the electrosleeve technique as a future repair candidate using electroplating.

  11. Repair technology for steam generator tubes

    International Nuclear Information System (INIS)

    Kim, Seung Ho; Jung, Hyun Kyu; Jung, Seung Ho; Kim, Chang Hoi; Jung, Young Moo; Seo, Yong Chil; Kim, Jung Su; Seo, Moo Hong

    2001-02-01

    The most commonly used sleeving materials are thermally treated Alloy 600 and thermally treated Alloy 690 Alloy. Currently, thermally treated Alloy 690 and Alloy 800 are being offered although Alloy 800 has not been licensed in the US. To install sleeve, joint strength, leak tightness, PWSCC resistance, evaluation on process parameter range and the effect of equipments and procedures on repair plan and radiation damage have to be investigated before sleeving. ABB CE provides three type of leak tight Alloy 690 TIG welded and PLUSS sleeve. Currently, Direct Tube Repair technique using Nd:YAG laser has been developed by ABB CE and Westinghouse. FTI has brazed and kinetic sleeve designs for recirculating steam generator and hydraulic and rolled sleeve designs for one-through steam generators. Westinghouse provides HEJ, brazed and laser welded sleeve design. When sleeve is installed in order to repair the damaged S/G tubes, it is certain that defects can be occurred due to the plastic induced stress and thermal stress. Therefore it is important to minimize the residual stress. FTI provides the electrosleeve technique as a future repair candidate using electroplating

  12. Method for steam generator water level measurement

    International Nuclear Information System (INIS)

    Srinivasan, J.S.

    1991-01-01

    This paper describes a nuclear power plant, a method of controlling the steam generator water level, wherein the steam generator has an upper level tap corresponding to an upper level, a lower level, a riser positioned between the lower and upper taps, and level sensor means for indicating water level between a first range limit and a second range limit, the sensor means being connected to at least the lower tap. It comprises: calculating a measure of velocity head at about the lower level tap; calculating a measure of full water level as the upper level less the measure of velocity head; calibrating the level sensor means to provide an output at the first limit corresponding to an input thereto representative of the measure of full level; calculating a high level setpoint equal to the level of the riser less a bias amount which is a function of the position of the riser relative to the span between the taps; and controlling the water level when the sensor means indicates that the high level setpoint has been reached

  13. Heysham II/Torness AGR steam generator

    International Nuclear Information System (INIS)

    Charcharos, A.N.; Wood, M.B.; Glasgow, J.R.

    1988-01-01

    The AGR Steam Generators for Heysham II and Torness Power Stations have been installed at site and are being operated in the initial low temperature commissioning plant engineering tests. In this paper a description of the high pressure once-through steam generators together with layout arrangements, materials employed, operating parameters, plant operating conditions and constraints is given. An outline of the development of the design through thermo-hydraulic considerations, mechanical design, instrumentation to component testing is presented. Special features of the design directed to accommodate such requirements as seismic loadings, waterside static and dynamic stability, gas flow induced vibration, thermal expansions are described in detail. The fabrication facilities employed and techniques selected and developed for the manufacture and assembly of the heating surfaces are presented. These include welding processes, tube manipulation and heat treatment with details of the automation applied to the processes. Operating experience in the early commissioning plant engineering tests at Site is described with an emphasis on those tests which provide the final confirmation of the design prior to operation at full load. The paper concludes with a description of the outstanding commissioning activities up to raise power. (author)

  14. Heysham II/Torness AGR steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Charcharos, A N [National Nuclear Corporation Ltd., Knutsford (United Kingdom); Wood, M B; Glasgow, J R [NEI Power Projects Ltd., Gateshead (United Kingdom)

    1988-07-01

    The AGR Steam Generators for Heysham II and Torness Power Stations have been installed at site and are being operated in the initial low temperature commissioning plant engineering tests. In this paper a description of the high pressure once-through steam generators together with layout arrangements, materials employed, operating parameters, plant operating conditions and constraints is given. An outline of the development of the design through thermo-hydraulic considerations, mechanical design, instrumentation to component testing is presented. Special features of the design directed to accommodate such requirements as seismic loadings, waterside static and dynamic stability, gas flow induced vibration, thermal expansions are described in detail. The fabrication facilities employed and techniques selected and developed for the manufacture and assembly of the heating surfaces are presented. These include welding processes, tube manipulation and heat treatment with details of the automation applied to the processes. Operating experience in the early commissioning plant engineering tests at Site is described with an emphasis on those tests which provide the final confirmation of the design prior to operation at full load. The paper concludes with a description of the outstanding commissioning activities up to raise power. (author)

  15. The testing of a steam-water separating device used for vertical steam generators

    International Nuclear Information System (INIS)

    Ding Xunshen; Cui Baoyuan; Xue Yunkui; Liu Shixun

    1989-01-01

    The air-water screening tests of a steam-water separating device used for vertical steam generators at low pressure are introduced. The article puts emphasis on the qualification test of the steam-water separating device at hot conditions in a high temperature and pressure water test rig. The performance of the comprehensive test of the steam-water separating device indicates that the humidity of the steam at the drier exit is much less than the specified amount of 0.25%

  16. Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

    Energy Technology Data Exchange (ETDEWEB)

    Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate.

  17. Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

    International Nuclear Information System (INIS)

    Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo

    2016-01-01

    Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate

  18. Degradation monitoring in IRIS steam generators

    International Nuclear Information System (INIS)

    Alpay, B.; Holloway, J. P.; Lee, J. C.

    2007-01-01

    We present a degradation monitoring technique based on unscented Kalman filtering (UKF), which uses a nonlinear system model without linearization to estimate the status of the component/state variables. To test the applicability of the methodology, the fouling of tubes is chosen among various degradation mechanisms for the IRIS (International Reactor Innovative and Secure) steam generators (SGs). The degradation monitoring algorithm diagnoses the tube fouling and estimates the thickness of the crud deposited on the secondary side of the SG along with the increase in the pressure drop triggered by fouling. A stand-alone SG model developed with the RELAP5 code was used to simulate the transient behavior of the SG and drive an UKF state estimate. By using the secondary side outlet temperature as the measurement and the nodal pressures along the secondary side as states, UKF generated accurate estimates of the crud layer thicknesses for different crud formations. (authors)

  19. The ageing of CANDU steam generator due to localized corrosion

    International Nuclear Information System (INIS)

    Lucan, D.; Fulger, M.; Jinescu, Ghe.

    2001-01-01

    The principal types of corrosion are presented which can occur in CANDU steam generator. There are also presented the operation conditions, the specifications referring to the water chemistry and the construction materials of Steam Generator, the factors that have a great influence on the corrosion behaviour during the whole exploitation period of this equipment. The most important elements of CANDU Steam Generator ageing management program are also discussed. (R. P.)

  20. Future development LMFBR-steam generators SNR2

    International Nuclear Information System (INIS)

    Essebaggers, J.; Pors, J.G.

    1975-01-01

    The development work for steam generators for large LMFBR plants by Neratoom will be reviewed consisting of: 1. Development engineering information. 2. Concept select studies followed by conceptual designs of selected models. 3. Development manufacturing techniques. 4. Detail design of a prototype unit. 5. Testing of sub-constructions for prototype steam generators. In this presentation item 1 and 2 above will be high lighted, identifying the development work for the SNR-2 steam generators on short term basis. (author)

  1. The Creys Malville FBR Super Phenix steam generators

    International Nuclear Information System (INIS)

    Baque, P.; Zuber, T.; Saur, J.M.; Cambillard, E.

    1980-08-01

    After briefly recalling the French experience on sodium steam generators, the authors describe the design concepts of the Superphenix units and give their main characteristics. A short summary of the realized R and D program precedes the description of the four 750-MWt steam generators, the fabrication of which is in progress by Creusot-Loire at Chalon sur Saone (France). The studies started for the next French fast breeder reactors and their steam generators are mentioned

  2. Thermal-hydraulic characteristic of the PGV-1000 steam generator

    International Nuclear Information System (INIS)

    Ubra, O.; Doubek, M.

    1995-01-01

    Horizontal steam generators are typical parts of nuclear power plants with pressure water reactor type VVER. By means of this computer program, a detailed thermal-hydraulic study of the horizontal steam generator PGV-1000 has been carried out and a special attention has been paid to the thermal-hydraulics of the secondary side. A set of important steam generator characteristics has been obtained and analyzed. Some of the interesting results of the analysis are presented in the paper. (author)

  3. Thermal hydraulic studies in steam generator test facility

    International Nuclear Information System (INIS)

    Vinod, V.; Suresh Kumar, V.A.; Noushad, I.B.; Ellappan, T.R.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G.

    2005-01-01

    Full text of publication follows: A 500 MWe fast breeder reactor is being constructed at Kalpakkam, India. This is a sodium cooled reactor with two primary and two secondary sodium loops with total 8 steam generators. The typical advantage of fast breeder plants is the high operating temperature of steam cycles and the high plant efficiency. To produce this high pressure and high temperature steam, once through straight tube vertical sodium heated steam generators are used. The steam is generated from the heat produced in the reactor core and being transported through primary and secondary sodium circuits. The steam generator is a 25 m high middle supported steam generator with expansion bend and 23 m heat transfer length. Steam Generator Test Facility (SGTF) constructed at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam aims at performing various tests on a 5.5 MWt steam generator. This vertically simulated test article is similar in all respects to the proposed 157 MWt steam generator module for the Prototype Fast Breeder Reactor (PFBR), with reduced number of tubes. Heat transfer performance tests are done with this 19 tube steam generator at various load conditions. Sodium circuit for the SGTF is equipped with oil fired heater as heat source and centrifugal sodium pump, to pump sodium at 105 m 3 /hr flow rate. Other typical components like sodium to air heat exchanger, sodium purification system and hydrogen leak detection system is also present in the sodium circuit. High pressure steam produced in the steam generator is dumped in a condenser and recycled. Important tests planned in SGTF are the heat transfer performance test, stability test, endurance test and performance test of steam generator under various transients. The controlled operation of steam generator will be studied with possible control schemes. A steady state simulation of the steam generator is done with a mathematical model. This paper gives the details of heat transfer

  4. Vapor generator steam drum spray heat

    International Nuclear Information System (INIS)

    Fasnacht, F.A. Jr.

    1978-01-01

    A typical embodiment of the invention provides a combination feedwater and cooldown water spray head that is centrally disposed in the lower portion of a nuclear power plant steam drum. This structure not only discharges the feedwater in the hottest part of the steam drum, but also increases the time required for the feedwater to reach the steam drum shell, thereby further increasing the feedwater temperature before it contacts the shell surface, thus reducing thermal shock to the steam drum structure

  5. Evaluation of sulfur hexafluoride and helium for steam generator leak location: Final report

    International Nuclear Information System (INIS)

    Kassen, W.R.

    1987-01-01

    Since the use of sulfur hexafluoride as a tracer for identifying sources of primary to secondary leakage in PWR steam generators appeared to offer significant sensitivity advantages, the thermal stability of sulfur hexafluoride in water was evaluated at steam generator operating temperature. Significant decomposition was observed after 2 to 4 hours at temperature. Key decomposition products were fluoride and sulfide ions. Based on this observation and these limited test results, the use of SF 6 for PWR steam generator leak location can not be recommended at this time. A survey of 15 utilities was conducted in regard to their application experience with the helium tracer-mass spectroscopy technique for steam generator leak location. Although several successful steam generator integrity programs do not include use of this technique, it has proven to be a useful addition to the inspection program at some plants. No corrosion concerns appear to be associated with this technique

  6. SWAAM code development, verification and application to steam generator design

    International Nuclear Information System (INIS)

    Shin, Y.W.; Valentin, R.A.

    1990-01-01

    This paper describes the family of SWAAM codes developed by Argonne National Laboratory to analyze the effects of sodium/water reactions on LMR steam generators. The SWAAM codes were developed as design tools for analyzing various phenomena related to steam generator leaks and to predict the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The theoretical foundations and numerical treatments on which the codes are based are discussed, followed by a description of code capabilities and limitations, verification of the codes by comparison with experiment, and applications to steam generator and IHTS design. (author). 25 refs, 14 figs

  7. Corrosion aspects in steam generators of nuclear power plants

    International Nuclear Information System (INIS)

    Visoni, E.; Santos Pinto, M. dos

    1988-01-01

    Steam generators of pressurized water reactors (PWR), transfer heat from a primary coolant system to a secondary coolant system. Primary coolant water is heated in the core and passes through the steam generator that transfer heat to the secondary coolant water. However, the steam generator is dead for ionic impurities, corrosion products and fabrication/maintenence residues. These impurities concentrate between crevice and cracks. Many types of degradation mechanisms affect the tubes. The tubes are dented, craked, ovalized, wasted, etc. This paper describes the main corrosion problems in steam generators and includes the corrective actions to considered to reduce or eliminate these corrosion problems. (author) [pt

  8. Steam generator operating experience update, 1982-1983

    International Nuclear Information System (INIS)

    Frank, L.

    1984-06-01

    This report is a continuation of earlier reports by the staff addressing pressurized water reactor steam generator operating experience. NUREG-0886, Steam Generator Tube Experience, published in February 1982 summarized experience in domestic and foreign plants through December 1981. This report summarizes steam generator operating experience in domestic plants for the years 1982 and 1983. Included are new problems encountered with secondary-side loose parts, sulfur-induced stress-assisted corrosion cracking, and flow-induced vibrational wear in the new preheater design steam generators. The status of Unresolved Safety Issues A3, A4, and A5 is also discussed

  9. Condensate polisher application for PWR steam generator corrosion control

    International Nuclear Information System (INIS)

    Sawochka, S.G.; Leibovitz, J.; Siegwarth, D.P.; Pearl, W.L.

    1981-01-01

    The evolution of corrosion attack modes particularly in recirculating U-tube PWR steam generators has dictated a thorough review of the advantages and disadvantages of condensate polishing. Analytical modeling techniques to qualitatively predict crevice chemistry variations resulting from steam generator bulk water variations have allowed valuable insights to be developed. Modeling results complemented by steam generator and laboratory corrosion data will be employed to set condensate demineralizer effluent specifications consistent with control of steam generator corrosion. Laboratory and plant studies are being performed to demonstrate achievability of necessary effluent specifications. (author)

  10. Dismantling of the 50 MW steam generator test facility

    International Nuclear Information System (INIS)

    Nakai, S.; Onojima, T.; Yamamoto, S.; Akai, M.; Isozaki, T.; Gunji, M.; Yatabe, T.

    1997-01-01

    We have been dismantling the 50MW Steam Generator Test Facility (50MWSGTF). The objectives of the dismantling are reuse of sodium components to a planned large scale thermal hydraulics sodium test facility and the material examination of component that have been operated for long time in sodium. The facility consisted of primary sodium loop with sodium heater by gas burner as heat source instead of reactor, secondary sodium loop with auxiliary cooling system (ACS) and water/steam system with steam temperature and pressure reducer instead of turbine. It simulated the 1 loop of the Monju cooling system. The rated power of the facility was 50MWt and it was about 1/5 of the Monju power plant. Several sodium removal methods are applied. As for the components to be dismantled such as piping, intermediate heat exchanger (IHX), air cooled heat exchangers (AC), sodium is removed by steam with nitrogen gas in the air or sodium is burned in the air. As for steam generators which material tests are planned, sodium is removed by steam injection with nitrogen gas to the steam generator. The steam generator vessel is filled with nitrogen and no air in the steam generator during sodium removal. As for sodium pumps, pump internal structure is pulled out from the casing and installed into the tank. After the installation, sodium is removed by the same method of steam generator. As for relatively small reuse components such as sodium valves, electromagnet flow meters (EMFs) etc., sodium is removed by alcohol process. (author)

  11. Mathematical modeling of a steam generator for sensor fault detection

    International Nuclear Information System (INIS)

    Prock, J.

    1988-01-01

    A dynamic model for a nuclear power plant steam generator (vertical, preheated, U-tube recirculation-type) is formulated as a sixth-order nonlinear system. The model integrates nodal mass and energy balances for the primary water, the U-tube metal and the secondary water and steam. The downcomer flow is determined by a static balance of momentum. The mathematical system is solved using transient input data from the Philippsburg 2 (FRG) nuclear power plant. The results of the calculation are compared with actual measured values. The proposed model provides a low-cost tool for the automatic control and simulation of the steam generating process. The ''parity-space'' algorithm is used to demonstrate the applicability of the mathematical model for sensor fault detection and identification purposes. This technique provides a powerful means of generating temporal analytical redundancy between sensor signals. It demonstrates good detection rates of sensor errors using relatively few steps of scanning time and allows the reconfiguration of faulty signals. (author)

  12. Steam generator corrosion 2007; Dampferzeugerkorrosion 2007

    Energy Technology Data Exchange (ETDEWEB)

    Born, M. (ed.)

    2007-07-01

    Between 8th and 9th November, 2007, SAXONIA Standortentwicklungs- und -verwertungsgesellschaft GmbH (Freiberg, Federal Republic of Germany) performed the 3rd Freiberger discussion conference ''Fireside boiler corrosion''. The topics of the lectures are: (a) Steam generator corrosion - an infinite history (Franz W. Alvert); (b) CFD computations for thermal waste treatment plants - a contribution for the damage recognition and remedy (Klaus Goerner, Thomas Klasen); (c) Experiences with the use of corrosion probes (Siegfried R. Horn, Ferdinand Haider, Barbara Waldmann, Ragnar Warnecke); (d) Use of additives for the limitation of the high temperature chlorine corrosion as an option apart from other measures to the corrosion protection (Wolfgang Spiegel); (e) Current research results and aims of research with respect to chlorine corrosion (Ragnar Warnecke); (f) Systematics of the corrosion phenomena - notes for the enterprise and corrosion protection (Thomas Herzog, Wolfgang Spiegel, Werner Schmidl); (g) Corrosion protection by cladding in steam generators of waste incinerators (Joerg Metschke); (h) Corrosion protection and wear protection by means of thermal spraying in steam generators (Dietmar Bendix); (i) Review of thick film nickelized components as an effective protection against high-temperature corrosion (Johann-Wilhelm Ansey); (j) Fireproof materials for waste incinerators - characteristics and profile of requirement (Johannes Imle); (k) Service life-relevant aspects of fireproof linings in the thermal recycling of waste (Till Osthoevener and Wolfgang Kollenberg); (l) Alternatives to the fireproof material in the heating space (Heino Sinn); (m) Cladding: Inconal 625 contra 686 - Fundamentals / applications in boiler construction and plant construction (Wolfgang Hoffmeister); (n) Thin films as efficient corrosion barriers - thermal spray coating in waste incinerators and biomass firing (Ruediger W. Schuelein, Steffen Hoehne, Friedrich

  13. Implementation status of performance demonstration program for steam generator tubing analysts in Korea

    International Nuclear Information System (INIS)

    Cho, Chan Hee; Lee, Hee Jong; Yoo, Hyun Ju; Nam, Min Woo; Hong, Sung Yull

    2013-01-01

    Some essential components in nuclear power plants are periodically inspected using non destructive examinations, for example ultrasonic, eddy current and radiographic examinations, in order to determine their integrity. These components include nuclear power plant items such as vessels, containments, piping systems, pumps, valves, tubes and core support structure. Steam generator tubes have an important safety role because they constitute one of the primary barriers between the radioactive and non radioactive sides of the nuclear power plant. There is potential that if a tube bursts while a plant is operating, radioactivity from the primary coolant system could escape directly to the atmosphere. Therefore, in service inspections are critical in maintaining steam generator tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due to its high inspection speed and flaw detectability on non magnetic tubes. However, it is not easy to analyze correctly eddy current signals because they are influenced by many factors. Therefore, the performance of eddy current data analysts for steam generator tubing should be demonstrated comprehensively. In Korea, the performance of steam generator tubing analysts has been demonstrated using the Qualified Data Analyst program. This paper describes the performance demonstration program for steam generator tubing analysts and its implementation results in Korea. The pass rate of domestic analysts for this program was 71.4%

  14. Implementation status of performance demonstration program for steam generator tubing analysts in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Chan Hee; Lee, Hee Jong; Yoo, Hyun Ju; Nam, Min Woo [KHNP Central Research Institute, Daejeon (Korea, Republic of); Hong, Sung Yull [Yeungnam Univ., Gyeongsan (Korea, Republic of)

    2013-02-15

    Some essential components in nuclear power plants are periodically inspected using non destructive examinations, for example ultrasonic, eddy current and radiographic examinations, in order to determine their integrity. These components include nuclear power plant items such as vessels, containments, piping systems, pumps, valves, tubes and core support structure. Steam generator tubes have an important safety role because they constitute one of the primary barriers between the radioactive and non radioactive sides of the nuclear power plant. There is potential that if a tube bursts while a plant is operating, radioactivity from the primary coolant system could escape directly to the atmosphere. Therefore, in service inspections are critical in maintaining steam generator tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due to its high inspection speed and flaw detectability on non magnetic tubes. However, it is not easy to analyze correctly eddy current signals because they are influenced by many factors. Therefore, the performance of eddy current data analysts for steam generator tubing should be demonstrated comprehensively. In Korea, the performance of steam generator tubing analysts has been demonstrated using the Qualified Data Analyst program. This paper describes the performance demonstration program for steam generator tubing analysts and its implementation results in Korea. The pass rate of domestic analysts for this program was 71.4%.

  15. The Steam Generating Heavy Water Reactor

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1975-01-01

    An account is given of the SGHWR, the prototype of which was built by the United Kingdom Atomic Energy Authority at Winfrith, under the following headings: Introduction; origin of the SGHWR concept; conceptual design (choice of reactor type, steam cycle, reactor coolant system, nuclear behaviour, fuel design, core design, and protective, auxiliary and containment systems); operation and control (integrity of core cooling, reactivity control, power trimming, long term reactivity control, xenon override, load following, power shaping, spatial stability control, void coefficient); protective systems (breached coolant circuit trip, intact coolant circuits trip, power set-back trip); dynamic characteristics; reactor control; station control (decoupled control system, coupled control system, rate of response); Winfrith prototype (design and safety philosophy, conceptual features and parameters, reactor coolant system, protective systems, emergency core cooling, core structure, fuel design, vented containment). (U.K.)

  16. Fabrication and inspection development for CRBRP steam generators

    International Nuclear Information System (INIS)

    McClung, R.W.; Slaughter, G.M.; Spalaris, C.N.; Lillie, A.F.

    1975-09-01

    One of the critical nonnuclear elements of the CRBRP is the steam generator that transfers the heat from the sodium system to the high-pressure steam system but must maintain integrity and separation of the two fluids. The construction material is 2 1 / 4 Cr--1 Mo alloy steel with high-purity (e.g. vacuum arc remelt) material being used for the tubing and tubesheets. For confidence in successful manufacturing of the several evaporator and superheater modules, key development activities are under way (1) for procurement of high-quality components, (2) to assure proper assembly (with emphasis on welding), and (3) to assure that adequate nondestructive testing methods are available to examine the units. (auth)

  17. Advanced life-cycle management for an increased steam generator performance

    International Nuclear Information System (INIS)

    Beck, J.; Schwarz, T.; Bouecke, R.; Schneider, V.

    2006-01-01

    High steam generators performance is a prerequisite for high plant availability and possible life time extension. During operation, the performance is reduced by fouling of the heating tubes and by corrosion, resulting on a reduction of the heat-exchange area. Such steam generator degradation problems arise from mechanical degradation and a continuous ingress of non-volatile contaminants, i.e. corrosion products and salt impurities accumulated in the steam generators. In addition, the tube scales in general affect the steam generator thermal performance, which ultimately cause a reduction of power output. AREVA applied an integrated service for utilities to evaluate all operational parameters influencing the steam generator performance. The evaluation is assisted by a systematic approach to evaluate the major steam generator operational data. The different data are structured and indexed in a Cleanling-Matrix. The result of this matrix is a quantified, dimensionless figure, given as the Fouling Index. The Fouling Index allows to monitor the condition of steam generators, compare it to other plants and, in combination with a life-time management applied at several German utilities, it allows verified statements on the past operation. Based on these data, an extrapolation of the potential additional life-time of the component is possible. As such, the Fouling Index is a valuable tool concerning life-time extension considerations. The application of the cleanliness criteria in combination with operational data with respect to life-time monitoring and improvements of steam generator performance are presented. (author)

  18. Comments on US LMFBR steam generator base technology

    International Nuclear Information System (INIS)

    Simmons, W.R.

    1984-01-01

    The development of steam generators for the LMFBR was recognized from the onset by the AEC, now DOE, as a difficult, challenging, and high-priority task. The highly reactive nature of sodium with water/steam requires that the sodium-water/steam boundaries of LMFBR steam generators possess a degree of leak-tightness reliability not normally attempted on a commercial scale. In addition, the LMFBR steam generator is subjected to high fluid temperatures and severe thermal transients. These requirements place great demand on materials, fabrication processes, and inspection methods; and even greater demands on the designer to provide steam generators that can meet these demanding requirements, be fabricated without unreasonable shop requirements, and tolerate off-normal effects

  19. Radioactivity transport following steam generator tube rupture

    International Nuclear Information System (INIS)

    Hopenfeld, J.

    1985-03-01

    A review of the capabilities of the CITADEL computer code as well as plant experience to project radioactivity releases following a steam generator tube rupture in PWR's shows that certain experimental data are needed for reliable off-site dose predictions. This article defines five parameters which are the key for such predictions and discusses the functional dependence of these parameters on various operational variables. The above parameters can be used in conjunction with CITADEL or they can be inserted in the appropriate equations which then conveniently can be programmed as a subroutine in thermal-hydraulic system codes. A joint Westinghouse, Electric Power Research Institute and Nuclear Regulatory Commission Program aimed at obtaining the five parameters empirically is described

  20. Internal ultrasonic testing of steam generator tubes

    International Nuclear Information System (INIS)

    Furlan, J.; Soleille, G.; Chalaye, H.

    1983-01-01

    The ''in situ'' inspection of steam generator tubes uses generally Foucault currents before starting and along its life. This inspection aims at searching cracks and corrosion defects. The Foucault current method is quite badly adapted to ''closed crack'' detection, for it doesn't introduce neither resistivity or magnetic permeability variation, or lack of matter. More, it is sensible to the magnetic properties of the tube itself and to its environment (tubular or support plates). It is why, this first systematic inspection has to be completed by an ultrasonic one allowing to bring new elements in the uncertain cases. A device with an internal probe has been developed. It ''lights'' the tube wall with the aid of a transducer of which beam reflects on a mirror. Operating conditions are the same as for Foucault current testing, that is to say the probe moves inside the tube without rotation of the device (bent parts are excluded) [fr

  1. Welding for the CRBRP steam generators

    International Nuclear Information System (INIS)

    Spalaris, C.N.; Ring, P.J.; Durand, R.E.; Wright, E.A.

    1979-01-01

    The rationale for selecting weld design, welding procedures and inspection methods was based upon the desire to obtain the highest reliability welds for the CRBRP steam generators. To assure the highest weld reliability, heavy emphasis was placed on the control of material cleanliness and composition substantially exceeding the requirements of the ASME Code for 2-1/4Cr--1Mo. The high tube/tubesheet weld quality was achieved through close material control, an extensive weld development program and the selection of high reliability welding equipment. Shell and nozzle weld fabrication using TIG, MIG, and submerged arc procedures are also being controlled through precise specifications, including preheat and postheat programs, together with radiography and ultrasonic inspection to ascertain the weld quality desired. Details of the tube/tubesheet welding and shell welding are described and results from the weld testing program are discussed

  2. Condensation induced water hammer in steam generators

    International Nuclear Information System (INIS)

    Jones, O.C. Jr.; Saha, P.; Wu, B.J.C.; Ginsberg, T.

    1979-06-01

    The case of condensation induced water hammer in nuclear steam generators is summarized, including both feed ring-type and economizer-type geometries. A slug impact model is described and used to demonstrate the parametric dependence of the impact pressures on heat transfer rates, initial pressures, and relative initial slug and void lengths. The results of the parametric study are related also to the economizer geometry and a suggested alternative model is presented. The importance of concerns regarding attenuation of shocks in two-phase media is delineated, and a simple experiment is described which was used to determine negligible attenuation within the accuracy of the experiment for void fractions up to over 30% in bubbly and slug flows

  3. Improved servicing equipment for steam generators

    International Nuclear Information System (INIS)

    Hedtke, James C.

    1998-01-01

    To help keep personnel exposure as low as reasonably achievable and reduce critical path outage time, most nuclear plants of PWR design in the USA are now using improved equipment to service their steam generators (SGs) during outages. Because of the success of this equipment in the USA, two Belgian plants and one English plant have purchased this equipment, and other nuclear plants in Europe are also considering procurement. The improved SG servicing equipment discussed in this paper discusses consists of nozzle dams, segmented multi-stud tensioner, primary manway cover handling tool set, shield door and fastener cleaner. This equipment is specifically designed for the individual plant application and can also be specified for replacement SG projects. All of the equipment can be used without modification of the existing SGs. (author)

  4. Steam generator issues in the United States

    International Nuclear Information System (INIS)

    Strosnider, J.R.

    1997-01-01

    Alloy 600 steam generator tubes in the US have exhibited degradation mechanisms similar to those observed in other countries. Effective programs have been implemented to address several degradation mechanisms including: wastage; mechanical wear; pitting; and fatigue. These degradation mechanisms are fairly well understood as indicated by the ability to effectively mitigate/manage them. Stress corrosion cracking (SCC) is the dominant degradation mechanism in the US. SCC poses significant inspection and management challenges to the industry and the regulators. The paper also addresses issues of research into SCC, inspection programs, plugging, repair strategies, water chemistry, and regulatory control. Emerging issues in the US include: parent tube cracking at sleeve joints; detection and repair of circumferential cracks; free span cracking; inspection and cracking of dented regions; and severe accident analysis

  5. Flaw analysis in steam generator tube

    International Nuclear Information System (INIS)

    Hutin, J.P.; Billon, F.

    1985-08-01

    Operating more than 30 PWR units, Electricite de France has to face several steam generator tube problems. One of the most serious difficulties is the stress corrosion cracking due to primary fluid, just above the tube sheet, in the roll transition region. With regard to availability it is, of course, a major concern; with regard to safety, the point is that tube rupture should be preceded by a significant primary-to-secondary leak during normal operation so that the reactor should be shut down before failure occurs. The demonstration of this assessment asks for experimental and analytical evidences. In 1981, Elecricite de France started a comprehensive program on that subject. A general description of this program and the main results are to be presented during the SMIRT-8 Conference. The purpose of the present paper is to develop in greater detail the analytical part of the work

  6. Development of advanced PWR steam generator

    International Nuclear Information System (INIS)

    Saito, Itaru; Nakamura, Tomomichi

    1999-01-01

    In response to the increased power of the advanced PWR, it is necessary to develop a steam generator (SG) which has a large capacity with high performance and high reliability as well as being economical to produce. In this paper, the development of the design of a new SG for the advanced PWRs is described and compared with the design of a conventional SG. Moreover, an outline of a seismic verification test for the U-bend tube bundle which includes advanced anti-vibration bars (AVB) which are very important is described. As a result, it was verified that the bundle has sufficient strength and a relatively high attenuation to seismic loads. These results will be reflected in the detailed design of advanced AVBs. (author)

  7. Wastage of Steam Generator Tubes by Sodium-Water Reaction

    International Nuclear Information System (INIS)

    Jeong, Ji Young; Kim, Jong Man; Kim, Tae Joon; Choi, Jong Hyeun; Kim, Byung Ho; Lee, Yong Bum; Park, Nam Cook

    2010-01-01

    The Korea Advanced LIquid MEtal Reactor (KALIMER) steam generator is a helical coil, vertically oriented, shell-and-tube type heat exchanger with fixed tube-sheet. The conceptual design and outline drawing of the steam generator are shown in Figure 1. Flow is counter-current, with sodium on the shell side and water/steam on the tube side. Sodium flow enters the steam generator through the upper inlet nozzles and then flows down through the tube bundle. Feedwater enters the steam generator through the feedwater nozzles at the bottom of steam generator. Therefore, if there is a hole or a crack in a heat transfer tube, a leakage of water/steam into the sodium may occur, resulting in a sodium-water reaction. When such a leak occurs, so-called 'wastage' is the result which may cause damage to or a failure of the adjacent tubes. If a steam generator is operated for some time in this condition, it is possible that it might create an intermediate leak state which would then give rise to the problems of a multi-target wastage in a very short time. Therefore, it is very important to predict these phenomena quantitatively from the view of designing a steam generator and its leak detection systems. For this, multi-target wastage tests for modified 9Cr-1Mo steel tube bundle by intermediate leaks are being prepared

  8. Chooz-A Steam Generators Characterization

    International Nuclear Information System (INIS)

    Aitammar, Laurie

    2016-01-01

    EDF nuclear waste management requires a deep understanding of characterization, classification and waste sorting operations. In fact, French nuclear waste management defines several classes with specific management, treatment and storage facilities. Based on particular criteria, the more the radiological risk of the nuclear waste is important, the more its management will be complex and expensive. During the dismantling of the first French pressurized reactor Chooz-A, decontamination of the primary water circuits (not including the reactor vessel), the steam generators and the pressurizer have been carried out in order to reduce their activity levels. Thanks to these decontamination operations, and a specific characterization methodology, EDF was able to Re-classify the 4 steam generators and store them in one piece at the ANDRA Very Low Level Activity disposal facilities instead of the Low and Intermediate Level activity one. This re-classification allowed EDF to avoid important cutting and packaging processes. To characterize and declare Chooz-A SG activity, EDF-CIDEN used a methodology defined by the French institute of atomic energy, CEA. The method is based on external gamma spectrometry measurements performed with NaI collimated detectors, associated with MERCURAD simulations providing the transfer functions for the detectors and activity sources. Internal measurements are carried out with a CZT (CdZnTe) probe inside the SG tubes to refine the 3D model. In fact, the primary side represents the main source of activity, and understanding its contamination distribution is important to reduce the model and calculation uncertainties. Measurements eventually provided SG 60 Co global activity, from which the activities of other radionuclides of the spectrum were determined using scaling factors. The final activity declaration takes into account the standard deviation of the measurements in order to cover the uncertainties of the methodology. Thereby, the declaration

  9. Comparison of steam generator methods in PISC

    International Nuclear Information System (INIS)

    Lahdenperae, K.; Kankare, M.

    1996-01-01

    The main objective of the study (PISC III, action 5) was the experimental evaluation of the performance of methods used in in-service inspection of steam generator tubes used in nuclear power plants. The study was organized by the Joint Research Center of the European Community (JRC). The round robin test with blind boxes started in 1991. During the study training boxes and blind boxes were circulated in 29 laboratories in Europe, Japan and the USA. The boxes contained steam generator tubes with artificial and natural (chemically induced) flaws. The material was inconell. The blind boxes contained 66 tubes and 95 flaws. All flaws were introduced into different discontinuities, under support plates, above the tube sheet and into U-bends. The flaws included volumetric flaws (wastage, pitting, wear), axial and circumferential notches and chemically induced SCC cracks and IGA. After the round robin test the reference laboratory performed the destructive examination of reported flaws. The flaw detection probability (FDP) for all flaws and for teams inspecting all tubes was 60-85%. The detection of flaws deeper than 40% of the wall thickness was good. Flaws with a depth of less than 20% were not detected. When all flaws were considered, depth sizing was found to have a wide dispersion. Similarly, measured lengths did not as a rule correlate with true lengths. The classification of flaws in cracks and of volumetric flaws was not very successful, the correct classification probability being only about 70%. Evaluation of the flaws showed some shortcomings. The correct rejection probability was at best 83% for teams inspecting all boxes. (3 refs.)

  10. Modeling and Simulation of U-tube Steam Generator

    Science.gov (United States)

    Zhang, Mingming; Fu, Zhongguang; Li, Jinyao; Wang, Mingfei

    2018-03-01

    The U-tube natural circulation steam generator was mainly researched with modeling and simulation in this article. The research is based on simuworks system simulation software platform. By analyzing the structural characteristics and the operating principle of U-tube steam generator, there are 14 control volumes in the model, including primary side, secondary side, down channel and steam plenum, etc. The model depends completely on conservation laws, and it is applied to make some simulation tests. The results show that the model is capable of simulating properly the dynamic response of U-tube steam generator.

  11. Control concepts for direct steam generation in parabolic troughs

    Energy Technology Data Exchange (ETDEWEB)

    Valenzuela, Loreto; Zarza, Eduardo [CIEMAT, Plataforma Solar de Almeria, Tabernas (Almeria) (Spain); Berenguel, Manuel [Universidad de Almeria, Dept. de Lenguajes y Computacion, Almeria (Spain); Camacho, Eduardo F. [Universidad de Sevilla, Dept. de Ingenieria de Sistemas y Automatica, Sevilla (Spain)

    2005-02-01

    A new prototype parabolic-trough collector system was erected at the Plataforma Solar de Almeria (PSA) (1996-1998) to investigate direct steam generation (DSG) in a solar thermal power plant under real solar conditions. The system has been under evaluation for efficiency, cost, control and other parameters since 1999. The main objective of the control system is to obtain steam at constant temperature and pressure at the solar field outlet, so that changes in inlet water conditions and/or in solar radiation affect the amount of steam, but not its quality or the nominal plant efficiency. This paper presents control schemes designed and tested for two operating modes, 'Recirculation', for which a proportional-integral-derivative (PI/PID) control functions scheme has been implemented, and 'Once-through', requiring more complex control strategies, for which the scheme is based on proportional-integral (PI), feedforward and cascade control. Experimental results of both operation modes are discussed. (Author)

  12. Methodology for demonstrating the integrity of Steam Generator Tubes NPP Almaraz; Metodologia para la demostracion de la integridad de los tubos de Generador de Vapor de C. N. Almaraz

    Energy Technology Data Exchange (ETDEWEB)

    Campana Martin, J.; Cueto-Felgueroso Garcia, C.

    2013-07-01

    Steam Generator Program requires the performance of a Degradation Assessment prior to each refueling outage. The overall purpose of DA is to ensure that appropriate inspections are performed during the upcoming outage, and that the requisite information for integrity assessment is provided. Integrity assessment is performed after each SG tube inspection and includes two stages. The first one, Condition Monitoring is an assessment which confirms that SG tubes have met Performance Criteria during previous inspection interval. The second stage, Operational Assessment is an assessment which demonstrates that Performance Criteria will be met during the next inspection interval.

  13. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

    2009-01-01

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  14. Status of Siemens steam generator design and measures to assure continuous long-term reliable operation

    International Nuclear Information System (INIS)

    Hoch, G.

    1999-01-01

    Operating pressurized water reactors with U-tube steam generators have encountered difficulties with either one or a combination of inadequate material selection, poor design or manufacturing and an insufficient water chemistry control which resulted in excessive tube degradation. In contrast to the above mentioned problems, steam generators from Siemens/KWU are proving by operating experience that all measures undertaken at the design stage as well as during the operating and maintenance phase were effective enough to counteract any tube corrosion phenomena or other steam generator related problem. An Integrated Service Concept has been developed, applied and wherever necessary improved in order to ensure reliable steam generator operation. The performance of the steam generators is updated continuously, evaluated and implemented in lifetime databases. The main indicator for steam generator integrity are the results of the eddy current testing of the steam generator tubes. Tubes with indications are rated with lifetime threshold values and if necessary plugged, based on individual assessment criteria.(author)

  15. Numerical modeling of secondary side thermohydraulics of horizontal steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Melikhov, V.I.; Melikhov, O.I.; Nigmatulin, B.I. [Research and Engineering Centre of LWR Nuclear Plants Safety, Moscow (Russian Federation)

    1995-12-31

    A mathematical model for the transient three-dimensional secondary side thermal hydraulics of the horizontal steam generator has been developed. The calculations of the steam generator PGV-1000 and PGV-4 nominal regimes and comparison of numerical and experimental results have been carried out. 7 refs.

  16. Influence of recycling ratio on steam generator thermal recycling

    International Nuclear Information System (INIS)

    Bassel, W.S.; Rodrigues, V.G.

    1989-01-01

    A mathematical model was developed to simulate thermal performance of steam generator. The simulation was done with 3 control volumes. The coupled non-linear algebric equations, where the heat transfer was calculated with logarithmic meam temperature difference, was solved by iterative method. The developed model is suitable for calculation the parameters which effect the performance of steam generator. (author) [pt

  17. Modelling of a Coil Steam Generator for CSP applications

    DEFF Research Database (Denmark)

    Pelagotti, Leonardo; Sørensen, Kim; Condra, Thomas Joseph

    2014-01-01

    The project investigates a new design for a CSP plant steam generation system, the Coil Steam Generator (CSG). This system allows faster start-ups and therefore higher daily energy production from the Sun. An analytical thermodynamic simulation model of the evaporator and a mechanical analysis...

  18. Materials choices for the advanced LWR steam generators

    International Nuclear Information System (INIS)

    Paine, J.P.N.; Shoemaker, C.E.; McIlree, A.R.

    1987-01-01

    Current light water reactor (LWR) steam generators have been affected by a variety of corrosion and mechanical damage degradation mechanisms. Included are wear caused by tube vibration, intergranular corrosion, pitting, and thinning or wastage of the steam generator tubing and accelerated corrosion of carbon steel supports (denting). The Electric Power Research Institute (EPRI) and the Steam Generator Owners Groups (I, II) have sponsored laboratory and field studies to provide ameliorative actions for the majority of the damage forms experienced to date. Some of the current corrosion mechanisms are aggravated or caused by unique materials choices or materials interactions. New materials have been proposed and at least partially qualified for use in replacement model steam generators, including an advanced LWR design. In so far as possible, the materials choices for the advanced LWR steam generator avoid the corrosion pitfalls seemingly inherent in the current designs. The EPRI Steam Generator Project staff has recommended materials and design choices for a new steam generator. Based on these recommendations we believe that the advanced LWR steam generators will be much less affected by corrosion and mechanical damage mechanisms than are now experienced

  19. Numerical modeling of secondary side thermohydraulics of horizontal steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Melikhov, V I; Melikhov, O I; Nigmatulin, B I [Research and Engineering Centre of LWR Nuclear Plants Safety, Moscow (Russian Federation)

    1996-12-31

    A mathematical model for the transient three-dimensional secondary side thermal hydraulics of the horizontal steam generator has been developed. The calculations of the steam generator PGV-1000 and PGV-4 nominal regimes and comparison of numerical and experimental results have been carried out. 7 refs.

  20. Considerations in selecting tubing materials for CANDU steam generators

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    1978-01-01

    Corrosion resistance is the major consideration in selecting tubing material for CANDU steam generators. Corrosion, and additional considerations, lead to the following steam generator tubing material recommendations: for CANDU-BPHWR's (boiling pressurized heavy water reactors) low-cobalt Incoloy-800; for CANDU-PHWR's (pressurized, non-boiling, heavy water reactors), low-cobalt Monel-400

  1. Dynamic analysis of CHASNUPP steam generator structure during shipping

    International Nuclear Information System (INIS)

    Han Liangbi; Xu Jinkang; Zhou Meiwu; He Yinbiao

    1998-07-01

    The dynamic analysis of CHASNUPP steam generator during shipping is described, including the simplified mathematical model, acceleration power spectrum of ocean wave induced random vibration, the dynamic analysis of steam generator structure under random loading, the applied computer code and calculated results

  2. Reconstruction of steam generators super emergency feadwater supply system (SHNC) and steam dump stations to the atmosphere system PSA

    International Nuclear Information System (INIS)

    Kuzma, J.

    2001-01-01

    Steam Generators Super Emergency Feadwater Supply System (SHNC) and Steam Dump Stations to the Atmosphere System (PSA) are two systems which cooperate to remove residual heat from reactor core after seismic event. SHNC assure feeding of the secondary site of steam generator (Feed) where after heat removal.from primary loops, is relieved to the atmosphere by PSA (Bleed) in form of steam. (author)

  3. Design of a steam generator for PWR power plants and steady state simulation

    International Nuclear Information System (INIS)

    Ferreira, W.J.

    1982-01-01

    A procedure and a computer code for the thermal design of a steam generator for PWR power plants is developed. A vertical integral steam generator with inverted U-tubes and natural circulation of the secondary side is selected for modelling. Primary fluid velocity and recirculation ratio are varied to obtain the preliminary dimensions. Further, adjustments are made through iteractive solution of the equations of conservation of mass, energy and momentum. An agreement is found between design calculations for steam generators of different capacities and existing designs. (Author) [pt

  4. Improved in-service inspection program for management of degradation in steam generator tubing

    International Nuclear Information System (INIS)

    Kurtz, R.; Heasler, P.; Muscara, J.

    1992-01-01

    This paper presents an overview of significant results from NRC-sponsored research on steam generator tube integrity and inspection. Burst test results are described along with empirical models to relate flaw geometry and size to tube burst pressure. Results of round robin examinations of a retired-from-service steam generator to determine eddy current inspection reliability are presented. An evaluation and comparison of various sampling plans for in-service inspection of steam generators is discussed. Finally, performance demonstration qualification efforts for eddy current inspection systems are described

  5. Steam generators under construction for the SNR-300 power plant

    Energy Technology Data Exchange (ETDEWEB)

    Essebaggers, J

    1975-07-01

    The prototype straight tube and the helical coil-steam generator has been designed and fabricated of which the straight tube steam generator has been successfully tested for over 3000 hours at prototypical conditions and is presently being dismantled for detailed examination of critical designed features. The prototype helical coil steam generator is presently under testing in the 50 MWt test facility at TNO-Hengelo with approximately 500 hours of operation at full load conditions. In an earlier presentation the design and fabrication of the prototype steam generators have been presented, while for this presentation the production units for SNR-300 will be discussed. Some preliminary information will be presented at this meeting of the dismantling operations of the prototype straight tube steam generator. (author)

  6. Steam generators under construction for the SNR-300 power plant

    International Nuclear Information System (INIS)

    Essebaggers, J.

    1975-01-01

    The prototype straight tube and the helical coil-steam generator has been designed and fabricated of which the straight tube steam generator has been successfully tested for over 3000 hours at prototypical conditions and is presently being dismantled for detailed examination of critical designed features. The prototype helical coil steam generator is presently under testing in the 50 MWt test facility at TNO-Hengelo with approximately 500 hours of operation at full load conditions. In an earlier presentation the design and fabrication of the prototype steam generators have been presented, while for this presentation the production units for SNR-300 will be discussed. Some preliminary information will be presented at this meeting of the dismantling operations of the prototype straight tube steam generator. (author)

  7. Equipment transporter for nuclear steam generator

    International Nuclear Information System (INIS)

    Hayes, L.R.

    1987-01-01

    A transporter is described for use in a steam generator of a nuclear power installation. The generator is essentially a heat exchanger having a vertically extended shell. Across the lower portion extends a horizontal tube sheet having an upper surface which supports a bundle of vertically extending tubes forming a limited annular space with the inside of the shell wall and the upper surface. An opening of limited dimensions through the shell wall gains manual access to the limited annular space. The transporter has means for locating and removing solid debris from the upper surface of the tube sheet in the annular space and has a means for assembly and disassembly of the transporter so that it may be manually passed through the shell opening to and from a position on the upper surface of the tube sheet in the annular space. The transporter includes: a body; at least three wheels mounted on the body for engaging the upper surface of the tube sheet; a first motor mounted on the body drivingly connected to the wheels for moving the transporter along the upper surface of the tube sheet in the annular space; a remotely operated means on the body for locating solid debris on the upper surface of the tube sheet; and means for securing and removing solid debris on the upper surface of the tube sheet located by the means for locating

  8. Large-leak sodium-water reaction analysis for steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Sakano, K; Shindo, Y; Hori, M

    1975-07-01

    The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

  9. Large-leak sodium-water reaction analysis for steam generators

    International Nuclear Information System (INIS)

    Sakano, K.; Shindo, Y.; Hori, M.

    1975-01-01

    The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

  10. System for combustion of sunflower shells in industrial steam generators

    International Nuclear Information System (INIS)

    Todoriev, Kh.

    2000-01-01

    The paper presents an economically efficient solution for reconstruction of steam generators with steam production over 5 t/h using foregoing cyclone chamber for sunflower shells combustion. For average fuel caloricity 9 445 ccal/kg and sunflower shells caloricity between 3 485 and 3 750 ccal/kg, the petroleum saving is 68.78% for an average boiler efficiency 4.6 t/h steam

  11. THTR steam generator licensing experience as seen by the manufacturer

    International Nuclear Information System (INIS)

    Fricker, H.W.

    1981-01-01

    This paper describes the licensing procedures of the manufacture of the 300 MWe THTR steam generator. The following problems are discussed: operating data, design, materials used, manufacture and installation of the generator, and also quality control

  12. Computer codes for simulation of Angra 1 reactor steam generator

    International Nuclear Information System (INIS)

    Pinto, A.C.

    1978-01-01

    A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt

  13. Mechanical design of a sodium heated steam generator

    International Nuclear Information System (INIS)

    Chetal, S.C.

    1975-01-01

    FBTR steam generator is a once through type unit consisting of four 12.5 MW thermal modules generating a total of 74 tons per hour of steam at 125 bar and 480 0 C. This paper outlines the mechanical design of such type of steam generator with emphasis on special design problems associated with this type of sodium to water steam heat exchanger, namely, thermal cycling of transition zone where nucleate boiling changes over to film boiling, application of pressure vessel design criteria for transient pressures, thermal stress evaluation resulting from differential expansion between shell and tube in this typical configuration, sodium headers support design, thermal sleeve, design, thermal shock analysis in thick tubes, thermal stress resulting from stratification and stability of expansion bends against vibration. Some of the possible design changes for the future large size steam generator are outlined. (author)

  14. Seed-a distributed data base architecture for global management of steam-generator inspection data

    International Nuclear Information System (INIS)

    Soon Ju Kang; Yu Rak Choi; Hee Gon Woo; Seong Su Choi

    1996-01-01

    This paper deals with a data management system - called SEED (Steam-generator Eddy-current Expert Database) for global handling of SG (steam generator) tube inspection data in nuclear power plants. The SEED integrates all stages in SG tube inspection process and supports all data such as raw eddy current data, inspection history data, SG tube information, etc. SEED implemented under client/server computing architecture for supporting LAN/WAN based graphical user interface facilities using WWW programming tools. (author)

  15. An homogeneous model of steam generator to simulate operational transiento and accidents in PWR nuclear power plants

    International Nuclear Information System (INIS)

    Souza, A.L. de.

    1981-07-01

    GEVAP - A digital computer code was developed to simulate the thermodynamic transient behaviour of steam generators. The steam generator is divided in heating sections. In each section, the conservation equations of mass and energy are integrated numerically, using a predictor-corrector method. As good reslts where obtained, as compared to transients simulated using more detainled codes, it is concluded that GEVAP can be included as the steam generator module of a more complete systems simulation code for PWR's. (E.G.) [pt

  16. Dancing with STEAM: Creative Movement Generates Electricity for Young Learners

    Science.gov (United States)

    Simpson Steele, Jamie; Fulton, Lori; Fanning, Lisa

    2016-01-01

    The integration of science, technology, engineering, arts, and mathematics (STEAM) serves to develop creative thinking and twenty-first-century skills in the classroom (Maeda 2012). Learning through STEAM promotes novelty, innovation, ingenuity, and task-specific purposefulness to solve real-world problems--all aspects that define creativity. Lisa…

  17. Advanced boron soaking procedure for steam generators

    International Nuclear Information System (INIS)

    Ueno, T.; Tsuge, A.; Kawanishi, K.; Ochi, T.; Kadokami, E.

    1991-01-01

    An experimental study on boric acid penetration into tube to tube-support-plate crevices and Inter-Granular-Attack (hereinafter called 'IGA') cracks in crevices has been performed to obtain the optimum boric acid soaking procedure in operating steam generators with IGA. The penetration rate into the crevice is closely related to various parameters such as heat flux, crevice gap, and porosity of the sludge deposited in crevices. Two experimental crevice models were set up. One was of the packed crevice type; crevice gap is completely packed by sludge, and the other was of the open crevice type; crevice gap is not packed, but reduced by sludge. The porosity of the crevice varied from 100% open porosity to the highly sludge packed porosity of 10∼20%. The relation between heat flux and boric acid penetration rate of the packed crevice type was investigated. For the open crevice type, from the viewpoint that boric acid penetration into the dryout region produces no effects, tube wall superheat in the crevice was measured in order to obtain the dryout heat flux. And it was investigated the boron in IGA cracks using Ion Micro Analysis in order to confirm existence of an anticorrosive film in IGA propagation. The optimum reactor power for effective boric acid penetration onto the tube surface and into the IGA cracks within the tube to tube-support-plate crevice was found to be about a 5% and 30% power level, which are applicable to both the packed and open crevice type. (author)

  18. WWER Steam Generators Tubing Performance and Aging Management

    International Nuclear Information System (INIS)

    Trunov, Nikolay B.; Davidenko, Stanislav E.; Grigoriev, Vladimir A.; Popadchuk, Valery S.; Brykov, Sergery I.; Karzov, Georgy P.

    2006-01-01

    At WWER NPPs the horizontal steam generators (SGs), are used that differ in design concept from vertical SGs mostly used at western NPPs. Reliable operation of SG heat-exchanging tubes is the crucial worldwide problem for NPP of various types. According to the operation feedback the water chemistry is the governing factor affecting operability of SG tubing. The secondary side corrosion is considered to be the main mechanism of SG heat-exchanging tubes damage at WWER plants. To make the assessment of the tubing integrity the combination of pressure tests and eddy-current tests is used. Assessment of the tubing performance is an important part of SG life extension practice. The given paper deals with the description of the tube testing strategy and the approach to tube integrity assessment based on deterministic and probabilistic methods of fracture mechanics. Requirements for eddy-current test are given as well. Practice of condition monitoring and implementing the database on steam generators operation are presented. The approach to tubes plugging criteria is described. The research activities on corrosion mechanism studies and residual lifetime evaluation are mentioned. (authors)

  19. Optimum fuel allocation in parallel steam generator systems

    International Nuclear Information System (INIS)

    Bollettini, U.; Cangioli, E.; Cerri, G.; Rome Univ. 'La Sapienza'; Trento Univ.

    1991-01-01

    An optimization procedure was developed to allocate fuels into parallel steam generators. The procedure takes into account the level of performance deterioration connected with the loading history (fossil fuel allocation and maintenance) of each steam generator. The optimization objective function is the system hourly cost, overall steam demand being satisfied. Costs are due to fuel and electric power supply and to plant depreciation and maintenance as well. In order to easily updata the state of each steam generator, particular care was put in the general formulation of the steam production function by adopting a special efficiency-load curve description based on a deterioration scaling parameter. The influence of the characteristic time interval length on the optimum operation result is investigated. A special implementation of the method based on minimum cost paths is suggested

  20. A Receding Horizon Controller for the Steam Generator Water Level

    International Nuclear Information System (INIS)

    Na, Man Gyun; Lee, Yoon Joon

    2003-01-01

    In this work, the receding horizon control method was used to control the water level of nuclear steam generators and applied to two linear models and also a nonlinear model of steam generators. A receding horizon control method is to solve an optimization problem for finite future steps at current time and to implement the first optimal control input as the current control input. The procedure is then repeated at each subsequent instant. The dynamics of steam generators is very different according to power levels. The receding horizon controller is designed by using a reduced linear steam generator model fixed over a certain power range and applied to a Westinghouse-type (U-tube recirculating type) nuclear steam generator. The proposed controller designed at a fixed power level shows good performance for any other power level within this power range. The steam generator shows actually nonlinear characteristics. Therefore, the proposed algorithm is implemented for a nonlinear model of the nuclear steam generator to verify its real performance and also shows good responses

  1. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

    A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

  2. Babcock and Wilcox Canada steam generators past, present and future

    International Nuclear Information System (INIS)

    Smith, J.C.

    1998-01-01

    The steam generators in all of the domestic CANDU Plants, and most of the foreign CANDU plants, were supplied by Babcock and Wilcox Canada, either on their own or in co-operation with local manufacturers. More than 200 steam generators have been supplied. In addition, Babcock and Wilcox Canada has taken the technology which evolved out of the CANDU steam generators and has adapted the technology to supply of replacement steam generators for PWR's. There is enough history and operating experience, plus laboratory experience, to point to the future directions which will be taken in steam generator design. This paper documents the steam generators which have been supplied, the experience in operation and maintenance, what has worked and not worked, and how the design, materials, and operating and maintenance philosophy have evolved. The paper also looks at future requirements in the market, and the continuing research and product development going on at Babcock and Wilcox to address the future steam generator requirements. (author)

  3. Babcock and Wilcox Canada steam generators past, present and future

    Energy Technology Data Exchange (ETDEWEB)

    Smith, J.C. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada)

    1998-07-01

    The steam generators in all of the domestic CANDU Plants, and most of the foreign CANDU plants, were supplied by Babcock and Wilcox Canada, either on their own or in co-operation with local manufacturers. More than 200 steam generators have been supplied. In addition, Babcock and Wilcox Canada has taken the technology which evolved out of the CANDU steam generators and has adapted the technology to supply of replacement steam generators for PWR's. There is enough history and operating experience, plus laboratory experience, to point to the future directions which will be taken in steam generator design. This paper documents the steam generators which have been supplied, the experience in operation and maintenance, what has worked and not worked, and how the design, materials, and operating and maintenance philosophy have evolved. The paper also looks at future requirements in the market, and the continuing research and product development going on at Babcock and Wilcox to address the future steam generator requirements. (author)

  4. Study on steam separation in steam generators of a NPP with the WWER-440 reactors

    International Nuclear Information System (INIS)

    Dmitriev, A.I.; Kolzov, Yu.V.; Titov, V.F.; Dubrovin, A.V.; Ilyushin, V.F.; Volkov, A.P.

    1977-01-01

    The separation characteristics as well as the actual level position in steam generators with and without a submerged holy sheet have been determined at a WWER-440 reactor nuclear power plant. It has been shown, that without changing the design of steam generators their load at the WWER-440 reactor nuclear power plant can be increased by about 10%. In this case the vapour humidity does not exceed the permissible value equal to 0.25%. The submerged holy sheet considerably decreases load irregularity and swelling of the water-steam mixture layer

  5. Sludge cleaning in the steam generators: sludge Lancing e IBL

    International Nuclear Information System (INIS)

    Montoro, E.; Gonzalez, S.; Calderon, N.

    2013-01-01

    IBERDROLA Engineering and Construction has echoed the need for plants to remove oxide deposits (sludge) located on the secondary side, on the bottom plate and into the tube bundle steam steam generators. Therefore, and with its partner SAVAC SRA has developed a specific system consisting of applying a capillary water at very high pressure applied directly to the location of these oxides. (Author)

  6. Small leak shutdown, location, and behavior in LMFBR steam generators

    International Nuclear Information System (INIS)

    Sandusky, D.W.

    1976-01-01

    The paper summarizes an experimental study of small leaks tested under LMFBR steam generator conditions. Defected tubes were exposed to flowing sodium and steam. The observed behavior of the defected tubes is reported along with test results of shutdown methods. Leak location methods were investigated. Methods were identified to open plugged defects for helium leak testing and detect plugged leaks by nondestructive testing

  7. The progress of test and study for steam dryer in vertical steam generator

    International Nuclear Information System (INIS)

    Ding Xunshen

    1993-01-01

    Constructions, tests and test results are reviewed for three types of steam generator dryer that are concentric vertical corrugated separator, centrifugal conic separator and chevron separator. The last type is considered as the best one in comparison, which has been applied to Qinshan 300 MW steam generator. A number of pertinent remarks to draining scheme, hydraulic loss reduction, and conduct of test are given based on experiences

  8. Coupled Calculations in Helical Steam Generator: Validation on Legacy Data

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Yuan, Haomin [Argonne National Lab. (ANL), Argonne, IL (United States); Kraus, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Solberg, Jerome [Argonne National Lab. (ANL), Argonne, IL (United States); Ferencz, Robert M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-30

    The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental

  9. Coal fired steam generation for heavy oil recovery

    International Nuclear Information System (INIS)

    Firmin, K.

    1992-01-01

    In Alberta, some 21,000 m 3 /d of heavy oil and bitumen are produced by in-situ recovery methods involving steam injection. The steam generation requirement is met by standardized natural-gas-fired steam generators. While gas is in plentiful supply in Alberta and therefore competitively priced, significant gas price increases could occur in the future. A 1985 study investigating the alternatives to natural gas as a fuel for steam generation concluded that coal was the most economic alternative, as reserves of subbituminous coal are not only abundant in Alberta but also located relatively close to heavy oil and bitumen production areas. The environmental performance of coal is critical to its acceptance as an alternate fuel to natural gas, and proposed steam generator designs which could burn Alberta coal and control emissions satisfactorily are assessed. Considerations for ash removal, sulfur dioxide sorption, nitrogen oxides control, and particulate emission capture are also presented. A multi-stage slagging type of coal-fired combustor has been developed which is suitable for application with oilfield steam generators and is being commissioned for a demonstration project at the Cold Lake deposit. An economic study showed that the use of coal for steam generation in heavy oil in-situ projects in the Peace River and Cold Lake areas would be economic, compared to natural gas, at fuel price projections and design/cost premises for a project timing in the mid-1990s. 7 figs., 3 tabs

  10. Accident alarm equipment for steam generator, especially liquid sodium heated steam generator

    International Nuclear Information System (INIS)

    Matal, O.; Jung, J.; Banovec, J.

    1982-01-01

    The alarm equipment consists of a system of sensors mounted onto the steam generator and its accessories. Each of the sensors is used for a different accident characteristic, such as the flow of sodium, the acoustic spectrum, the concentration of hydrogen in sodium. The system of sensors is connected to the common accident alarm system. The equipment will not issue the alarm signal if it receives a message from only one sensor, only when the message is confirmed from other sensors. This excludes false alarm. (M.D.)

  11. Three Mile Island Nuclear Station steam generator chemical cleaning

    International Nuclear Information System (INIS)

    Hansen, C.A.

    1992-01-01

    The Three Mile Island-1 steam generators were chemically cleaned in 1991 by the B and W Nuclear Service Co. (BWNS). This secondary side cleaning was accomplished through application of the EPRI/SGOG (Electric Power Research Institute - Steam Generator Owners Group) chemical cleaning iron removal process, followed by sludge lancing. BWNS also performed on-line corrosion monitoring. Corrosion of key steam generator materials was low, and well within established limits. Liquid waste, subsequently processed by BWNS was less than expected. 7 tabs

  12. A reflux capsule steam generator for sodium cooled reactors

    International Nuclear Information System (INIS)

    Lantz, E.

    Pressurized water reactor plants at numerous sites have sustained significant leakage through their steam generators. The consequent shutdowns for repairs and replacements have damaged their economics. This experience suggests that if steam generators for liquid metal fast breeder reactors (LMFBR's) continue to be built as presently designed some of them will have similar problems. Because of their larger capital investment, the consequent damage to the economics of LMFBR's could be more serious. Reflux capsules provide a way to separate sodium from water and to reduce thermal stresses in steam generators for sodium cooled reactors. Their use would also eliminate the need for a primary heat exchanger and a secondary sodium loop pump. (author)

  13. 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

  14. Thermo hydrodynamical analyses of steam generator of nuclear power plant

    International Nuclear Information System (INIS)

    Petelin, S.; Gregoric, M.

    1984-01-01

    SMUP computer code for stationary model of a U-tube steam generator of a PWR nuclear power plant was developed. feed water flow can enter through main and auxiliary path. The computer code is based on the one dimensional mathematical model. Among the results that give an insight into physical processes along the tubes of steam generator are distribution of temperatures, water qualities, heat transfer rates. Parametric analysis permits conclusion on advantage of each design solution regarding heat transfer effects and safety of steam generator. (author)

  15. Lower head integrity under steam explosion loads

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Yuen, W.W.; Angelini, S.; Freeman, K.; Chen, X.; Salmassi, T. [Center for Risk Studies and Safety, Univ. of California, Santa Barbara, CA (United States); Sienicki, J.J.

    1998-01-01

    Lower head integrity under steam explosion loads in an AP600-like reactor design is considered. The assessment is the second part of an evaluation of the in-vessel retention idea as a severe accident management concept, the first part (DOE/ID-10460) dealing with thermal loads. The assessment is conducted in terms of the Risk Oriented Accident Analysis Methodology (ROAAM), and includes the comprehensive evaluation of all relevant severe accident scenarios, melt conditions and timing of release from the core region, fully 3D mixing and explosion wave dynamics, and lower head fragility under local, dynamic loading. All of these factors and brought together in a ROAAM Probabilistic Framework to evaluate failure likelihood. The conclusion is that failure is `physically unreasonable`. (author)

  16. Soviet steam generator technology: fossil fuel and nuclear power plants

    International Nuclear Information System (INIS)

    Rosengaus, J.

    1987-01-01

    In the Soviet Union, particular operational requirements, coupled with a centralized planning system adopted in the 1920s, have led to a current technology which differs in significant ways from its counterparts elsewhere in the would and particularly in the United States. However, the monograph has a broader value in that it traces the development of steam generators in response to the industrial requirements of a major nation dealing with the global energy situation. Specifically, it shows how Soviet steam generator technology evolved as a result of changing industrial requirements, fuel availability, and national fuel utilization policy. The monograph begins with a brief technical introduction focusing on steam-turbine power plants, and includes a discussion of the Soviet Union's regional power supply (GRES) networks and heat and power plant (TETs) systems. TETs may be described as large central co-generating stations which, in addition to electricity, provide heat in the form of steam and hot water. Plants of this type are a common feature of the USSR today. The adoption of these cogeneration units as a matter of national policy has had a central influence on Soviet steam generator technology which can be traced throughout the monograph. The six chapters contain: a short history of steam generators in the USSR; steam generator design and manufacture in the USSR; boiler and furnace assemblies for fossil fuel-fired power stations; auxiliary components; steam generators in nuclear power plants; and the current status of the Soviet steam generator industry. Chapters have been abstracted separately. A glossary is included containing abbreviations and acronyms of USSR organizations. 26 references

  17. Modelling of steam condensation in the primary flow channel of a gas-heated steam generator

    International Nuclear Information System (INIS)

    Kawamura, H.; Meister, G.

    1982-10-01

    A new simulation code has been developed for the analysis of steam ingress accidents in high temperatures reactors which evaluates the heat transfer in a steam generator headed by a mixture of helium and water steam. Special emphasis is laid on the analysis of steam condensation in the primary circuit of the steam generator. The code takes wall and bulk condensation into account. A new method is proposed to describe the entrainment of water droplets in the primary gas flow. Some typical results are given. Steam condensation in the primary channel may have a significant effect on temperature distributions. The effect on the heat transferred by the steam generator, however, is found to be not so prominent as might be expected. The reason is discussed. A simplified code will also be described, which gives results with reasonable accuracy within much shorter execution times. This code may be used as a program module in a program simulating the total primary circuit of a high temperature reactor. (orig.) [de

  18. PACTEL: Experiments on the behaviour of the new horizontal steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kouhia, J.; Riikonen, V.; Purhonen, H. [VTT Energy, Lappeenranta (Finland)

    1995-12-31

    Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

  19. PACTEL: Experiments on the behaviour of the new horizontal steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kouhia, J; Riikonen, V; Purhonen, H [VTT Energy, Lappeenranta (Finland)

    1996-12-31

    Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

  20. Analysis of induced steam generator tube rupture using MAAP 4.0

    International Nuclear Information System (INIS)

    Kenton, M.; Epstein, M.; Henry, R.E.; Paik, C.; Fuller, E.

    1996-01-01

    The nuclear industry has initiated a program of Steam Generator Degradation Specific Management (SGDSM) to cope with the various types of corrosion that have been observed in pressurized water reactor (PWR) steam generators. In parallel, the U.S. Nuclear Regulatory Commission is promulgating revised rules on steam generator tube integrity. To support these efforts, the Electric Power Research Institute has sponsored calculations with the MAAP 4 code. The principal objective of these calculations is to estimate the peak temperatures experienced by the steam generator tubes during high-pressure severe accidents. These results are used to evaluate the potential for degraded tubes to leak or rupture. Attention was focused on station blackout (SBO) accidents with loss of turbine-driven auxiliary feedwater because these generally result in the greatest threat to the tubes

  1. Prototype steam generator test at SCTI/ETEC. Acoustic program test plan

    International Nuclear Information System (INIS)

    Greene, D.A.; Thiele, A.; Claytor, T.N.

    1981-10-01

    This document is an integrated test plan covering programs at General Electric (ARSD), Rockwell International (RI) and Argonne National Laboratory (CT). It provides an overview of the acoustic leak detection test program which will be completed in conjunction with the prototype LMFBR steam generator at the Energy Technology Engineering Laboratory. The steam generator is installed in the Sodium Components Test Installation (SCTI). Two acoustic detection systems will be used during the test program, a low frequency system developed by GE-ARSD (GAAD system) and a high frequency system developed by RI-AI (HALD system). These systems will be used to acquire data on background noise during the thermal-hydraulic test program. Injection devices were installed during fabrication of the prototype steam generator to provide localized noise sources in the active region of the tube bundle. These injectors will be operated during the steam generator test program, and it will be shown that they are detected by the acoustic systems

  2. Steam generator secondary pH during a steam generator tube rupture

    International Nuclear Information System (INIS)

    Adams, J.P.; Peterson, E.S.

    1991-12-01

    The Nuclear Regulatory Commission requires utilizes to determine the response of a pressurized water reactor to a steam generator tube rupture (SGTR) as part of the safety analysis for the plant. The SGTR analysis includes assumptions regarding the partitioning of iodine between liquid and vapor in steam generator secondary. Experimental studies have determined that the partitioning of iodine in water is very sensitive to the pH. Based on this experimental evidence, the NRC requested the INEL to perform an analytical assessment of secondary coolant system (SCS) pH during an SGTR. Design basis thermal and hydraulic calculations were used together with industry standard chemistry guidelines to determine the SCS chemical concentrations during an SGTR. These were used as input to the Facility for Analysis of Chemical Thermodynamics computer system to calculate the equilibrium pH in the SCS at various discrete time during an SGTR. The results of this analysis indicate that the SCS pH decreases from the initial value of 8.8 to approximately 6.5 by the end of the transient, independent of PWR design

  3. Steam generator life cycle management - B and W perspective

    International Nuclear Information System (INIS)

    Dhar, D.; Fluit, S.; Millman, J.

    2006-01-01

    This paper is an effort towards the B and W perspective about the effective life cycle management (LCM) of the CANDU Steam Generators (SGs) based on the identification of active and plausible degradation mechanisms for various SG components and the need to achieve a safe and economic operating interval for the station. The objective of this paper is to provide the long-term effective strategy for inspections, maintenance and design modifications as necessary for the safe and reliable operation of the SGs during the plant design life. The derived activities of this strategy need to be integrated with the station outage specific work scope plan for an effective life cycle management. The technical basis for these activities is based on the review of previous field inspection records, maintenance work and modifications at the station and operational experience (OPEX) from other CANDU steam generators with similar design. These activities need to be performed in order to ensure that the SGs perform within an acceptable level of safety and reliability as per the licensing bases, while optimizing station production and cost effectiveness. (author)

  4. Study of tritium permeation through Peach Bottom Steam Generator tubes

    International Nuclear Information System (INIS)

    Yang, L.; Baugh, W.A.; Baldwin, N.L.

    1977-06-01

    The report describes the equipment developed, samples tested, procedures used, and results obtained in the tritium permeation tests conducted on steam generator tubing samples which were removed from the Peach Bottom Unit No. 1 reactor

  5. Improvements during fabrication of the Spanish steam generators

    International Nuclear Information System (INIS)

    Alvarez Miranda, A.

    1994-01-01

    The ENSA company is manufacturing the tubes of the steam generator in NP Asco 1. The improvements of fabrication and production are broken down into 3 chapters: - Mechanizing process - Welding process - Clean area activities

  6. Horizontal steam generator PGV-1000 thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ubra, O. [Skoda Company, Prague (Switzerland); Doubek, M. [Czech Technical Univ., Prague (Switzerland)

    1995-12-31

    A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

  7. Horizontal steam generator PGV-1000 thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Ubra, O.; Doubek, M.

    1995-01-01

    A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.)

  8. Inverted Steam Generators for Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Matal, Oldřich; Šimo, Tomáš; Matal, Oldřich Jr.

    2013-01-01

    Conclusions: Two inverted steam generators of the Czech industry provenience have still been in successful operation with no water into sodium leaks at BOR 60 (RIAR Dimitrovgrad, Russian Federation). Micromodular inverted steam generator (MMISG) since 1981 and modular inverted steam generator (MISG) since 1991. In the framework of the CP ESFR project predesign studies of 100 MW (thermal) ISG modules were performed with the consideration of MMISG and MISG design, operational and safety benefits and experience. Development of material and technology for sodium heated steam generators components reflecting contemporary domestic industrial conditions in the Czech Republic was restarted in the years 2003 to 2004 and supported in the years 2008 to 2011 by the European CP ESFR project and by the Ministry of Industry and Trade of the Czech Republic

  9. Horizontal steam generator PGV-1000 thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ubra, O [Skoda Company, Prague (Switzerland); Doubek, M [Czech Technical Univ., Prague (Switzerland)

    1996-12-31

    A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

  10. Chemical-cleaning process evaluation: Westinghouse steam generators. Final report

    International Nuclear Information System (INIS)

    Cleary, W.F.; Gockley, G.B.

    1983-04-01

    The Steam Generator Owners Group (SGOG)/Electric Power Research Institute (EPRI) Steam Generator Secondary Side Chemical Cleaning Program, under develpment since 1978, has resulted in a generic process for the removal of accumulated corrosion products and tube deposits in the tube support plate crevices. The SGOG/EPRI Project S150-3 was established to obtain an evaluation of the generic process in regard to its applicability to Westinghouse steam generators. The results of the evaluation form the basis for recommendations for transferring the generic process to a plant specific application and identify chemical cleaning corrosion guidelines for the materials in Westinghouse Steam Generators. The results of the evaluation, recommendations for plant-specific applications and corrosion guidelines for chemical cleaning are presented in this report

  11. Nuclear steam generator sludge lance method and apparatus

    International Nuclear Information System (INIS)

    Shirey, R.A.; Murray, D.E.

    1991-01-01

    This paper describes a sludge lancing system for removing sludge deposits from an interior region of a steam generator. It comprises: a peripheral fluid injection means for injecting a fluid at a high pressure about a periphery of the steam generator, the peripheral fluid injection means comprising at least one elongated fluid conduit, at least one injection nozzle and a joint positioned at a predetermined point along the elongated fluid conduit for permitting the peripheral fluid injection means to bend to a predetermined angle at the joint within the steam generator; a reciprocable fluid injection means for injecting a fluid at a high pressure toward the sludge deposits and dislodging the sludge deposits; and a supporting means positioned within the interior of the steam generator for supporting the reciprocable fluid injection means throughout the reciprocation of the reciprocable fluid injection means

  12. Improvements in steam cycle electric power generating plants

    International Nuclear Information System (INIS)

    Bienvenu, Claude.

    1973-01-01

    The invention relates to a steam cycle electric energy generating plants of the type comprising a fossil or nuclear fuel boiler for generating steam and a turbo alternator group, the turbine of which is fed by the boiler steam. The improvement is characterized in that use is made of a second energy generating group in which a fluid (e.g. ammoniac) undergoes a condensation cycle the heat source of said cycle being obtained through a direct or indirect heat exchange with a portion of the boiler generated steam whereby it is possible without overloading the turbo-alternator group, to accomodate any increase of the boiler power resulting from the use of another fuel while maintaining a maximum energy output. This can be applied to electric power stations [fr

  13. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  14. Modular sludge collection system for a nuclear steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.; Bein, J.D.; Powasaki, F.S.

    1986-01-01

    A sludge collection system is described for a vertically oriented nuclear steam generator wherein vapors produced in the steam generator pass through means for separating entrained liquid from the vapor prior to the vapor being discharged from the steam generator. The sludge collection system comprises: an upwardly open chamber for collecting the separated liquid and feedwater entering the steam generator; upwardly open sludge collecting containers positioned within the chamber, wherein each of the containers includes a top rim encompassing an opening leading to the interior of each container; generally flat, perforated covers, each of the covers being positioned over one of the openings such that a gap is formed between the cover and the adjacent top rim; sludge agitating means on at least one of the containers; and sludge removal means on at least one of the containers

  15. Technical development and its application on steam generator replacement

    International Nuclear Information System (INIS)

    Morita, Sadahiko; Hanzawa, Katsumi; Sato, Hajime; Kannoto, Yasuo.

    1995-01-01

    Twenty-two PWR nuclear power plants are now under commercial operation in Japan. Eight of these plants are scheduled to have their steam generators replaced by up-graded units as a social responsibility for improved reliability, economy and easier maintenance. To carry out steam generator replacement, main coolant pipe cutting and restoration techniques, remote controlled welding machines and other remote controlled equipment, templating techniques with which the new steam generator primary nozzles will fit the existing primary pipes correctly were developed. An adequate training program was carried out to establish these techniques and they were then applied in replacement work on site. The steam generators of the three plants were replaced completely in 1994. These newly developed techniques are to be applied in upcoming plants and replaced plants will be much reliable. (author)

  16. Evaluation of Oconee steam-generator debris. Final report

    International Nuclear Information System (INIS)

    Rigdon, M.A.; Rubright, M.M.; Sarver, L.W.

    1981-10-01

    Pieces of debris were observed near damaged tubes at the 14th support plate elevation in the Oconee 1-B steam generator. A project was initiated to evaluate the physical and chemical nature of the debris, to identify its source, and to determine its role in tube damage at this elevation. Various laboratory techniques were used to characterize several debris and mill scale samples. Data from these samples were then compared with each other and with literature data. It was concluded that seven of eight debris samples were probably formed in the steam generator. Six of these samples were probably formed by high temperature aqueous corrosion early in the life of the steam generator. The seventh sample was probably formed by the deposition and spalling of magnetite on the Inconel steam generator tubes. None of the debris samples resembled any of the mill scale samples

  17. Overview of the United States steam generator development programs

    Energy Technology Data Exchange (ETDEWEB)

    Kaspar, P W; Lowe, P A

    1975-07-01

    The LMFBR steam generator development program of the USA was initiated to support the development of reliable designs and meaningful performance data for these critical components. Since the steam generators include the structural boundary between heated sodium and water, the consequences of small flaws in the materials that form the boundary are significant. Successful development and demonstration of commercial LMFBR power plants requires the consideration of many factors in addition to the design, construction and operation of a particular plant. Additional factors which must be assessed include: economics, reliability, safety, environment, operability, maintainability and conservation of the resources. In terms of the steam generator these items led to the selection of a single wall tube design using a forced recirculating system for the present Clinch River Breeder Reactor. There are strong economic incentives to use a once-through steam generating system in future designs.

  18. Steam and sodium leak simulation in a fluidized-bed steam generator

    International Nuclear Information System (INIS)

    Vaux, W.G.; Keeton, A.R.; Keairns, D.L.

    1977-01-01

    A fluidized-bed steam generator for the liquid metal fast breeder reactor enhances plant availability and minimizes the probability of a water/sodium reaction. An experimental test program was conceived to assess design criteria and fluidized-bed operation under conditions of water, steam, and sodium leaks. Sodium, steam, and water were leaked into helium-fluidized beds of metal and ceramic particles at 900 F. Test results show the effects of leaks on the heat transfer coefficient, quality of fluidization, leak detection, and cleanup procedures

  19. Steam Generator Owners Group PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Green, S.J.

    1985-01-01

    In 1981 the Steam Generator Owners Group (SGOG), a group of domestic and foreign pressurized water reactor (PWR) owners, developed and issued the PWR secondary water chemistry guidelines. The guidelines were prepared in response to the growing recognition that a majority of the problems causing reduced steam generator reliability (e.g., denting, wasteage, pitting, etc.) were related to secondary (steam) side water purity. The guidelines were subsequently issued as an Electric Power Research Institute (EPRI) report. In 1984 they were revised to reflect industry experience in adopting the original issuance and to incorporate new information on causes of corrosion damage. The guidelines have been endorsed and their adoption recommended by the SGOG

  20. Analysis and design of flow limiter used in steam generator

    International Nuclear Information System (INIS)

    Liu Shixun; Gao Yongjun

    1995-10-01

    Flow limiter is an important safety component of PWR steam generator. It can limit the blowdown rate of steam generator inventory in case of the main steam pipeline breaks, so that the rate of the primary coolant temperature reduction can be slowed down in order to prevent fuel element from burn-out. The venturi type flow limiter is analysed, its flow characteristics are delineated, physical and mathematical models defined; the detail mathematical derivation provided. The research lays down a theoretic basis for flow limiter design. The governing equations and formulas given can be directly applied to computer analysis of the flow limiter. (3 refs., 3 figs.)

  1. Estimating probable flaw distributions in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gorman, J.A.; Turner, A.P.L.

    1997-01-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses

  2. Dynamic and control of a once through steam generator

    International Nuclear Information System (INIS)

    Gomes, Arivaldo Vicente

    1979-01-01

    This paper presents a non linear distributed parameter model for the dynamics and feedback control of a large countercurrent heat exchanger used as a once through steam generator for a breeder reactor power plant. A convergent, implicit method has been developed to solve simultaneously the equations of conservation of mass, momentum and energy. The model, applicable to heat exchanger systems in general, has been used specifically to study the performance of a once-through steam generator with respect to its load following ability and stability of throttle steam temperature and pressure. (author)

  3. Design and construction of a steam generator with feedback

    International Nuclear Information System (INIS)

    Camargo, Camila C.; Placco, Guilherme M.; Guimaraes, Lamartine N.F.

    2013-01-01

    The EARTH project aims to develop technologies to design and build systems that generate electricity in space, using microreactors. One of the activities within the TERRA project aims to build a closed thermal cycle Rankine type in order to test a Tesla turbine type. The objective of this work is to design and build a steam generator with feedback, which should ensure a satisfactory range of steam supply, security system, feedback system and heating system

  4. Fault tolerant control for steam generators in nuclear power plant

    International Nuclear Information System (INIS)

    Deng Zhihong; Shi Xiaocheng; Xia Guoqing; Fu Mingyu

    2010-01-01

    Based on the nonlinear system with stochastic noise, a bank of extended Kalman filters is used to estimate the state of sensors. It can real-time detect and isolate the single sensor fault, and reconstruct the sensor output to keep steam generator water level stable. The simulation results show that the methodology of employing a bank of extended Kalman filters for steam generator fault tolerant control design is feasible. (authors)

  5. Automated Diagnosis and Classification of Steam Generator Tube Defects

    International Nuclear Information System (INIS)

    Garcia, Gabe V.

    2004-01-01

    A major cause of failure in nuclear steam generators is tube degradation. Tube defects are divided into seven categories, one of which is intergranular attack/stress corrosion cracking (IGA/SCC). Defects of this type usually begin on the outer surface of the tubes and propagate both inward and laterally. In many cases these defects occur at or near the tube support plates. Several different methods exist for the nondestructive evaluation of nuclear steam generator tubes for defect characterization

  6. SG (steam generators) reliability project builds on Owners Group successes

    International Nuclear Information System (INIS)

    Green, S.

    1988-01-01

    In 1987, a five-year Steam Generator Reliability Project was established at the Electric Power Research Institute (EPRI) to deal with outstanding issues, following on from work initiated by the previous utility industry groupings (Steam Generator Owners Groups I and II). The work done by these groups is discussed and a listing of the major objectives of the new project is provided. (U.K.)

  7. Dynamic instability forecasting for through-out sodium steam generators

    International Nuclear Information System (INIS)

    Aleksandrov, V.V.; Rassokhin, N.G.

    1985-01-01

    Simplified technique for determining boundaries of dynamic instability of through-out sodium steam generators is presented. The technique is based on the application of autoresonance concept to autooscillating model of dynamic instability of a steam-generating channel. Estimated model parameters and basic investigational results for different conditions are given. Assessment is performed according to the instability degree. Use of the technique is effective for multiversion studying of SG design at early designing stages

  8. Development of data management system for steam generator inspection

    International Nuclear Information System (INIS)

    Jung, Yong Moo; Im, Chang Jae; Lee, Yoon Sang; Kang, Soon Joo; An, Jong Kwan

    1994-06-01

    The data communications environment for transferring Nuclear Power Plant Steam Generator Eddy Current testing data was investigated and after connecting LAN to Hinet-F network, the remote data transfer with the speed of 56 kbps was tested successfully. Data management system for Steam Generator Eddy current testing was also developed by using HP-UX, RMB (Rock Mountain Basic) 21 figs, 13 tabs, 5 refs. (Author)

  9. Development of data management system for steam generator inspection

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yong Moo; Im, Chang Jae; Lee, Yoon Sang; Kang, Soon Joo; An, Jong Kwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    The data communications environment for transferring Nuclear Power Plant Steam Generator Eddy Current testing data was investigated and after connecting LAN to Hinet-F network, the remote data transfer with the speed of 56 kbps was tested successfully. Data management system for Steam Generator Eddy current testing was also developed by using HP-UX, RMB (Rock Mountain Basic) 21 figs, 13 tabs, 5 refs. (Author).

  10. Steam generators in indirect-cycle water-cooled reactors

    International Nuclear Information System (INIS)

    Fajeau, M.

    1976-01-01

    In the indirect cycle water-cooled nuclear reactors, the steam generators are placed between the primary circuit and the turbine. They act both as an energy transmitter and as a leaktigh barrier against fission or corrosion products. Their study is thus very important from a performance and reliability point of view. Two main types are presented here: the U-tube and the once-through steam generators [fr

  11. NRC concerns about steam generator tube U-bend failures

    International Nuclear Information System (INIS)

    Dillon, R.L.

    1981-01-01

    This paper concerns itself with genralized NRC regulatory policy regarding SGT failures and staff reports and opinions which may tend to influence the developing policy specific to U-bend failures. The most significant analysis at hand in predicting NRC policy on SGT U-bend failures is Marsh's Evaluation of Steam Generator Tube Rupture Events. Marsh sets out to describe and analyze the five steam generator tube ruptures that are known to NRC. All have occurred in the period 1975 to 1980

  12. Some engineering aspects of the steam generator system for the United States LMFBR demonstration plant

    International Nuclear Information System (INIS)

    Tippets, F.E.

    1975-01-01

    This paper describes the main design features of the steam generator system for the Clinch River Breeder Reactor Plant and the engineering approach being employed for some of the critical elements of this system, including in particular the sodium-steam/water boundary, the efforts to have this boundary be of highest integrity, and the system features to safely accommodate any failure of the boundary. (author)

  13. Some engineering aspects of the steam generator system for the United States LMFBR demonstration plant

    Energy Technology Data Exchange (ETDEWEB)

    Tippets, F E

    1975-07-01

    This paper describes the main design features of the steam generator system for the Clinch River Breeder Reactor Plant and the engineering approach being employed for some of the critical elements of this system, including in particular the sodium-steam/water boundary, the efforts to have this boundary be of highest integrity, and the system features to safely accommodate any failure of the boundary. (author)

  14. Acoustic detection for water/steam leak from a tube of LMFBR steam generator

    International Nuclear Information System (INIS)

    Sonoda, Masataka; Shindo, Yoshihisa

    1989-01-01

    Acoustic leak detector is useful for detecting more quickly intermediate leak than the existing hydrogen detector and is available for identification of leak location on the accident of water/steam leak from a tube of LMFBR steam generator. This paper presents the overview of HALD (High frequency Acoustics Leak Detection) system, which is more sensitive for leak detection and lower cost of equipment for identification of leak location than a low frequency type detector. (author)

  15. Steam generator replacement at the Obrigheim nuclear power station

    International Nuclear Information System (INIS)

    Pickel, E.; Schenk, H.; Huemmler, A.

    1984-01-01

    The Obrigheim Nuclear Power Station (KWO) is equipped with a dual-loop pressurized water reactor of 345 MW electric power; it was built by Siemens in the period 1965 to 1968. By the end of 1983, KWO had produced some 35 billion kWh in 109,000 hours of operation. Repeated leaks in the heater tubes of the two steam generators had occurred since 1971. Both steam generators were replaced in the course of the 1983 annual revision. Kraftwerk Union AG (KWU) was commissioned to plant and carry out the replacement work. Despite the leakages the steam generators had been run safely and reliably over a period of 14 years until their replacement. Replacing the steam generators was completed within twelve weeks. In addition to the KWO staff and the supervising crew of KWU, some 400 external fitters were employed on the job at peak work-load periods. For the revision of the whole plant, work on the emergency systems and replacement of the steam generators a maximum number of approx. 900 external fitters were employed in the plant in addition to some 250 members of the plant crew. The exposure dose of the personnel sustained in the course of the steam generator replacement was 690 man-rem, which was clearly below previous estimates. (orig.) [de

  16. Highly Flexible and Efficient Solar Steam Generation Device.

    Science.gov (United States)

    Chen, Chaoji; Li, Yiju; Song, Jianwei; Yang, Zhi; Kuang, Yudi; Hitz, Emily; Jia, Chao; Gong, Amy; Jiang, Feng; Zhu, J Y; Yang, Bao; Xie, Jia; Hu, Liangbing

    2017-08-01

    Solar steam generation with subsequent steam recondensation has been regarded as one of the most promising techniques to utilize the abundant solar energy and sea water or other unpurified water through water purification, desalination, and distillation. Although tremendous efforts have been dedicated to developing high-efficiency solar steam generation devices, challenges remain in terms of the relatively low efficiency, complicated fabrications, high cost, and inability to scale up. Here, inspired by the water transpiration behavior of trees, the use of carbon nanotube (CNT)-modified flexible wood membrane (F-Wood/CNTs) is demonstrated as a flexible, portable, recyclable, and efficient solar steam generation device for low-cost and scalable solar steam generation applications. Benefitting from the unique structural merits of the F-Wood/CNTs membrane-a black CNT-coated hair-like surface with excellent light absorbability, wood matrix with low thermal conductivity, hierarchical micro- and nanochannels for water pumping and escaping, solar steam generation device based on the F-Wood/CNTs membrane demonstrates a high efficiency of 81% at 10 kW cm -2 , representing one of the highest values ever-reported. The nature-inspired design concept in this study is straightforward and easily scalable, representing one of the most promising solutions for renewable and portable solar energy generation and other related phase-change applications. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  17. Steam generator replacement: a story of continuous improvement

    International Nuclear Information System (INIS)

    Sills, M.S.; Wilkerson, R.

    2009-01-01

    This paper provides a review of the history of steam generator replacement in the US focusing on the last five years. From the early replacements in the 1980s, there have been major technology improvements resulting in dramatically shorter outages and reduced radiological exposure for workers. Even though the changes for the last five years have been less dramatic, the improvement trend continues. No two steam generator replacement (SGR) projects are the same and there are some major differences including; the access path for the components to containment (is a construction opening in containment required), type of containment, number of steam generators, one piece or two piece replacement, plant type (Westinghouse, CE or B and W) and plant layout. These differences along with other variables such as delays due to plant operations and other activities not related to the steam generator replacement make analysis of performance data difficult. However, trends in outage performance and owner expectations can be identified. How far this trend will go is also discussed. Along with the trend of improved performance, there is also a significant variation in performance. Some of the contributors to this variation are identified. This paper addresses what is required for a successful outage, meeting the increasing expectations and setting new records. The authors will discuss various factors that contribute to the success of a steam generator replacement. These factors include technical issues and, equally important, organizational interface and the role the customer plays. Recommendations are provided for planning a successful steam generator replacement outage. (author)

  18. Upgraded Steam Generator Lancing System for Uljin NPP no.2

    International Nuclear Information System (INIS)

    Kim, Seok Tae; Jeong, Woo Tae; Hong, Sung Yull

    2005-01-01

    KEPRI(Korea Electric Power Research Institute) has developed various types of steam generator lancing systems since 1998. In this paper, we introduce a new lancing system with new improvements from the previous steam generator lancing system for Uljin NPP #2(nuclear power plant) constructed by KEPRI. The previous lancing system is registered as KALANS R -II and was developed for System-80 type steam generators. The previous lancing system was applied to Uljin unit #3 and it lowered radiation exposure of operators in comparison to manually operated lancing systems. And it effectively removed sludge accumulated around kidney bean zone in the Uljin unit #3 steam generators. But the previous lancing system could only clean partially the steam generators of Uljin unit #4. This was because the rail of the previous lancing system interfered with a part of the steam generator. Therefore we developed a new lancing system that can solve the interference problem. This new lancing system was upgraded from the previous lancing system. Also, a new lancing system for System-80 S/G will be introduced in this paper

  19. Draining down of a nuclear steam generating system

    International Nuclear Information System (INIS)

    Jawor, J.C.

    1987-01-01

    The method is described of draining down contained reactor-coolant water from the inverted vertical U-tubes of a vertical-type steam generator in which the upper, inverted U-shaped ends of the tubes are closed and the lower ends thereof are open. The steam generator is part of a nuclear powered steam generating system wherein the reactor coolant water is normally circulated from and back into the reactor via a loop comprising the steam generator and inlet and outlet conduits connected to the lower end of the steam generator. The method comprises continuously introducing a gas which is inert to the system and which is under pressure above atmospheric pressure into at least one of the downwardly facing open ends of each of the U-tubes from below the tube sheet in which the open ends of the U-tubes are mounted adjacent the lower end of the steam generator, while permitting the water to flow out from the open ends of the U-tubes

  20. Analysis of Decay Heat Removal by Natural Convection in LMR with a Combined Steam Generator

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Eoh, Jae Hyuk; Han, Ji Woong; Lee, Tae Ho

    2011-01-01

    Liquid metal reactors (LMRs) conventionally employ an intermediate heat transport system (IHTS) to protect the nuclear core during a sodium-water reaction (SWR) event. However these SWR-related components increase plant construction costs. In order to eliminate the need for an IHTS, a combined steam generator, which is an integrated heat exchanger of a steam generator and intermediate heat exchanger (IHX), was proposed by the Korea Atomic Energy Research Institute (KAERI). The objective of this work is to analyze the natural circulation heat removal capability of the rector system using a combined steam generator. As a means of decay heat removal, a normal heat transport path is composed of a primary sodium system, intermediate lead-bismuth circuit combined with SG and steam/water system. This paper presents the results of the possible temperature and natural circulation flows in all circuits during a steady state for a given reactor power level varied as a function of time

  1. Hydrogen-oxygen steam generator applications for increasing the efficiency, maneuverability and reliability of power production

    Science.gov (United States)

    Schastlivtsev, A. I.; Borzenko, V. I.

    2017-11-01

    The comparative feasibility study of the energy storage technologies showed good applicability of hydrogen-oxygen steam generators (HOSG) based energy storage systems with large-scale hydrogen production. The developed scheme solutions for the use of HOSGs for thermal power (TPP) and nuclear power plants (NPP), and the feasibility analysis that have been carried out have shown that their use makes it possible to increase the maneuverability of steam turbines and provide backup power supply in the event of failure of the main steam generating equipment. The main design solutions for the integration of hydrogen-oxygen steam generators into the main power equipment of TPPs and NPPs, as well as their optimal operation modes, are considered.

  2. High-definition radiography of tube-to-tubesheet welds of steam generator of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Venkatraman, B.; Sethi, V.K.; Jayakumar, T.; Raj, B.

    1995-01-01

    In the steam generator of the Prototype Fast Breeder Reactor (PFBR), steam is generated by the transfer of heat from secondary sodium to water. Due to the inherent dangers of sodium-water reaction, the integrity of weld joints separating sodium and water/steam is of paramount importance. This is particularly true and very important for the tube-to-tubesheet joints. This paper discusses the use of projective magnification technique by microfocal radiography for the quality evaluation and optimisation of the welding parameters of such small tube-to-tubesheet welds of the steam generator of PFBR. (author)

  3. Three-dimensional modeling of nuclear steam generator

    International Nuclear Information System (INIS)

    Bogdan, Z.; Afgan, N.

    1985-01-01

    In this paper mathematical model for steady-state simulation of thermodynamic and hydraulic behaviour of U-tube nuclear steam generator is described. The model predicts three-dimensional distribution of temperatures, pressures, steam qualities and velocities in the steam generator secondary loop. In this analysis homogeneous two phase flow model is utilized. Foe purpose of the computer implementation of the mathematical model, a special flow distribution code NUGEN was developed. Calculations are performed with the input data and geometrical characteristics related to the D-4 (westinghouse) model of U-tube nuclear steam generator built in Krsko, operating under 100% load conditions. Results are shown in diagrams giving spatial distribution of pertinent variables in the secondary loop. (author)

  4. A fault detection and diagnosis in a PWR steam generator

    International Nuclear Information System (INIS)

    Park, Seung Yub

    1991-01-01

    The purpose of this study is to develop a fault detection and diagnosis scheme that can monitor process fault and instrument fault of a steam generator. The suggested scheme consists of a Kalman filter and two bias estimators. Method of detecting process and instrument fault in a steam generator uses the mean test on the residual sequence of Kalman filter, designed for the unfailed system, to make a fault decision. Once a fault is detected, two bias estimators are driven to estimate the fault and to discriminate process fault and instrument fault. In case of process fault, the fault diagnosis of outlet temperature, feed-water heater and main steam control valve is considered. In instrument fault, the fault diagnosis of steam generator's three instruments is considered. Computer simulation tests show that on-line prompt fault detection and diagnosis can be performed very successfully.(Author)

  5. Steam generator deposit control program assessment at Comanche Peak

    International Nuclear Information System (INIS)

    Stevens, J.; Fellers, B.; Orbon, S.

    2002-01-01

    Comanche Peak has employed a variety of methods to assess the effectiveness of the deposit control program. These include typical methods such as an extensive visual inspection program and detailed corrosion product analysis and trending. In addition, a recently pioneered technique, low frequency eddy current profile analysis (LFEC) has been utilized. LFEC provides a visual mapping of the magnetite deposit profile of the steam generator. Analysis of the LFEC results not only provides general area deposition rates, but can also provide local deposition patterns, which is indicative of steam generator performance. Other techniques utilized include trending of steam pressure, steam generator hideout-return, and flow assisted corrosion (FAC) results. The sum of this information provides a comprehensive assessment of the deposit control program effectiveness and the condition of the steam generator. It also provides important diagnostic and predictive information relative to steam generator life management and mitigative strategies, such as special cleaning procedures. This paper discusses the techniques employed by Comanche Peak Chemistry to monitor the effectiveness of the deposit control program and describes how this information is used in strategic planning. (authors)

  6. Primary separator replacement for Bruce Unit 8 steam generators

    International Nuclear Information System (INIS)

    Roy, S.B.; Mewdell, C.G.; Schneider, W.G.

    2000-01-01

    During a scheduled maintenance outage of Bruce Unit 8 in 1998, it was discovered that the majority of the original primary steam separators were damaged in two steam generators. The Bruce B steam generators are equipped with GXP type primary cyclone separators of B and W supply. There were localized perforations in the upper part of the separators and large areas of generalized wall thinning. The degradation was indicative of a flow related erosion corrosion mechanism. Although the unit- restart was justified, it was obvious that corrective actions would be necessary because of the number of separators affected and the extent of the degradation. Repair was not considered to be a practical option and it was decided to replace the separators, as required, in Unit 8 steam generators during an advanced scheduled outage. GXP separators were selected for replacement to minimize the impact on steam generator operating characteristics and analysis. The material of construction was upgraded from the original carbon steel to stainless steel to maximize the assurance of full life. The replacement of the separators was a first of a kind operation not only for Ontario Power Generation and B and W but also for all CANDU plants. The paper describes the degradations observed and the assessments, the operating experience, manufacture and installation of the replacement separators. During routine inspection in 1998, many of the primary steam separators in two of steam generators at Bruce Nuclear Division B Unit 8 were observed to have through wall perforations. This paper describes assessment of this condition. It also discusses the manufacture and testing of replacement primary steam separators and the development and execution of the replacement separator installation program. (author)

  7. Sodium and steam leak simulation studies for fluidized bed steam generators

    International Nuclear Information System (INIS)

    Keeton, A.R.; Vaux, W.G.; Lee, P.K.; Witkowski, R.E.

    1976-01-01

    An experimental program is described which was conducted to study the effects of sodium or steam leaking into an operating fluidized bed of metal or ceramic particles at 680 to 800 0 K. This effort was part of the early development studies for a fluidized-bed steam generator concept using helium as the fluidizing gas. Test results indicated that steam and small sodium leaks had no effect on the quality of fluidization, heat transfer coefficient, temperature distribution, or fluidizing gas pressure drop across the bed. Large sodium leaks, however, immediately upset the operation of the fluidized bed. Both steam and sodium leaks were detected positively and rapidly at an early stage of a leak by instruments specifically selected to accomplish this

  8. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

    1995-10-01

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  9. Emergency systems and protection equipment of modular steam generators for fast reactors

    International Nuclear Information System (INIS)

    Matal, O.

    The requirements are discussed for accident protection of modular steam generators for fast reactors. Accident protection is assessed for a modular through-flow steam generator and for a natural circulation modular steam generator. Benefits and constraints are shown and possible improvements are outlined for accident protection of liquid sodium fired modular steam generators. (Kr)

  10. Steam generator and condenser design of WWER-1000 type of nuclear power plant

    International Nuclear Information System (INIS)

    Zare Shahneh, Abolghasem.

    1995-03-01

    Design process of steam generator and condenser at Russian nuclear power plant type WWER-1000 is identified. The four chapter of the books are organized as nuclear power plant, types of steam generators specially horizontal steam generator, process of steam generator design and the description of condenser and its process design

  11. Steam Generator Group Project. Task 6. Channel head decontamination

    International Nuclear Information System (INIS)

    Allen, R.P.; Clark, R.L.; Reece, W.D.

    1984-08-01

    The Steam Generator Group Project utilizes a retired-from-service pressurized-water-reactor steam generator as a test bed and source of specimens for research. An important preparatory step to primary side research activities was reduction of the radiation field in the steam generator channel head. This task report describes the channel head decontamination activities. Though not a programmatic research objective it was judged beneficial to explore the use of dilute reagent chemical decontamination techniques. These techniques presented potential for reduced personnel exposure and reduced secondary radwaste generation, over currently used abrasive blasting techniques. Two techniques with extensive laboratory research and vendors prepared to offer commercial application were tested, one on either side of the channel head. As indicated in the report, both techniques accomplished similar decontamination objectives. Neither technique damaged the generator channel head or tubing materials, as applied. This report provides details of the decontamination operations. Application system and operating conditions are described

  12. The SNR-300 steam generator small leak detection system

    International Nuclear Information System (INIS)

    Dumm, K.

    1984-01-01

    Small leak detection in the SNR-300 steam generator moduls is achieved by hydrogen meters. Development and design of the Nickel membrane - ion getter pump combination are described and sensitivity requests derived. Results of calibration tests by water/steam injections in a sodium loop are presented. The arrangement and interconnection of signals in SNR-300 are given and possibilities for inservice calibrations are discussed, supported by long time operation tests in the KNK-reactor plant. (author)

  13. Hydrogen-based power generation from bioethanol steam reforming

    Energy Technology Data Exchange (ETDEWEB)

    Tasnadi-Asztalos, Zs., E-mail: tazsolt@chem.ubbcluj.ro; Cormos, C. C., E-mail: cormos@chem.ubbcluj.ro; Agachi, P. S. [Babes-Bolyai University, Faculty of Chemistry and Chemical Engineering, 11 Arany Janos, Postal code: 400028, Cluj-Napoca (Romania)

    2015-12-23

    This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO{sub 2} emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint.

  14. Hydrogen-based power generation from bioethanol steam reforming

    International Nuclear Information System (INIS)

    Tasnadi-Asztalos, Zs.; Cormos, C. C.; Agachi, P. S.

    2015-01-01

    This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO 2 emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint

  15. Hydrogen-based power generation from bioethanol steam reforming

    Science.gov (United States)

    Tasnadi-Asztalos, Zs.; Cormos, C. C.; Agachi, P. S.

    2015-12-01

    This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO2 emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint.

  16. Decontamination of Steam Generator tube using Abrasive Blasting Technology

    International Nuclear Information System (INIS)

    Min, B. Y.; Kim, G. N.; Choi, W. K.; Lee, K. W.; Kim, D. H.; Kim, K. H.; Kim, B. T.

    2010-01-01

    As a part of a technology development of volume reduction and self disposal for large metal waste project, We at KAERI and our Sunkwang Atomic Energy Safety (KAES) subcontractor colleagues are demonstrating radioactively contaminated steam generator tube by abrasive blasting technology at Kori-1 NPP. A steam generator is a crucial component in a PWR (pressurized Water Reactor). It is the crossing between the primary, contaminated, circuit and the secondary waste-steam circuit. The heat from the primary reactor coolant loop is transferred to the secondary side in thousands of small tubes. Due to several problems in the material of those tube, like SCC (Stress Corrosion Cracking), insufficient control in water chemistry, which can be cause of tube leakage, more and more steam generators are replaced today. Only in Korea, already 2 of them are replaced and will be replaced in the near future. The retired 300 ton heavy Steam generator was stored at the storage waste building of Kori NPP site. The steam generator waste has a large volume, so that it is necessary to reduce its volume by decontamination. A waste reduction effect can be obtained through decontamination of the inner surface of a steam generator. Therefore, it is necessary to develop an optimum method for decontamination of the inner surface of bundle tubes. The dry abrasive blasting is a very interesting technology for the realization of three-dimensional microstructures in brittle materials like glass or silicon. Dry abrasive blasting is applicable to most surface materials except those that might be shattered by the abrasive. It is most effective on flat surface and because the abrasive is sprayed and can also applicable on 'hard to reach' areas such as inner tube ceilings or behind equipment. Abrasive decontamination techniques have been applied in several countries, including Belgium, the CIS, France, Germany, Japan, the UK and the USA

  17. Drying system for steam generators, particularly for steam generators of nuclear power stations

    International Nuclear Information System (INIS)

    Lavalerie, Claude; Borrel, Christian.

    1982-01-01

    A drying system is described which allows for modular construction and which provides a significant available exchange area in a reduced volume. All the drying elements are identical and are distributed according to a ternay circular symmetry and are placed radially and associated to steam guiding facilities which alternately provide around the axis of revolution an output volume of dry steam from one element and an input volume of wet steam in the following element [fr

  18. Microfabricated rankine cycle steam turbine for power generation and methods of making the same

    Science.gov (United States)

    Frechette, Luc (Inventor); Muller, Norbert (Inventor); Lee, Changgu (Inventor)

    2009-01-01

    In accordance with the present invention, an integrated micro steam turbine power plant on-a-chip has been provided. The integrated micro steam turbine power plant on-a-chip of the present invention comprises a miniature electric power generation system fabricated using silicon microfabrication technology and lithographic patterning. The present invention converts heat to electricity by implementing a thermodynamic power cycle on a chip. The steam turbine power plant on-a-chip generally comprises a turbine, a pump, an electric generator, an evaporator, and a condenser. The turbine is formed by a rotatable, disk-shaped rotor having a plurality of rotor blades disposed thereon and a plurality of stator blades. The plurality of stator blades are interdigitated with the plurality of rotor blades to form the turbine. The generator is driven by the turbine and converts mechanical energy into electrical energy.

  19. Steam generator assessment for sustainable power plant operation

    International Nuclear Information System (INIS)

    Drexler, Andreas; Fandrich, Joerg; Ramminger, Ute; Montaner-Garcia, Violeta

    2012-09-01

    Water and steam serve in the water-steam cycle as the energy transport and work media. These fluids shall not affect, through corrosion processes on the construction materials and their consequences, undisturbed plant operation. The main objectives of the steam water cycle chemistry consequently are: - The metal release rates of the structural materials shall be minimal - The probability of selective / localized forms of corrosion shall be minimal. - The deposition of corrosion products on heat transfer surfaces shall be minimized. - The formation of aggressive media, particularly local aggressive environments under deposits, shall be avoided. These objectives are especially important for the steam generators (SGs) because their condition is a key factor for plant performance, high plant availability, life time extension and is important to NPP safety. The major opponent to that is corrosion and fouling of the heating tubes. Effective ways of counteracting all degradation problems and thus of improving the SG performance are to keep SGs in clean conditions or if necessary to plan cleaning measures such as mechanical tube sheet lancing or chemical cleaning. Based on more than 40 years of experience in steam-water cycle water chemistry treatment AREVA developed an overall methodology assessing the steam generator cleanliness condition by evaluating all available operational and inspection data together. In order to gain a complete picture all relevant water chemistry data (e.g. corrosion product mass balances, impurity ingress), inspection data (e.g. visual inspections and tube sheet lancing results) and thermal performance data (e.g. heat transfer calculations) are evaluated, structured and indexed using the AREVA Fouling Index Tool Box. This Fouling Index Tool Box is more than a database or statistical approach for assessment of plant chemistry data. Furthermore the AREVA's approach combines manufacturer's experience with plant data and operates with an

  20. PMK-2. Experimental study on steam generator behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Ezsoel, G.; Szabados, L.; Trosztel, I. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1995-12-31

    The PMK-2 is a full pressure scaled-down model of the Paks Nuclear Power Plant, with a 1:2070 scaling ratio for the volume and power. It has a steam generator model which is a vertical section of the horizontal steam generator. The model has hot and cold collectors similarly to the steam generators of the plant. The heat transfer tubes are horizontal tubes. There are 82 rows of tubes and the elevations, as well as the heat transfer surface distribution is the same as in the plant. The elevation of the feed water supply is similar to that of the plant. To study the temperature distribution in both the primary and the secondary side several thermocouples are built in, in addition to the overall instrumentation of the loop which has again a high number of measurement channels. Paper gives a description and results of SPE-4, with special respect to the steam generator behaviour in both steady state and transient conditions. Axial distribution of coolant and feedwater temperatures are given for the primary and the secondary side of hot and cold collectors and the temperature distribution in the centre of steam generator. (orig.).

  1. PMK-2. Experimental study on steam generator behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Ezsoel, G; Szabados, L; Trosztel, I [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1996-12-31

    The PMK-2 is a full pressure scaled-down model of the Paks Nuclear Power Plant, with a 1:2070 scaling ratio for the volume and power. It has a steam generator model which is a vertical section of the horizontal steam generator. The model has hot and cold collectors similarly to the steam generators of the plant. The heat transfer tubes are horizontal tubes. There are 82 rows of tubes and the elevations, as well as the heat transfer surface distribution is the same as in the plant. The elevation of the feed water supply is similar to that of the plant. To study the temperature distribution in both the primary and the secondary side several thermocouples are built in, in addition to the overall instrumentation of the loop which has again a high number of measurement channels. Paper gives a description and results of SPE-4, with special respect to the steam generator behaviour in both steady state and transient conditions. Axial distribution of coolant and feedwater temperatures are given for the primary and the secondary side of hot and cold collectors and the temperature distribution in the centre of steam generator. (orig.).

  2. Steam generator inspection activities at the EPRI NDE Center

    International Nuclear Information System (INIS)

    Krzywosz, K.

    1988-01-01

    Various types of corrosion and mechanical damage continue to affect the availability of both recirculating and once-through steam generators. Both the tube bundle and its supporting structure are affected. Intergranular attack and stress corrosion cracking (SCC) are the corrosion-assisted tube-wall damage mechanisms of most concern at this time. Fatigue cracking and fretting at antivibration bars are currently the mechanical damage forms causing most concern. Improved NDE equipment and techniques are providing better detection and characterization of adverse conditions within the steam generators and doing it at an earlier stage. This allows timely corrective action. To maintain the projected life expectancy of existing and new steam generators, remedial measures have been implemented. These measures include shot- or roto-peening, U-bend stress relief, chemical cleaning of secondary side, and sleeving of tubes. The improved NDE technology will also be instrumental in monitoring and assessing the effectiveness of the remedial measures. The revision of guidance documents for steam generator in-service inspection (ISI) is providing more relevant information to support this complex operation. A multitasked project is described that includes evaluation of steam generator tube NDE technology, transfer of this technology to utilities, and rapid response utility assistance

  3. Automation of steam generator services at public service electric & gas

    Energy Technology Data Exchange (ETDEWEB)

    Cruickshank, H.; Wray, J.; Scull, D. [Public Service Electric & Gas, Hancock`s Bridge, NJ (United States)

    1995-03-01

    Public Service Electric & Gas takes an aggressive approach to pursuing new exposure reduction techniques. Evaluation of historic outage exposure shows that over the last eight refueling outages, primary steam generator work has averaged sixty-six (66) person-rem, or, approximately tewenty-five percent (25%) of the general outage exposure at Salem Station. This maintenance evolution represents the largest percentage of exposure for any single activity. Because of this, primary steam generator work represents an excellent opportunity for the development of significant exposure reduction techniques. A study of primary steam generator maintenance activities demonstrated that seventy-five percent (75%) of radiation exposure was due to work activities of the primary steam generator platform, and that development of automated methods for performing these activities was worth pursuing. Existing robotics systems were examined and it was found that a new approach would have to be developed. This resulted in a joint research and development project between Westinghouse and Public Service Electric & Gas to develop an automated system of accomplishing the Health Physics functions on the primary steam generator platform. R.O.M.M.R.S. (Remotely Operated Managed Maintenance Robotics System) was the result of this venture.

  4. Modular steam generator for use in nuclear power plants

    International Nuclear Information System (INIS)

    Cella, A.

    1979-01-01

    An improved steam generator for a PWR is described. A turbine generator is driven by the steam output of the steam generator to provide electrical power. The improvement provides vertically assemblable modules which are removably mounted together in sealing relationship. The modules comprising a base module, a tube bundle module removably mountable on the base module in sealing relationship, and an uppermost dryer module removably mountable on the tube bundle module in sealing relationship. Ready access to and removal of the tube bundle module in situ from the nuclear power plant steam generator is facilitated. The dryer module contains moisture separator for drying the generated steam. The base module, upon which the associated weight of the vertically assembled dryer module and tube bundle module are supported, contains the inlet and outlet for the heat exchange fluid. The tube bundle module contains the tube bundle through which the heat exchange fluid flows as well as an inlet for feedwater. The tube sheet serves as a closure flange for the tube bundle module, with the associated weight of the vertically assembled dryer module and tube bundle module on the tube sheet closure flange effectuating the sealing relationship between the base module and the tube bundle module for facilitating closure

  5. Leak detection in Phenix and Super Phenix steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Cambillard, E [Centre d' Etudes Nucleaires de Saclay, Gif-sur-Yvette (France)

    1978-10-01

    Water leak detection Phenix and Super Phenix steam generators is based on measurement of the hydrogen produced by the reaction of sodium with water. The hydrogen evolves in the sodium in which the steam generator tubes are completely immersed. Depending on service conditions, however (sodium temperature and flow velocity), the hydrogen may appear in the argon existing above the free levels. This is why, although the Phenix steam generators do not feature free levels, measurement systems were added to measure the hydrogen concentration in the argon in the expansion tanks. Super Phenix steam generators are fitted at their outlet with systems for measuring hydrogen in the sodium, and above their free level with a system for measuring hydrogen in the argon. The measurement systems have nickel tube probes connected to circuits kept under vacuum by an ion pump. The hydrogen partial pressure is measured by a mass spectrometer. Super Phenix measurement systems differ from Phenix systems essentially in the temperature regulation of the sodium reaching the nickel tube probes, and in the centralization of the supply and measurement systems of the ion pumps and mass spectrometers. This paper deals with description, calibration and operating conditions of the hydrogen detection systems in sodium and argon in Phenix and Super Phenix steam generators. (author)

  6. Maintenance and plugging technology for CANDU steam generator tubing

    International Nuclear Information System (INIS)

    Prince, J.; Nicholson, A.; Hare, J.; McGoey, L.; Stafford, T.; Gowthorpe, P.

    2006-01-01

    In order to keep aging steam generators in service and to successfully manage the life of these critical components, the capability must exist to perform tube plugging and other complex maintenance activities in-situ. In the early days of CANDU steam generator operation, the only option was to perform these activities manually, which had inherent safety and quality risks. The challenge was to be able to perform these activities remotely thus eliminating some of the confined space and radiological exposure risks. The additional challenge was to develop equipment and techniques which would result in significantly improved quality, particularly for the completed plug welds which would be returned to service. Over the past fifteen years, this technology has matured and has produced remarkable results in field application. Some 14000 tube plugs have been successfully installed to date using automated plugging techniques. This paper presents an overview of the development of techniques available to utilities for steam generator tube plugging as well as some highlights of other steam generator tube maintenance activities such as primary side tube removal and tube end damage repair. Aspects covered in the paper include plug and procedure development, automated equipment and manipulators for tool deployment, process controls and personnel requirements. Recently, the steam generator tube plugging performed by OPG has been incorporated into a formal quality program under the requirements of ASME NCA 4000. An overview of the quality program will be presented and details of some of the important aspects of the quality program will be discussed. (author)

  7. Steam generator replacement at Kansai Electric Power Co., Inc

    International Nuclear Information System (INIS)

    Kimura, S.; Dodo, Takashi; Negishi, Kazuo

    1995-01-01

    Eleven nuclear units are in operation at the Kansai Electric Power Co., Inc.. In seven of them, Mihama-1·2·3, Takahama-1·2, and Ohi-1·2, comparatively long duration for tube inspection and repair have been required during late annual outages. KEPCO decided to replace all steam generators in these 7 units with the latest model which was improved upon the past degradation experiences, as a result of comprehensive considerations including public confidence in nuclear power generation, maintenability, and economic efficiency. This report presents the design improvements in new steam generators, replacement techniques, and so on. (author)

  8. Automated analysis technique developed for detection of ODSCC on the tubes of OPR1000 steam generator

    International Nuclear Information System (INIS)

    Kim, In Chul; Nam, Min Woo

    2013-01-01

    A steam generator (SG) tube is an important component of a nuclear power plant (NPP). It works as a pressure boundary between the primary and secondary systems. The integrity of a SG tube can be assessed by an eddy current test every outage. The eddy current technique(adopting a bobbin probe) is currently the main technique used to assess the integrity of the tubing of a steam generator. An eddy current signal analyst for steam generator tubes continuously analyzes data over a given period of time. However, there are possibilities that the analyst conducting the test may get tired and cause mistakes, such as: missing indications or not being able to separate a true defect signal from one that is more complicated. This error could lead to confusion and an improper interpretation of the signal analysis. In order to avoid these possibilities, many countries of opted for automated analyses. Axial ODSCC (outside diameter stress corrosion cracking) defects on the tubes of OPR1000 steam generators have been found on the tube that are in contract with tube support plates. In this study, automated analysis software called CDS (computer data screening) made by Zetec was used. This paper will discuss the results of introducing an automated analysis system for an axial ODSCC on the tubes of an OPR1000 steam generator.

  9. In situ ultrasonic examination of high-strength steam generator support bolts

    International Nuclear Information System (INIS)

    Jusino, A.

    1985-01-01

    Currently employed high-strength steam generator support bolting material (designed prior to ASME Section III Part NF or Component Supports), 38.1 mm in diameter, in combination with high preloads are susceptible to stress corrosion cracking because of the relatively low stress corrosion resistance (K/sub ISCC/) properties. These bolts are part of the pressurized water reactor steam generator supports at the integral support pads (three per steam generator, with each pad housing six, eight, or ten bolts depending on the design). The US Nuclear Regulatory Commission concerns for high-strength bolting were identified in NUREG-0577, ''Potential for Low Fracture Toughness and Laminar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports,'' which was issued for comment on unresolved safety issue A-12. Subsequently, the bolting issues were addressed in generic issue B29. One of the issues deals specifically with high-strength bolting materials, which are vulnerable to stress corrosion cracking. A Westinghouse Owners Group funded program was established to develop in situ ultrasonic examination techniques to determine steam generator support bolting integrity at the head-to-shank and first-thread locations. This program was established in order to determine bolting integrity in place. Ultrasonic techniques were developed for both socket-head and flat-head bolt configurations. As a result of this program, in situ ultrasonic examination techniques were developed for examination of PWR steam generator support bolts. By employing these techniques utilities will be able to ensure the integrity of this in-place bolting without incurring the costs previously experienced during removal for surface examinations

  10. Steam turbine generators for Sizewell 'B' nuclear power station

    International Nuclear Information System (INIS)

    Hesketh, J.A.; Muscroft, J.

    1990-01-01

    The thermodynamic cycle of the modern 3000 r/min steam turbine as applied at Sizewell 'B' is presented. Review is made of the factors affecting thermal efficiency including the special nature of the wet steam cycle and the use of moisture separation and steam reheating. Consideration is given to the optimization of the machine and cycle parameters, including particular attention to reheating and to the provision of feedheating, in order to achieve a high overall level of performance. A modular design approach has made available a family of machines suitable for the output range 600-1300 MW. The constructional features of the 630 MW Sizewell 'B' turbine generators from this range are described in detail. The importance of service experience with wet steam turbines and its influence on the design of modern turbines for pressurized water reactor (PWR) applications is discussed. (author)

  11. Evaluation of a dryer in a steam generator

    International Nuclear Information System (INIS)

    Xue Yunkui; Liu Shixun; Guandao, Xie; Chen Junliang

    1998-01-01

    The hooked-vane-type dryer is used in vertical, natural circulation steam generators used in PWR-type nuclear power stations. it separates the fine droplets of water carried by steam so that the steam generator outlet steam moisture is below 0.25%. Such low moisture is demanded to ensure a safe and economic operation of the unit. The dryer is composed of hooked vanes and a draining structure. A series of tests to screen different designs were performed using air-water mixture. The paper presents the results of the investigation of the effect of the number of drainage hooks , the bending angle , distance between two adjacent vanes, and other geometrical parameters on the performance of a hooked-vane-type steam dryer. It indicates that the dryer still works effectively when the moisture of the steam at the dryer inlet changes in a wide range, and that the performance of the dryer is closely related to the geometry of the draining structure . On the basis of the results of this program, a draining structure with an original design was selected and it is presented in the paper. The performance of the selected draining structure is better than that of similar structures in China and abroad. (author)

  12. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54... ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g)). (a) Pressure vessels in which steam is generated are classed as “Unfired Steam Boilers” except as...

  13. Chemical cleaning as an essential part of steam generator asset management

    International Nuclear Information System (INIS)

    Stiepani, C.; Ammann, F.; Jones, D.; Evans, S.; Harper, K.

    2010-01-01

    Accumulation of deposits is intrinsic for the operation of Steam Generators in PWRs. Such depositions often lead to reduction of thermal performance, loss of component integrity and, in some cases to power restrictions. Accordingly removal of such deposits is an essential part of the asset management of the Steam Generators in a Nuclear Power Plant. Every plant has its individual condition, history and constraints which need to be considered when planning and performing a chemical cleaning. Typical points are: Sludge load amount and constitution of the deposits; Sludge distribution in the steam generator; Existing or expected corrosion problems; Amount and treatment possibilities for the waste generated. Depending on these points the strategy for chemical cleaning shall be evolved. The range of treatment starts with very soft cleanings with a removal of approx 100 kg per steam generator and goes to a full scale cleaning which can remove up to several thousand kilograms of deposits from a steam generator. Depending on the goal to be achieved and the steam generator present an adequate cleaning method shall be selected. Flexible and 'customizable' cleaning methods that can be adapted to the individual needs of a plant are therefore a must. Particular for the application of preventive cleanings where repeated or even regular application are intended, special focus has to be put on low corrosion and easy waste handling. Therefore AREVA has developed the 'C3' concept, Customized Chemical Cleaning concept. This concept covers the entire range of steam generator cleaning. Particular for the preventive maintenance cleanings processes with extreme low corrosion rates and easy waste handling are provided which make repeated applications safe and cost efficient. (author)

  14. Current forgings and their properties for steam generator of nuclear plant

    International Nuclear Information System (INIS)

    Tsukada, Hisashi; Suzuki, Komei; Kusuhashi, Mikio; Sato, Ikuo

    1997-01-01

    Current steel forgings for steam generator (SG) of PWR plant are reviewed in the aspect of design and material improvement. The following three items are introduced. The use of integral type steel forgings for the fabrication of steam generator enhances the structural integrity and makes easier fabrication and inspection including in-service inspection. The following examples of current integral type forgings developed by the Japan Steel Works, Ltd. (JSW) are introduced: (1) primary head integrated with nozzles, manways and supports; (2) steam drum head integrated with nozzle and handhole; (3) conical shell integrated with cylindrical sections and handholes. In order to decrease the weight of steam generator, the high strength materials such as SA508, Cl.3a steel have been adopted in some cases. The properties of this steel are introduced and the chemistry and heat treatment condition are discussed. As one of the methods to minimize the macro- and micro-segregations, the use of vacuum carbon deoxidation (VCD), i.e. deoxidization of steel by gaseous CO reaction, with addition of Al for grain refining was investigated. The properties of SA508, Cl.3 steels with Low Si content are compared with those of conventional one

  15. Current forgings and their properties for steam generator of nuclear plant

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Hisashi; Suzuki, Komei; Kusuhashi, Mikio; Sato, Ikuo [Japan Steel Works Ltd., Muroran (Japan)

    1997-12-31

    Current steel forgings for steam generator (SG) of PWR plant are reviewed in the aspect of design and material improvement. The following three items are introduced. The use of integral type steel forgings for the fabrication of steam generator enhances the structural integrity and makes easier fabrication and inspection including in-service inspection. The following examples of current integral type forgings developed by the Japan Steel Works, Ltd. (JSW) are introduced: (1) primary head integrated with nozzles, manways and supports; (2) steam drum head integrated with nozzle and handhole; (3) conical shell integrated with cylindrical sections and handholes. In order to decrease the weight of steam generator, the high strength materials such as SA508, Cl.3a steel have been adopted in some cases. The properties of this steel are introduced and the chemistry and heat treatment condition are discussed. As one of the methods to minimize the macro- and micro-segregations, the use of vacuum carbon deoxidation (VCD), i.e. deoxidization of steel by gaseous CO reaction, with addition of Al for grain refining was investigated. The properties of SA508, Cl.3 steels with Low Si content are compared with those of conventional one.

  16. Thermo-hydraulic stability study of a steam generator

    International Nuclear Information System (INIS)

    Magni, M C; Marcel, C P; Delmastro, D F

    2012-01-01

    In this work a mathematical model developed to investigate the thermalhydraulic stability of a helically coiled steam generator is presented. Such a steam generator is prone to experiment density wave oscillations. The model is therefore used to analyze the stability of the CAREM-25 reactor steam generators. The model is linear, numerically non-diffusive and nodal. In addition, it is able to represent non-uniform heat transfer fluxes between the primary and secondary coolant circuits. By using this model the marginal stability condition is found by varying the inlet friction coefficient for different conditions. The results are then compared with those obtained with a different model for which a simple uniform heat flux profiled is assumed. It is found that with such simplification the density waves instability mechanism is overestimated in a wide range of operating powers. For very low powers, in the contrary, the so-called uniform model underestimates the stabilizing inlet friction and therefore it gives non-conservative results. With the use of the more realistic non-uniform power profile model, it was possible to determine that, for a CAREM-25 steam generator, the most stable conditions is found at 60MW when the reactor operates at nominal pressure. Moreover, it is found that at high power levels the stability performance is dominated by the two-phase friction component while at low power levels the friction component originated in the over heated steam region prevail (author)

  17. Mode Selection for Axial Flaw Detection in Steam Generator Tube Using Ultrasonic Guided Wave

    International Nuclear Information System (INIS)

    Yoon, Byung Sik; Yang, Seung Han; Guon, Ki Il; Kim, Yong Sik

    2009-01-01

    The eddy current testing method is mainly used to inspect steam generator tube during in-service inspection period. But the general problem of assessing the structural integrity of the steam generator tube using eddy current inspection is rather complex due to the presence of noise and interference signal under various conditions. However, ultrasonic testing as a nondestructive testing tool has become quite popular and effective for the flaw detection and material characterization. Currently, ultrasonic guided wave is emerging technique in power industry because of its various merits. But most of previous studies are focused on detection of circumferential oriented flaws. In this study, the steam generator tube of nuclear power plant was selected to detect axially oriented flaws and investigate guided wave mode identification. The longitudinal wave mode is generated using piezoelectric transducer frequency from 0.5 MHz, 1.0 MHz, 2.25MHz and 5MHz. Dispersion based STFT algorithm is used as mode identification tool

  18. Questions raised in developing fast reactor steam generator designs

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, P A; Hayden, O

    1975-07-01

    The most important component in the achievement of satisfactory LMFBR reliability is the steam generator. When the failure statistics of other nuclear steam generators and the implications of a sodium water reaction are considered, there is some cause for concern. It is apparent that considerable improvement in technology is necessary and until more experience on operating plant is available a conservative design approach must be taken. Many solutions have been proposed, varying from forced circulation straight tube modular to large single vessel once through helical designs. The paper poses what are considered to be the main questions which arise when making a choice of fast reactor steam generator type and tube configuration. The aim is to promote discussion amongst the assembled experts on their relative design approaches and the importance placed upon the various factors in reaching our common goal of ensuring the success of the LMFBR in its essential role of conserving world energy resources. (author)

  19. Design and operating experiences with 50MW steam generator

    International Nuclear Information System (INIS)

    Kawara, M.; Yamaki, H.; Kanamori, A.; Tanaka, K.; Takahashi, T.

    1975-01-01

    The main purpose of the 50 MW steam generator is to have experiences of manufacturing and operation with large scale steam generator including necessary research and development works which can be reflected on the design and fabrication of 'Monju' (Japan 300 MWe prototype LMFBR). The detailed design of the 50 MW steam, generator was begun on March, 1972 and succeeded in the demonstration of 72 hours continuous operation with full power on June, 1974. It has been successfully operated since then, the performances of which have been evaluated through various kinds of tests. In this paper, the following items are mainly discussed system design, thermal and hydraulic design, structure and fabrication and some experiences on testing operation including cleaning and sodium flushing of equipment, sodium level control system, the behavior of hydrogen detection system and general outlook of the performance. (author)

  20. Forced circulation type steam generator simulation code: HT4

    International Nuclear Information System (INIS)

    Okamoto, Masaharu; Tadokoro, Yoshihiro

    1982-08-01

    The purpose of this code is a understanding of dynamic characteristics of the steam generator, which is a component of High-temperature Heat Transfer Components Test Unit. This unit is a number 4th test section of Helium Engineering Demonstration Loop (HENDEL). Features of this report are as follows, modeling of the steam generator, a basic relationship for the continuity equation, numerical analysis techniques of a non-linear simultaneous equation and computer graphics output techniques. Forced circulation type steam generator with strait tubes and horizontal cut baffles, applied in this code, have be designed at the Over All System Design of the VHTRex. The code is for use with JAERI's digital computer FACOM M200. About 1.5 sec required for each time step reiteration, then about 40 sec cpu time required for a standard problem. (author)

  1. Review of EPRI's steam generator R and D program

    International Nuclear Information System (INIS)

    Millett, P.J.; Welty, C.J.

    1998-01-01

    EPRI has carried out an extensive R and D program on SG technology since the mid 1970's. Very early efforts under the auspices of the Steam Generator Owners Group (SGOG) focused on developing remedial actions for the critical SG corrosion issues of denting, wastage and pitting. Fundamental work was also carried out in the development of thermal hydraulic models for vibration and wear, chemical cleaning and tube repair techniques. In the late 1980's and continuing through today, the program has shifted emphasis towards management of steam generator degradation, primarily stress corrosion cracking of the SG tubes on both the primary and secondary sides. The current Steam Generator Management Program (SGMP) carries out R and D in four areas; materials, chemistry, thermal hydraulics and non-destructive testing. The strategic goals of this program and projects put in place to achieve these goals will be reviewed in detail in this paper. (author)

  2. Primary manway shielding and exhaust covers for a steam generator

    International Nuclear Information System (INIS)

    Wallace, W.R.; Immel, A.K.; Boro, I.; Lester, W.E. II.

    1990-01-01

    This paper discusses a radiation emission shielding cover in combination with a steam generator of a nuclear reactor for covering at least a portion of a manway of the steam generator for protecting an operator from radiation emission. It comprises a plate; a mounting assembly including a mounting flange for securing the mounting assembly adjacent the manway of the steam generator and a mounting bracket; a slide means mounted on the mounting bracket adjacent the manway; and guide means mounted on the plate for receiving the slide means such that the plate can be moved from an open position adjacent the manway to a closed position over at least a portion of the manway

  3. Generic steam generator life cycle management from a utility perspective

    International Nuclear Information System (INIS)

    Baker, R.L.

    1993-01-01

    Steam generator repairs and replacements, which have occurred over the last 10 years, have lead many utilities to evaluate the economics of continued maintenance on existing steam generators against the economics of steam generator replacement. Such an endeavor requires an assessment of the expected rate and types of degradation. In addition, an identification of possible preventative or remedial measures and their potential effectiveness must be made. To arrive at an assessment of the economic impact of various combinations of potential courses of action many utilities have employed in-house developed or customized commercial programs to convert technical assessments into economic impact evaluations. This paper intends to give the reader an introduction to the technical issues and insight into a method of addressing the economic impact of possible management strategies. 52 refs., 17 figs., 2 tabs

  4. Questions raised in developing fast reactor steam generator designs

    International Nuclear Information System (INIS)

    Taylor, P.A.; Hayden, O.

    1975-01-01

    The most important component in the achievement of satisfactory LMFBR reliability is the steam generator. When the failure statistics of other nuclear steam generators and the implications of a sodium water reaction are considered, there is some cause for concern. It is apparent that considerable improvement in technology is necessary and until more experience on operating plant is available a conservative design approach must be taken. Many solutions have been proposed, varying from forced circulation straight tube modular to large single vessel once through helical designs. The paper poses what are considered to be the main questions which arise when making a choice of fast reactor steam generator type and tube configuration. The aim is to promote discussion amongst the assembled experts on their relative design approaches and the importance placed upon the various factors in reaching our common goal of ensuring the success of the LMFBR in its essential role of conserving world energy resources. (author)

  5. Study group meeting on steam generators for LMFBR's. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1975-07-01

    The Meeting organised by IAEA international working group on fast reactors which considered that the subject of sodium heated steam generators was a topic which needed study by the experts of several disciplines. For example: people who design such steam generators, specialists in the field of sodium water reactions, experts in material and water chemistry and members of the utilities who would be the customers for such units. Besides the exchange of large amount of information, it was considered that further special studies were necessary for the following subjects: materials; maintenance and repair; operating procedures and control of steam generators. A separate study of sodium-water reactions was recommended considering the safety aspects related to large water leakage and economic advantage of possible detection and protection against small water leaks.

  6. Televisual inspection of cleanliness in secondary side of steam generator

    International Nuclear Information System (INIS)

    Zhou Shoukang; Ding Xunshen

    1995-01-01

    For U-type tube steam generator (UTSG), the deposit will be piled on the low velocity area of tubesheet, tube bundle and support plates during operation. The deposit is the hidden region of pollutant, and the concentration effect will create the corrosion situation in secondary side of steam generator. So, it is very necessary to inspect the cleanliness of SG. The paper introduces the technique of cleanliness inspection with the VP3 1TV apparatus. The steam generator cleanliness inspection of GNP unit 1 and 2, after installation, major transients and commercial operation, shows that the 1TV technique not only provides the cleanliness condition of tubesheet, and removes the loose foreign object, but also provides the basis for the determination of lancing procedure, confirms the result of lancing, and gets the cleanliness conclusion according to the appropriate qualified standard

  7. Water leak detection in steam generator of SUPER PHENIX

    International Nuclear Information System (INIS)

    Brunet, M.; Garnaud, P.; Ghaleb, D.; Kong, N.

    1988-01-01

    With the intent of detecting water leaks inside steam generators, we developed a third system, called acoustic detector, to complement hydrogen detectors and rupture disks (burst disks). The role of the acoustic system is to enable rapid intervention in the event of a leak growing rapidly which could rupture neighbouring tubes. In such a case, the detectable flow rate of the leak varies from a few tens of g/s to a few hundred g/s. At the SUPER PHENIX, three teams work in [20-100 kHz] and CEA/STA* [50-300 kHz]. The simulation of water leaks in the steam generator by the argon injections performed to date at 50% of the rated power has shown promising results. An anomaly in the evolution of the background noise at more than 50% loading of one of the two instrumented steam generators would make difficult any extrapolation to full power behaviour. (author)

  8. Dependence of steam generator vibrations on feedwater pressure

    International Nuclear Information System (INIS)

    Sadilek, J.

    1989-01-01

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

  9. Strain measurements of nuclear power plant steam generator antiseismic supports

    International Nuclear Information System (INIS)

    Kulichevsky, R.

    1997-01-01

    The nuclear power plants steam generators have different types of structural supports. One of these types are the antiseismic supports, which are intended to be under stress only if a seismic event takes place. Nevertheless, the antiseismic supports lugs, that are welded to the steam generator vessel, are subjected to thermal fatigue because of the temperature cycles related with the shut down and start up operations performed during the life of the nuclear power plant. In order to evaluate the stresses that the lugs are subjected to, several strain gages were welded on two supports lugs, positioned at two heights of one of the Embalse nuclear power plant steam generators. In this paper, the instrumentation used and the strain measurements obtained during two start up operations are presented. The influence of the plant start up operation parameters on the lugs strain evolution is also analyzed. (author) [es

  10. Design and operating experiences with 50MW steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kawara, M; Yamaki, H; Kanamori, A; Tanaka, K; Takahashi, T

    1975-07-01

    The main purpose of the 50 MW steam generator is to have experiences of manufacturing and operation with large scale steam generator including necessary research and development works which can be reflected on the design and fabrication of 'Monju' (Japan 300 MWe prototype LMFBR). The detailed design of the 50 MW steam, generator was begun on March, 1972 and succeeded in the demonstration of 72 hours continuous operation with full power on June, 1974. It has been successfully operated since then, the performances of which have been evaluated through various kinds of tests. In this paper, the following items are mainly discussed system design, thermal and hydraulic design, structure and fabrication and some experiences on testing operation including cleaning and sodium flushing of equipment, sodium level control system, the behavior of hydrogen detection system and general outlook of the performance. (author)

  11. Study group meeting on steam generators for LMFBR's. Summary report

    International Nuclear Information System (INIS)

    1975-07-01

    The Meeting organised by IAEA international working group on fast reactors which considered that the subject of sodium heated steam generators was a topic which needed study by the experts of several disciplines. For example: people who design such steam generators, specialists in the field of sodium water reactions, experts in material and water chemistry and members of the utilities who would be the customers for such units. Besides the exchange of large amount of information, it was considered that further special studies were necessary for the following subjects: materials; maintenance and repair; operating procedures and control of steam generators. A separate study of sodium-water reactions was recommended considering the safety aspects related to large water leakage and economic advantage of possible detection and protection against small water leaks

  12. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  13. Reliability of eddy current examination of steam generator tubes

    International Nuclear Information System (INIS)

    Birks, A.S.; Ferris, R.H.; Doctor, P.G.; Clark, R.A.; Spanner, G.E.

    1985-04-01

    A unique study of nondestructive examination reliability is underway at the Pacific Northwest Laboratory under US Nuclear Regulatory Commission sponsorship. Project participants include the Electric Power Research Institute and consortiums from France, Italy, and Japan. This study group has conducted a series of NDE examinations of tubes from a retired-from-service steam generator, using commercially available multifrequency eddy current equipment and ASME procedures. The examination results have been analyzed to identify factors contributing to variations in NDE inspection findings. The reliability of these examinations will then be validated by destructive analyses of the steam generator tubes. The program is expected to contribute to development of a model for steam generator inservice inspection sampling plans and inspection periods, as well as to improved regulatory guidelines for tube plugging

  14. Functional performance of the helical coil steam generator, Consolidated Nuclear Steam Generator (CNSG) IV system. Executive summary report

    International Nuclear Information System (INIS)

    Watson, G.B.

    1975-10-01

    The objective of this project was to study the functional performance of the CNSG - IV helical steam generator to demonstrate that the generator meets steady-state and transient thermal-hydraulic performance specifications and that secondary flow instability will not be a problem. Economic success of the CNSG concepts depends to a great extent on minimizing the size of the steam generator and the reactor vessel for ship installation. Also, for marine application the system must meet stringent specifications for operating stability, transient response, and control. The full-size two-tube experimental unit differed from the CNSG only in the number of tubes and the mode of primary flow. In general, the functional performance test demonstrated that the helical steam generator concept will exceed the specified superheat of 35F at 100% load. The experimental measured superheat at comparable operating conditions was 95F. Testing also revealed that available computer codes accurately predict trends and overall performance characteristics

  15. A study on improving the performance of steam generator using thermal analysis

    International Nuclear Information System (INIS)

    Li, Zhen Zhe; Heo, Kwang Su; Choi, Jun Hoo; Seol, Seoung Yun

    2008-01-01

    Steam generation mechanism is the key technology of domestic steam cleaner. Not only weight and price of steam cleaner but also the performance of steam generation mechanism must be considered to improve the competitive power of the products. In order to find out the mechanism which can be used to improve the performance of steam generator, the process of steam generation was studied at first. In the following step, possibility of control, safety of mechanism and etc were compared about the two candidated steam generation mechanism. Finally, the merit and drawback of each mechanism were summarized

  16. Acoustic noises of the BOR-60 reactor steam generators when simulating leaks with argon and steam

    International Nuclear Information System (INIS)

    Sokolov, V.M.; Golushko, V.V.; Afanas'ev, V.A.; Grebenkin, Yu.P.; Muralev, A.B.

    1985-01-01

    Background acoustic noises of stea generators in different operational regimes and noises of argon and steam small leads (about 0.1 g/s) are studied to determine the possibility of designing the acoustic system for leak detection in sodium-water steamgenerators. Investigations are carried out at the 30 MW micromodule steam generator being in operation at the BOR-60 reactor as well as at the 20 MW tank type steam generator. Immersed ransduceres made of lithium niobate 6 mm in-diameter and waveguide transducers made of a stainless steel in the form of rods 10 mm in-diameter and 500 mm long are used as acoustic monitors. It is shown that the leak noise is more wide-band than the background noise of the steam generator and both high and low frequencies appear in the spectrum. The use of monitors of different types results in similar conslusions inrelation to the character of background noises and leak signals (spectral density, signal to-noise ratio) in the ase of similar bandroidths of the transduceres. A conclusion is made that the change of operational regimes leads to changes of background noise level, which can be close to the reaction of

  17. Proceedings of steam generator sludge deposition in recirculating and once through steam generator upper tube bundle and support plates

    International Nuclear Information System (INIS)

    Baker, R.L.; Harvego, E.A.

    1992-01-01

    The development of remedial measures of shot peening have given nuclear utilities viable measures to address primary water stress corrosion cracking to extend steam generator life. The nuclear utility industry is now faced with potential replacement of steam generators in nuclear power plants due to stress corrosion cracking and intergranular attach in crevice locations on the secondary side of steam generators at tube support plates and at the crevice at the top of the tube sheet. Significant work has been done on developing and understanding of the effects of sludge buildup on the corrosion process at these locations. This session was envisioned to provide a forum for the development of an understanding of the mechanisms which control the transport and deposition of sludge on the secondary side of steam generators. It is hoped that this information will aid utilities in monitoring the progression of fouling of these crevices by further knowledge in where to look for the onset of support plate crevice fouling. An understanding of the progression of fouling from upper tube support plates to those lower in the steam generator where higher temperatures cause the corrosion process to initiate first can aid the nuclear utility industry in developing remedial measures for this condition and in providing a forewarning of when to apply such remedial measures

  18. Design of jet manipulator for sludge lancing for steam generators

    International Nuclear Information System (INIS)

    Kumar, Kundan; Nathani, D.K.; Kayal, J.N.; Rupani, B.B.

    2006-01-01

    The sludge accumulation in secondary side of mushroom type steam generators of Indian Pressurised Heavy Water Reactors (PHWRs) may lead to loss of thermal efficiency and corrosion. Sludge removal is required to minimise such effects for safe and enhanced operating life of the steam generators. A sludge lancing system has been developed for sludge removal from the secondary side of the steam generators. Jet Manipulator is one of the various modules of the sludge lancing system. The JM consists of three modules namely walker, elevator and nozzle heads. Each module is designed to pass through hand hole, having 180 mm diameter and 100 mm wide gap between steam generator shell and shroud. These three modules are connected to each other by quick connecting type joints and are having their specific functions. The walker crawls by step of single pitch of the tube along the central no-tube lane of the steam generator by taking lateral supports on the nearest tubes. The elevator is capable of lifting the nozzle head to a suitable height required for lancing operation of entire tube sheet of the steam generator. The nozzle head directs the multiple jets along the narrow inter tube lanes having 3 mm width, on both sides of the central no-tube lane. The nozzle can be set to move at different elevations such that the multiple jets will graze along the narrow tube lane to create the sludge lancing action. The provision exists for movement of JM in both directions, i.e. forward and reverse. This paper highlights the objective, design and development, selection of nozzles, qualification and performance evaluation of JM. The manipulator is remotely operable by compressed air in the forward and reverse direction in the central no-tube lane to position the nozzle head in the horizontal direction. (author)

  19. Steam generator for use in nuclear power plants

    International Nuclear Information System (INIS)

    Cella, A.

    1980-01-01

    An improved steam generator is described for use in a nuclear power plant of the pressurized water type in which a turbine generator is driven by the steam output of the steam generator to provide electrical power therefrom. The improvement comprises providing a vertically movable grid structure vertically extending within the interior of the lower housing portion of the steam generator through which individual tubes comprising a vertically extending tube bundle extend. The tube bundle has a tube sheet at one end thereof supporting the tube bundle for the tubes extending through the tube sheet in flow through communication with a heat exchange fluid inlet. The grid structure defines grid apertures therein through which the individual tubes extend with each of the grid apertures being in surrounding relationship with a portion of an associated one of the tubes. The grid structure is movable for a predetermined vertical extent, such as by hydraulic means, such as a piston, along the tubes for vertically displacing the means defining the grid apertures by a sufficient amount for removing the previously surrounded portion of each of the tubes from the associated grid apertures whereby an enhanced reading of the condition of the tubes at the previously surrounded portion is enabled. The steam generator may comprise vertically assemblable modules which are removably mounted together in sealing relationship, with the modules comprising a base module, a tube bundle module removably mountable on the base module in sealing relationship therewith and an uppermost drier module removably mountable on the tube bundle module in sealing relationship therewith whereby ready access to removal of the tube bundle module in situ from the nuclear power plant steam generator is facilitated

  20. Nuclear steam generator tube to tubesheet joint optimization

    International Nuclear Information System (INIS)

    McGregor, Rod

    1999-01-01

    Industry-wide problems with Stress Corrosion Cracking in the Nuclear Steam Generator tube-to-tubesheet joint have led to costly repairs, plugging, and replacement of entire vessels. To improve corrosion resistance, new and replacement Steam Generator developments typically employ the hydraulic tube expansion process (full depth) to minimize tensile residual stresses and cold work at the critical transition zone between the expanded and unexpanded tube. These variables have undergone detailed study using specialized X-ray diffraction and analytical techniques. Responding to increased demands from Nuclear Steam Generator operators and manufacturers to credit the leak-tightness and strength contributions of the hydraulic expansion, various experimental tasks with complimentary analytical modelling were applied to improve understanding and control of tube to hole contact pressure. With careful consideration to residual stress impact, design for strength/leak tightness optimization addresses: Experimentally determined minimum contact pressure levels necessary to preclude incipient leakage into the tube/hole interface. The degradation of contact pressure at surrounding expansions caused by the sequential expansion process. The transient and permanent contact pressure variation associated with tubesheet hole dilation during Steam Generator operation. An experimental/analytical simulation has been developed to reproduce cyclic Steam Generator operating strains on the tubesheet and expanded joint. Leak tightness and pullout tests were performed during and following simulated Steam Generator operating transients. The overall development has provided a comprehensive understanding of the fabrication and in-service mechanics of hydraulically expanded joints. Based on this, the hydraulic expansion process can be optimized with respect to critical residual stress/cold work and the strength/leakage barrier criteria. (author)