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Sample records for steam generator tubing

  1. Steam generator tube failures

    SciTech Connect

    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.

  2. Steam generator tube integrity program

    SciTech Connect

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

    1996-03-01

    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.

  3. Reliability of steam generator tubing

    SciTech Connect

    Kadokami, E.

    1997-02-01

    The author presents results on studies made of the reliability of steam generator (SG) tubing. The basis for this work is that in Japan the issue of defects in SG tubing is addressed by the approach that any detected defect should be repaired, either by plugging the tube or sleeving it. However, this leaves open the issue that there is a detection limit in practice, and what is the effect of nondetectable cracks on the performance of tubing. These studies were commissioned to look at the safety issues involved in degraded SG tubing. The program has looked at a number of different issues. First was an assessment of the penetration and opening behavior of tube flaws due to internal pressure in the tubing. They have studied: penetration behavior of the tube flaws; primary water leakage from through-wall flaws; opening behavior of through-wall flaws. In addition they have looked at the question of the reliability of tubing with flaws during normal plant operation. Also there have been studies done on the consequences of tube rupture accidents on the integrity of neighboring tubes.

  4. Steam generator tubing NDE performance

    SciTech Connect

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

    1997-02-01

    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.

  5. Circumferential cracking of steam generator tubes

    SciTech Connect

    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.

  6. Workshop Proceedings: Pitting in Steam Generator Tubing

    SciTech Connect

    1984-10-01

    A two-day workshop focused on the probable causes of steam generator pitting at two nuclear plants and on whether pitting is a low-temperature or a high-temperature phenomenon. Participants also heard descriptions of various pit-resistant metals that are suitable for tube sleeving.

  7. Data analysis for steam generator tubing samples

    SciTech Connect

    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.

  8. Alternate tube plugging criteria for steam generator tubes

    SciTech Connect

    Cueto-Felgueroso, C.; Aparicio, C.B.

    1997-02-01

    The tubing of the Steam Generators constitutes more than half of the reactor coolant pressure boundary. Specific requirements governing the maintenance of steam generator tubes integrity are set in Plant Technical Specifications and in Section XI of the ASME Boiler and Pressure Vessel Code. The operating experience of Steam Generator tubes of PWR plants has shown the existence of some types of degradatory processes. Every one of these has an specific cause and affects one or more zones of the tubes. In the case of Spanish Power Plants, and depending on the particular Plant considered, they should be mentioned the Primary Water Stress Corrosion Cracking (PWSCC) at the roll transition zone (RTZ), the Outside Diameter Stress Corrosion Cracking (ODSCC) at the Tube Support Plate (TSP) intersections and the fretting with the Anti-Vibration Bars (AVBs) or with the Support Plates in the preheater zone. The In-Service Inspections by Eddy Currents constitutes the standard method for assuring the SG tubes integrity and they permit the monitoring of the defects during the service life of the plant. When the degradation reaches a determined limit, called the plugging limit, the SG tube must be either repaired or retired from service by plugging. Customarily, the plugging limit is related to the depth of the defect. Such depth is typically 40% of the wall thickness of the tube and is applicable to any type of defect in the tube. In its origin, that limit was established for tubes thinned by wastage, which was the predominant degradation in the seventies. The application of this criterion for axial crack-like defects, as, for instance, those due to PWSCC in the roll transition zone, has lead to an excessive and unnecessary number of tubes being plugged. This has lead to the development of defect specific plugging criteria. Examples of the application of such criteria are discussed in the article.

  9. Steam generator tube inspection in Japan

    SciTech Connect

    Fukui, Shigetaka

    1997-02-01

    Steam generator tube inspection was first carried out in 1971 at Mihama Unit-1 that is first PWR plant in Japan, when the plant was brought into the first annual inspection. At that time, inspection was made on sampling basis, and only bobbin coil probe was used. After experiencing various kinds of tube degradations, inspection method was changed from sampling to all number of tubes, and various kinds of probes were used to get higher detectability of flaw. At present, it is required that all the tubes shall be inspected in their full length at each annual inspection using standard bobbin coil probe, and some special probes for certain plants that have susceptibility of occurrence of flaw. Sleeve repaired portion is included in this inspection. As a result of analyses of eddy current testing data, all indications that have been evaluated to be 20% wall thickness or deeper shall be repaired by either plugging or sleeving, where flaw morphology is to be a wastage or wear. Other types of flaw such as IGA/SCC are not allowed to be left inservice when those indications are detected. These inspections are performed according to inspection procedures that are approved by regulatory authority. Actual inspections are witnessed by the Japan Power engineering and inspection corporation (JAPEIC)`s inspectors during data acquisition and analysis, and they issue inspection report to authority for review and approval. It is achieved high safety performance of steam generator through this method of inspections, however. some tube leakage problems were experienced in the past. To prevent recurrence of such events, government is conducting development and verification test program for new eddy current testing technology.

  10. Advanced Eddy current NDE steam generator tubing.

    SciTech Connect

    Bakhtiari, S.

    1999-03-29

    As part of a multifaceted project on steam generator integrity funded by the U.S. Nuclear Regulatory Commission, Argonne National Laboratory is carrying out research on the reliability of nondestructive evaluation (NDE). A particular area of interest is the impact of advanced eddy current (EC) NDE technology. This paper presents an overview of work that supports this effort in the areas of numerical electromagnetic (EM) modeling, data analysis, signal processing, and visualization of EC inspection results. Finite-element modeling has been utilized to study conventional and emerging EC probe designs. This research is aimed at determining probe responses to flaw morphologies of current interest. Application of signal processing and automated data analysis algorithms has also been addressed. Efforts have focused on assessment of frequency and spatial domain filters and implementation of more effective data analysis and display methods. Data analysis studies have dealt with implementation of linear and nonlinear multivariate models to relate EC inspection parameters to steam generator tubing defect size and structural integrity. Various signal enhancement and visualization schemes are also being evaluated and will serve as integral parts of computer-aided data analysis algorithms. Results from this research will ultimately be substantiated through testing on laboratory-grown and in-service-degraded tubes.

  11. Steam generator tube integrity flaw acceptance criteria

    SciTech Connect

    Cochet, B.

    1997-02-01

    The author discusses the establishment of a flaw acceptance criteria with respect to flaws in steam generator tubing. The problem is complicated because different countries take different approaches to the problem. The objectives in general are grouped in three broad areas: to avoid the unscheduled shutdown of the reactor during normal operation; to avoid tube bursts; to avoid excessive leak rates in the event of an accidental overpressure event. For each degradation mechanism in the tubes it is necessary to know answers to an array of questions, including: how well does NDT testing perform against this problem; how rapidly does such degradation develop; how well is this degradation mechanism understood. Based on the above information it is then possible to come up with a policy to look at flaw acceptance. Part of this criteria is a schedule for the frequency of in-service inspection and also a policy for when to plug flawed tubes. The author goes into a broad discussion of each of these points in his paper.

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

    SciTech Connect

    Cepcek, S.

    1997-02-01

    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.

  13. Automated Diagnosis and Classification of Steam Generator Tube Defects

    SciTech Connect

    Dr. Gabe V. Garcia

    2004-10-01

    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.

  14. Steam generator tubing development for commercial fast breeder reactors

    SciTech Connect

    Sessions, C.E.; Uber, C.F.

    1981-11-01

    The development work to design, manufacture, and evaluate pre-stressed double-wall 2/one quarter/ Cr-1 Mo steel tubing for commercial fast breeder reactor steam generator application is discussed. The Westinghouse plan for qualifying tubing vendors to produce this tubing is described. The results achieved to date show that a long length pre-stressed double-wall tube is both feasible and commercially available. The evaluation included structural analysis and experimental measurement of the pre-stress within tubes, as well as dimensional, metallurgical, and interface wear tests of tube samples produced. This work is summarized and found to meet the steam generator design requirements. 10 refs.

  15. Estimating probable flaw distributions in PWR steam generator tubes

    SciTech Connect

    Gorman, J.A.; Turner, A.P.L.

    1997-02-01

    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.

  16. Research of laser cleaning technology for steam generator tubing

    NASA Astrophysics Data System (ADS)

    Hou, Suixa; Luo, Jijun; Xu, Jun; Yuan, Bo

    2010-10-01

    Surface cleaning based on the laser-induced breakdown of gas and subsequent shock wave generation can remove small particles from solid surfaces. Accordingly, several studies in steam generator tubes of nuclear power plants were performed to expand the cleaning capability of the process. In this work, experimental apparatus of laser cleaning was designed in order to clean heat tubes in steam generator. The laser cleaning process is monitored by analyzing acoustic emission signal experimentally. Experiments demonstrate that laser cleaning can remove smaller particles from the surface of steam generator tubes better than other cleaning process. It has advantages in saving on much manpower and material resource, and it is a good cleaning method for heat tubes, which can be real-time monitoring in laser cleaning process of heat tubes by AE signal. As a green cleaning process, laser cleaning technology in equipment maintenance will be a good prospect.

  17. Overview of steam generator tube degradation and integrity issues

    SciTech Connect

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

    1996-10-01

    The degradation of steam generator tubes in pressurized water nuclear reactors continues to be a serious problem. Primary water stress corrosion cracking is commonly observed at the roll transition zone at U-bends, at tube denting locations, and occasionally in plugs and sleeves. Outer-diameter stress corrosion cracking and intergranular attack commonly occur near the tube support plate crevice, near the tube sheet in crevices or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of circumferential cracking at the RTZ on both the primary and secondary sides. Segmented axial cracking at the tubes support plate crevices is also becoming more common. Despite recent advances in in-service inspection technology, a clear need still exists for quantifying and improving the reliability of in- service inspection methods with respect to the probability of detection of the various types of flaws and their accurate sizing. Improved inspection technology and the increasing occurrence of such degradation modes as circumferential cracking, intergranular attack, and discontinuous axial cracking have led to the formulation of a new performance-based steam generator rule. This new rule would require the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes perform the required safety function over the next operating cycle. The new steam generator rule will also be applied to severe accident conditions to determine the continued serviceability of a steam generator with degraded tubes in the event of a severe accident. Preliminary analyses are being performed for a hypothetical severe accident scenario to determine whether failure will occur first in the steam generator tubes, which would lead to containment bypass, or instead in the hot leg nozzle or surge line, which would not.

  18. Health and safety impact of steam generator tube degradation

    SciTech Connect

    Marston T.

    1997-02-01

    In this paper the author addresses the problems inherent in evaluating the safety of steam generators with respect to tube rupture as part of a probabilistic safety analysis (PSA) of a reactor plant. He reviews the history of PSA as applied to reactors, and then looks at tube rupture histories as a start toward establishing event frequencies. He considers tube ruptures from the aspect of being an initiating event to being a conditional event to some other event, and then the question of performance of the steam generator in the face of a severe accident in the reactor.

  19. Evaluation of steam generator WWER 440 tube integrity criteria

    SciTech Connect

    Splichal, K.; Otruba, J.; Burda, J.

    1997-02-01

    The main corrosion damage in WWER steam generators under operating conditions has been observed on the outer surface of these tubes. An essential operational requirement is to assure a low probability of radioactive primary water leakage, unstable defect development and rupture of tubes. In the case of WWER 440 steam generators the above requirements led to the development of permissible limits for data evaluation of the primary-to-secondary leak measurements and determination of acceptable values for plugging of heat exchange tubes based on eddy current test (ECT) inspections.

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

    SciTech Connect

    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.

  1. Subcooled choked flow through steam generator tube cracks

    NASA Astrophysics Data System (ADS)

    Wolf, Brian J.

    The work presented here describes an experimental investigation into the choked flow of initially subcooled water through simulated steam generator tube cracks at pressures up to 6.9 MPa. The study of such flow is relevant to the prediction of leak flow rates from a nuclear reactor primary side to secondary side through cracks in steam generator tubes. An experimental approach to measuring such flow is de- scribed. Experimental results from data found in literature as well as the data collected in this work are compared with predictions from presented models as well as predictions from the thermal-hydraulic system code RELAP5. It is found that the homogeneous equilibrium model underpredicts choked flow rates of subcooled water through slits and artificial steam generator tube cracks. Additional modeling of thermal non-equilibrium improves the predictibility of choking mass flux for homogeneous models, however they fail to account for the characteristics of the two-phase pressure drop. An integral modeling approach is enhanced using a correlation developed from the data herein. Also, an assessment of the thermal-hydraulics code RELAP5 is performed and it’s applicability to predict choking flow rates through steam generator tube cracks is addressed. This assessment determined that the Henry & Fauske model, as coded in RELAP5, is best suited for modeling choked flow through steam generator tube cracks. Finally, an approach to applying choked flow data that is not at the same thermo-dynamic conditions as a prototype is developed.

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

    SciTech Connect

    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.

  3. Characterization of oxides on Bruce A NGS liner tubes and steam generator tubes

    SciTech Connect

    Miller, D.G.; Burrill, K.A.

    1998-12-31

    Oxide deposits on end-fitting liner tubes and steam generator tubes from the Bruce A Nuclear Generating Station (NGS) were characterized in advance of the decontamination of the heat transport system (HTS) of Bruce Unit 2. Oxide loadings, and Co-60 surface activities and specific activities were determined for the oxides on inlet and outlet end-fitting liner tubes from Bruce Unit l, Bruce Unit 2 and Bruce Unit 4. Oxides on the inner surfaces of steam generator tubes from Bruce NGS Units 1 and 2 were also characterized. The consistency in the deposit characteristics on the inlet liner tubes and steam generator tubes from Bruce A, along with the absence of magnetite on the outlet liner tubes has led to the development of a model for iron transport in the HTS of pressurized heavy water reactors (PHWRs). The activity transport/fouling mechanism involves flow-accelerated corrosion of the outlet feeder pipes, followed by deposition of iron in the steam generators, along the inlet feeder pipes, on the inlet end fittings, on the inlet fuel bundles and on the inlet region of the pressure tube. The results of loop experiments using decontamination solutions indicated that the oxide was rapidly removed from inlet liner tubes. However, removal of the Cr-rich oxide from the outlet liner tubes was less efficient, requiring the Alkaline Permangante (AP) oxidizing pre-treatment that is typically used in light water reactors (LWRs). The steam generator tubes were effectively decontaminated.

  4. WWER Steam Generators Tubing Performance and Aging Management

    SciTech Connect

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

    2006-07-01

    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)

  5. Modeling of eddy current NDE probe for steam generator tubes.

    SciTech Connect

    Chang, F. C.; Bakhtiari, S.; Kupperman, D.

    2003-01-29

    Calculations were performed with a three-dimensional (3-D) finite-element model to describe the response of an eddy current (EC) probe to defects in steam generator (SG) tubing of a nuclear reactor. Such calculations could be very helpful in understanding and interpreting the EC probe response to complex tube/defect geometries associated with longitudinal inner/outer notches, roll transitions, sludge, and through-wall holes in SG tubes. The governing field equations are derived in terms of coupled magnetic vector and electric scalar potentials in the conducting media and total or reduced scalar potentials in the non-conducting regions. To assess the validity of the model, we compared the signal responses for two numerical approaches, stored-energy-and-power-loss approach and magnetic-flux approach for various tube/defect geometries. Simulation results are also presented on the tube/defect geometries for the pancake coil response and the transmitter/receiver (T/R) probe response. The results indicate that the eddy-current NDE modeling is capable of predicting EC probe response to flaws in steam generator tubes.

  6. Heat transfer simulation in a helically coiled tube steam generator

    NASA Astrophysics Data System (ADS)

    Hassanzadeh, Bazargan; Keshavarz, Ali; Ebrahimi, Masood

    2014-01-01

    A symmetric helically coiled tube steam generator that operates by methane has been simulated analytically and numerically. In the analytical method, the furnace has been divided into five zones. The numerical method computes the total heat absorbed in the furnace, while the existing analytical methods compute only the radiation heat transfer. In addition, according to the numerical results, a correlation is proposed for the Nusselt number in the furnace.

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

    SciTech Connect

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

    1997-02-01

    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.

  8. Primary-side deposits on PWR steam-generator tubes

    SciTech Connect

    Bergmann, C.A.; Roesmer, J.; Perone, D.W.

    1983-03-01

    The evaluation of analyses of material removed from primary side steam generator tubing samples taken from nuclear plants that had operated for up to 7 effective full power years are presented in this report. The types of analyses incuded radiochemical, chemical, scanning electron microscope (SEM), and energy dispersive x-ray (EDAX) techniques to characterize the surfaces and composition of the tubing material. An evaluation of the data obtained and a comparison with in-core crud data and with values calculated by a mathematical activity transport model (CORA) are also given in the report.

  9. Steam generator tube integrity program leak rate tests. Progress report

    SciTech Connect

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

    1984-01-01

    This interim report presents preliminary results on leak rate tests performed on through-wall defected Inconel 600 steam generator tubing. Tube defects included an EDM (electro-discharge machine) notch and IGSCC (intergranular stress corrosion cracks) of various lengths. Tests were conducted at PWR operating temperatures with leakage of hot water/steam into air. A number of IGSCC cracks were unstable under the experiment conditions of these initial tests, continuing to grow until system capacity limitations resulted in decreased pressure differential. However, initial tesing also pointed to a need for reconfiguration of the test apparatus to sustain increased flow and, more importantly, alter the mode of control. The initial test configuration is based on flow control, with pressure differential across the specimen an independent variable. This often results in pressure increases too rapid to establish the initiation of crack instability. A reconfigured system based on pressure control with flow as an independent parameter is being recommended for future tests.

  10. Rupture pressure of wear degraded alloy 600 steam generator tubings

    NASA Astrophysics Data System (ADS)

    Hwang, Seong Sik; Namgung, Chan; Jung, Man Kyo; Kim, Hong Pyo; Kim, Joung Soo

    2008-02-01

    Fretting/wear degradation at the tube support in the U-bend region of a steam generator (SG) of a pressurized water reactor (PWR) has been reported. Simulated fretted flaws were machined on SG tubes of 195 mm in length. A pressure test was carried out with the tubes at room temperature by using a high pressure test facility which consisted of a water pressurizing pump, a test specimen section and a control unit. Water leak rates just after a ligament rupture or a burst were measured. Tubes degraded by up to 70% of the tube wall thickness (TW) showed a high safety margin in terms of the burst pressure during normal operating conditions. Tubes degraded by up to 50% of the TW did not show burst. Burst pressure depended on the defect depths rather than on the wrap angles. The tube with a wrap angle of 0° showed a fish mouth fracture, whereas the tube with a 45° wrap angle showed a three way fracture.

  11. Ultrasonic guided wave inspection concepts for steam generator tubing

    SciTech Connect

    Rose, J.L.; Rajana, K.M. ); Carr, F.T. )

    1994-02-01

    Some very exciting and promising results have been obtained with respect to the utilization and guided wave techniques for inspecting steam generator tubing. In addition to some theoretical considerations that were studied recently, work has been carried out in special probe design and development. This special probe is used in demonstration of law detection feasibility, and in understanding the conceptual development of a complete flaw detection system. This includes transducer, pulser-receiver system, and appropriate signal processing and pattern recognition software for reliable inspection. Ultrasonic NDE techniques have progressed quite rapidly during this decade for two principal reasons: advanced signal processing, and the use and understanding of multi-mode ultrasonic wave propagation. Both concepts are useful in the proposed work on guided wave propagation in steam generator tubing. These new directions go beyond the use of normal beam longitudinal waves and angle beam shear waves for inspection. Guided waves such as surface and Lamb waves can be used to monitor larger volumes of material with greater efficiency. The generation of these waves, however, is more complex. Theoretically one can produce a large number of modes in a structure with a simple loading arrangement. However, the generation of sufficient amounts of energy in a specific mode strongly depends on several factors. They include the loading system, angle of attack, probe frequency, frequency bandwidth, and a whole host of special transducer design and instrumentation parameters.

  12. Integrity of the tubes used in vertical and horizontal steam generators

    NASA Astrophysics Data System (ADS)

    Bergunker, V. D.

    2011-03-01

    Statistical data on experience gained from operation of steam generators around the world are presented, problems arising in vertical and horizontal steam generators are described, and the conditions of heattransfer tubes used in them are compared.

  13. Magnetic field simulation of magnetic phase detection sensor for steam generator tube in nuclear power plants

    NASA Astrophysics Data System (ADS)

    Ryu, Kwon-sang; Son, Derac; Park, Duck-gun; Kim, Yong-il

    2010-05-01

    Magnetic phases and defects are partly produced in steam generator tubes by stress and heat, because steam generator tubes in nuclear power plants are used under high temperature, high pressure, and radioactivity. The magnetic phases induce an error in the detection of the defects in steam generator tubes by the conventional eddy current method. So a new method is needed for detecting the magnetic phases in the steam generator tubes. We designed a new U-type yoke which has two kinds of coils and simulated the signal by the magnetic phases and defects in the Inconnel 600 tube.

  14. Circumferential cracking in steam generator tubes repaired by mechanical sleeving

    SciTech Connect

    Stubbe, J.; Pierson, E.; Laire, C.; Nedden, L. zur; Somville, P.; Royen, P. Van

    1995-12-31

    After one service cycle, leaks were detected in Doel 4 steam generator (SG) tubes repaired by mechanical sleeving (hydraulically + roll expanded). Two tubes were pulled and examined, one of them showing a big leak and the second being, pulled randomly. They both revealed through wall circumferential primary water stress corrosion cracking (PWSCC) at the upper hydraulic transition so that it was concluded that the problem was generic. A thorough assessment of the root causes of failure was undertaken, including stress and strain direct measurement by X-ray diffraction and photoelasticity, local stresses and temperature evaluation by calculation and stress corrosion cracking tests. Stress corrosion tests were carried out in 10 % NAOH environment, on mock-ups manufactured from reserve tubing of the plant simulating not only the upper joint but also the complete assembly (two joints). An estimate of the expected life was performed by comparison with reference mock-ups representative of the roll transitions (including the kiss roll). The findings are that the hydraulic expansion may generate high residual stresses, in spite of the very low residual deformations. Concerning, the temperature however, there are some indications that it could be substantially lower at the level of the cracking than at the tube to tubesheet roll transitions, which makes the quantified evaluation somewhat inaccurate. It is concluded that repair by mechanical sleeving is influenced by many parameters, including details of the installation procedure. Lifetime may be very limited when applied to PWSCC sensitive tubes and must be evaluated by appropriate testing. In particular, corrosion mock-ups should represent the entire sleeve, with both joints.

  15. Proceedings of steam generator sludge deposition in recirculating and once through steam generator upper tube bundle and support plates

    SciTech Connect

    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.

  16. Improved eddy-current inspection for steam generator tubing

    SciTech Connect

    Dodd, C.V.; Pate, J.R.; Allen, J.D. Jr.; Allen , Knoxville, TN )

    1989-01-01

    Computer programs have been written to allow the analysis of different types of eddy-current probes and their performance under different steam generator test conditions. The probe types include the differential bobbin probe, the absolute bobbin probe, the pancake probe and the reflection probe. The generator test conditions include tube supports, copper deposits, magnetite deposits, denting, wastage, pitting, cracking and IGA. These studies are based mostly on computed values, with the limited number of test specimens available used to verify the computed results. The instrument readings were computed for a complete matrix of the different test conditions, and then the test conditions determined as a function of the readings by a least-squares technique. A comparison was made of the errors in fit and instrument drift for the different probe types. The computations of the change in instrument reading due to the defects have led to an inversion'' technique in which the defect properties can be computed from the instrument readings. This has been done both experimentally and analytically for each of these probe types. 3 refs., 13 figs., 1 tab.

  17. Techniques for in-service inspection of heat-transfer tubes in steam generators

    SciTech Connect

    McClung, R.W.; Day, R.A.; Neely, H.H.; Powers, T.

    1981-01-01

    A multifaceted development program is in progress in the United States to study techniques for in-service inspection (ISI) of heat transfer tubes in breeder reactor steam generators. Several steam generator designs are involved. Although there are some similarities in the approaches, many of the details of techniques and capabilities are specific to the steam generator design. This paper describes the ultrasonic, eddy-current and penetrating radiation techniques being studied for the various steam generators, including the Large Leak Test Rig, the Clinch River Breeder Reactor design, and alternate steam generators being developed by Westinghouse and Babcock and Wilcox.

  18. Improvements in the simulation of a main steam line break with steam generator tube rupture

    NASA Astrophysics Data System (ADS)

    Gallardo, Sergio; Querol, Andrea; Verdú, Gumersindo

    2014-06-01

    The result of simultaneous Main Steam Line Break (MSLB) and a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR) is a depressurization in the secondary and primary system because both systems are connected through the SGTR. The OECD/NEA ROSA-2 Test 5 performed in the Large Scale Test Facility (LSTF) reproduces these simultaneous breaks in a Pressurized Water Reactor (PWR). A simulation of this Test 5 was made with the thermal-hydraulic code TRACE5. Some discrepancies found, such as an underestimation of SG-A secondary pressure during the depressurization and overestimation of the primary pressure drop after the first Power Operated Relief Valve (PORV) opening can be improved increasing the nodalization of the Upper Head in the pressure vessel and meeting the actual fluid conditions of Upper Head during the transient.

  19. Simulation of a main steam line break with steam generator tube rupture using trace

    SciTech Connect

    Gallardo, S.; Querol, A.; Verdu, G.

    2012-07-01

    A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)

  20. Clinch River Breeder Reactor Plant Steam Generator Few Tube Test model post-test examination

    SciTech Connect

    Impellezzeri, J.R.; Camaret, T.L.; Friske, W.H.

    1981-03-11

    The Steam Generator Few Tube Test (FTT) was part of an extensive testing program carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. This paper describes the problems encountered with the mechanical features of the FTT model design which led to premature test termination, and the results of the post-test examination. Conditions of tube bowing and significant tube and tube support gouging was observed. An interpretation of the visual and metallurgical observations is also presented. The CRBRP steam generator has undergone design evaluations to resolve observed deficiences found in the FFTM.

  1. Proceedings: Workshop on Thermally Treated Alloy 690 Tubes for Nuclear Steam Generators

    SciTech Connect

    1986-07-01

    Data presented at this workshop confirmed the superior corrosion resistance of thermally treated alloy 690. Pending further testing and optimization procedures, this material appears to be the best choice for manufacture of nuclear steam generator tubes.

  2. Proceedings: 1990 EPRI workshop on circumferential cracking of steam generator tubes

    SciTech Connect

    Lowenstein, D.B.; Gorman, J.A. )

    1991-03-01

    A meeting on circumferential crackling of steam generator tubes was organized to give those working in this area an opportunity to share their results, ideas and plans with regard to determining the causes of circumferential cracks, how to perform inspections for circumferential cracks, and possible remedial approaches to reduce the occurrence of circumferential cracks in PWR steam generator tubes. Topics discussed included: (1) field experience; (2) causative factors and modeling; (3) remedial actions, (4) inspection issues, and (5) licensing/regulatory issues.

  3. Studies of the steam generator degraded tubes behavior on BRUTUS test loop

    SciTech Connect

    Chedeau, C.; Rassineux, B.

    1997-04-01

    Studies for the evaluation of steam generator tube bundle cracks in PWR power plants are described. Global tests of crack leak rates and numerical calculations of crack opening area are discussed in some detail. A brief overview of thermohydraulic studies and the development of a mechanical probabilistic design code is also given. The COMPROMIS computer code was used in the studies to quantify the influence of in-service inspections and maintenance work on the risk of a steam generator tube rupture.

  4. French Regulatory practice and experience feedback on steam generator tube integrity

    SciTech Connect

    Sandon, G.

    1997-02-01

    This paper summarizes the way the French Safety Authority applies regulatory rules and practices to the problem of steam generator tube cracking in French PWR reactors. There are 54 reactors providing 80% of French electrical consumption. The Safety Authority closely monitors the performance of tubes in steam generators, and requires application of a program which deals with problems prior to the actual development of leakage. The actual rules regarding such performance are flexible, responding to the overall performance of operating steam generators. In addition there is an inservice inspection service to examine tubes during shutdown, and to monitor steam generators for leakage during operation, with guidelines for when generators must be pulled off line.

  5. Analysis of potential for jet-impingement erosion from leaking steam generator tubes during severe accidents.

    SciTech Connect

    Majumdar, S.; Diercks, D. R.; Shack, W. J.; Energy Technology

    2002-05-01

    This report summarizes analytical evaluation of crack-opening areas and leak rates of superheated steam through flaws in steam generator tubes and erosion of neighboring tubes due to jet impingement of superheated steam with entrained particles from core debris created during severe accidents. An analytical model for calculating crack-opening area as a function of time and temperature was validated with tests on tubes with machined flaws. A three-dimensional computational fluid dynamics code was used to calculate the jet velocity impinging on neighboring tubes as a function of tube spacing and crack-opening area. Erosion tests were conducted in a high-temperature, high-velocity erosion rig at the University of Cincinnati, using micrometer-sized nickel particles mixed in with high-temperature gas from a burner. The erosion results, together with analytical models, were used to estimate the erosive effects of superheated steam with entrained aerosols from the core during severe accidents.

  6. Eddy Current Signature Classification of Steam Generator Tube Defects Using A Learning Vector Quantization Neural Network

    SciTech Connect

    Gabe V. Garcia

    2005-01-03

    A major cause of failure in nuclear steam generators is degradation of their tubes. Although seven primary defect categories exist, one of the principal causes of tube failure is intergranular attack/stress corrosion cracking (IGA/SCC). This type of defect usually begins on the secondary side surface of the tubes and propagates both inwards and laterally. In many cases this defect is found at or near the tube support plates.

  7. Library of PWR (pressurized-water reactor) steam generator tubing samples: Final report

    SciTech Connect

    Albertin, L.; Clark, W.G. Jr.; Junker, W.R.; Kuchirka, P.J.; Madeyski, A.; Metala, M.J.; Taszarek, B.J.

    1988-01-01

    The PWR Steam Generator Tubing Sample Library is a Steam Generator Owners Group-EPRI program whose objective is to compile a library of well-characterized tubing samples to be used for performance evaluation of inspection systems and for training and qualification of signal interpretation systems. The library was created through the preparation of samples intended to replicate degradation encountered in actual field tubes. A limited number of tube segments removed from actual steam generators are included. Degradation categories include wear, pitting and fatigue cracks, as well as stress corrosion cracking (SCC) and intergranular attack (IGA). Eddy current and ultrasonic inspection techniques, along with supplementary radiography, dye penetrant, and optical techniques were used to characterize the library candidates. Advanced computer-aided NDE data collection, analysis and display techniques were used to assess test results. This report provides details of the library program, with major emphasis on the sampling protocol, characterization of degradation and recommendations for the use and future growth of the library. Also included is a compendium of steam generator tube degradation field observation, describing past destructive examinations of tubes removed for inspection from steam generators, and a description of a physical modeling approach, using mercury (metal) to assess the discontinuity characterization capabilities of a pancake-type eddy current probe. Computerized data analysis and display techniques were used to reconstruct the test results in both two-dimensional color-coded maps and three-dimensional pseudo-isometric plots.

  8. Reactance simulation for the defects in steam generator tube with outside ferrite sludge

    SciTech Connect

    Ryu, Kwon-sang; Kima, Yong-il; Son, Derac; Park, Duck-gun; Jung, Jae-kap

    2009-04-01

    A magnetic sludge is partly produced around the tube sheet outside a steam generator due to stress and heat. The sludge with magnetite is one of the important factors affecting eddy current signals. It causes trouble for the safety of the steam generator tubes and is difficult to detect by conventional eddy current methods. A new type of probe is needed to detect the signals for the magnetic sludge. We designed a new U-type yoke which has two kinds of coils--a magnetizing coil and the other a detecting coil--and we simulated the signal induced by the ferromagnetic sludge in the Inconel 600 tube.

  9. Reactance simulation for the defects in steam generator tube with outside ferrite sludge

    NASA Astrophysics Data System (ADS)

    Ryu, Kwon-sang; Son, Derac; Park, Duck-gun; Jung, Jae-kap; Kima, Yong-il

    2009-04-01

    A magnetic sludge is partly produced around the tube sheet outside a steam generator due to stress and heat. The sludge with magnetite is one of the important factors affecting eddy current signals. It causes trouble for the safety of the steam generator tubes and is difficult to detect by conventional eddy current methods. A new type of probe is needed to detect the signals for the magnetic sludge. We designed a new U-type yoke which has two kinds of coils—a magnetizing coil and the other a detecting coil—and we simulated the signal induced by the ferromagnetic sludge in the Inconel 600 tube.

  10. Evaluation and field validation of Eddy-Current array probes for steam generator tube inspection

    SciTech Connect

    Dodd, C.V.; Pate, J.R.

    1996-07-01

    The objective of the Improved Eddy-Current ISI for Steam Generator Tubing 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 describes the design of specialized high-speed 16-coil eddy-current array probes. Both pancake and reflection coils are considered. Test results from inspections using the probes in working steam generators are given. Computer programs developed for probe calculations are also supplied.

  11. Susceptibility of steam generator tubes in secondary conditions: Effects of lead and sulphate

    SciTech Connect

    Gomez Briceno, D.; Garcia, M.S.; Castano, M.L.; Lancha, A.M.

    1997-02-01

    IGA/SCC on the secondary side of steam generators is increasing every year, and represents the cause of some steam generator replacements. Until recently, caustic and acidic environments have been accepted as causes of IGA/SCC, particulary in certain environments: in sludge pile on the tube sheet; at support crevices; in free span. Lead and sulfur have been identified as significant impurities. Present thoughts are that some IGA/SCC at support crevices may have occurred in nearly neutral or mildly alkaline environments. Here the authors present experimental work aimed at studying the influence of lead and sulfur on the behaviour of steam generator tube alloys in different water environments typical of steam generators. Most test results ran for at least 2000 hours, and involved visual and detailed surface analysis during and following the test procedures.

  12. Indicated and actual mass inventory measurements for an inverted U-tube steam generator

    SciTech Connect

    Loomis, G.G.; Plessinger, M.P.; Boucher, T.J.

    1986-01-01

    Results from an experimental investigation of actual versus indicated secondary liquid level in a steam generator at steaming conditions are presented. The experimental investigation was performed in two different small scale U-tube-in-shell steam generators at typical pressurized water reactor operating conditions (5-7 MPa; saturated) in the Semiscale facility. During steaming conditions, the indicated secondary liquid level was found to vary considerably from the actual ''bottled-up'' liquid level. These difference between indicated and actual liquid level are related to the frictional pressure drop associated with the two-phase steaming condition in the riser. Data from a series of bottle-up experiments (Simultaneously, the primary heat source and secondary feed and steam are terminated) are tabulated and the actual liquid level is correlated to the indicated liquid level.

  13. Failure Pressure Estimates of Steam Generator Tubes Containing Wear-type Defects

    SciTech Connect

    Yoon-Suk Chang; Jong-Min Kim; Nam-Su Huh; Young-Jin Kim; Seong Sik Hwang; Joung-Soo Kim

    2006-07-01

    It is commonly requested that steam generator tubes with defects exceeding 40% of wall thickness in depth should be plugged to sustain all postulated loads with appropriate margin. The critical defect dimensions have been determined based on the concept of plastic instability. This criterion, however, is known to be too conservative for some locations and types of defects. In this context, the accurate failure estimation for steam generator tubes with a defect draws increasing attention. Although several guidelines have been developed and are used for assessing the integrity of defected tubes, most of these guidelines are related to stress corrosion cracking or wall-thinning phenomena. As some of steam generator tubes are also failed due to fretting and so on, alternative failure estimation schemes for relevant defects are required. In this paper, three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of steam generator tubes with different defect configurations; elliptical wastage type, wear scar type and rectangular wastage type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of the steam generator tube. After investigating the effect of key parameters such as wastage depth, wastage length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wastage region. Comparison of failure pressures predicted according to the proposed estimation scheme with some corresponding burst test data shows good agreement, which provides a confidence in the use of the proposed equations to assess the integrity of steam generator tubes with wear-type defects. (authors)

  14. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  15. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  16. Development of Performance Demonstration Programs for Steam Generator Tubing Analysts in Korea

    SciTech Connect

    Chan-Hee, Cho; Min-Woo, Nam; Dong-Hyun, Jee; Jee-Hong, Jung; Hee-Jong, Lee; Se-Kyoung, Kim

    2006-07-01

    Korea Electric Power Research Institute developed the performance demonstration programs for non-destructive examination personnel who analyze eddy current data for steam generator tubes of the nuclear power plant. The purpose of these programs is to ensure a continuing uniform knowledge base and skill level for data analysts and contribute to safely operate the nuclear power plant. In Korea, there have been many changes for the non-destructive examination of steam generator tubing in the nuclear power plant such as inspection scope, plugging criteria and qualification requirements. According to the Notice 2004-13 revised by the Ministry of Science and Technology in Korea, the analysts for steam generator tubing shall be qualified as the Qualified Data Analyst, and the Site Specific Performance Demonstration program shall be implemented for them. KEPRI developed these performance demonstration programs and they are being successfully implemented. The analyst performance is expected to be improved by the implementation of theses programs. (authors)

  17. Loads on steam generator tubes during simulated loss-of-coolant accident conditions. Final report. [PWR

    SciTech Connect

    Guerrero, H.N.; Hiestand, J.W.; Rossano, F.V.; Shah, P.K.; Thakkar, J.G.

    1982-11-01

    This report presents the work performed to verify the CEFLASH digital computer code modeling of the hydro-dynamic loads in a steam generator tube during a loss-of-coolant accident (LOCA). The test loop simulated the primary side thermal-hydraulic conditions in an operational nuclear steam generator. The loop consisted of 5 full size double 90/sup 0/ bend tubes and steam generator plena, a pressurizer, a reactor resistance simulator, a heater, a pump, and associated pipes and valves to complete the system. The tubes used were of typical length and the same outside diameter as those used in C-E steam generators. Prototypical supports were provided for the bundle of 5 tubes. Cold leg guillotine breaks were simulated using quick opening valve and rupture disks. Break opening times ranged from less than 1 msec to as much as 67 milliseconds. The loop instrumentation was designed to measure the transient pressure history at various locations and monitor the structural response of the tube to the LOCA hydrodynamic loading. A series of blowdown tests was performed for different operating and boundary conditions. Analytically predicted transient pressure histories and the differential pressure history across the tube span were compared with the experimental data.

  18. Numerical prediction of turbulence-induced steam generator tube vibration: Final report

    SciTech Connect

    Stuhmiller, J.H.

    1988-05-01

    This project investigates promising techniques for predicting turbulent buffeting of tubes leading to tube damage from wear given overall steam generator geometry and operating conditions. The specified overall steam generator operating conditions are used in a model for the steam generator inlet region to evaluate local measures of incoming turbulent flow such as velocity, pressure, turbulence intensity and spectra. A range of models differing in degree of completeness may be used to calculate the incoming flow turbulence. The simplest of the three models is to use a thermal-hydraulic code such as EPRI's ATHOS or PORTHOS code to calculate the steady state flow field (u, v, w and p). Crude, empirical estimates for turbulence intensities and spectra may be deduced from the steady flow results. The best approach, which is chosen for the present study, is Large Eddy Simulation (LES) which gives detailed transient flow results that are in essence a complete description of incoming turbulence. LES results for turbulent flow in the steam generator inlet region provide the necessary local flow conditions for input into tube structural dynamic simulations. This project uses transient thermal-hydraulic analysis of flow within the tube bank to determine the instantaneous, circumferentially integrated force on each tube as a function of position along its axis. The resulting force component time histories provide a complete description of the force imposed on a rigid tube due to the incoming flow turbulence. Tube motion under the action of flow induced forces is determined from models of structural dynamics. This project models one-dimensional motion of the multispan tube including finite tube support clearances and the resulting tube-support impact force. 25 refs., 99 figs., 11 tabs.

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

    SciTech Connect

    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.

  20. Experimental characterization of fluid film effects in various steam generator tube support geometries. Final report

    SciTech Connect

    Haslinger, K.H.

    1995-03-01

    Fluid film characteristics inside cylindrical steam generator tube support holes and near anti-vibration bar supports were determined experimentally. Test results were evaluated and empirical formulations were developed which adequately represent the observed fluid film phenomena. The empirical formulations are suited for incorporation into the ABAQUS computer code which has been developed by Foster Wheeler for EPRI for prediction of the dynamic behavior and work rates of vibrating steam generator tubes with non-linear support characteristics. A short rigid tube was cycled sinusoidally inside special, instrumented tube support samples. Alignment features enabled accurate positioning of the tube, thereby producing either non-contact or impact conditions with known excitation frequency, tube orbit, and amplitude. The complement of measurements consisted of the instantaneous values for tube motion, tube velocity, tube acceleration, contact condition, and the force exchange between tube and support. These measurements were digitized with high sampling rates and subsequently tabulated and graphed. Review of various 2-D and 3-D collages for a water environment at ambient revealed that the fluid film reaction forces, for reasonably large gaps between tube and support, are primarily dependent on tube acceleration, and to a lesser extent on tube velocity. For smooth cylindrical support surfaces there also exists a strong squeeze film effect for small gaps up to impact, and a suction effect during rebound. The squeeze film effect was found to be dependent on the instantaneous gap and tube velocity values. As influenced by the fluid viscosity, the dependency of the fluid reaction force on tube acceleration and on tube velocity was found to vary and was characterized in the experiments for one support clearance condition.

  1. Steam generator tube degradation at the Doel 4 plant influence on plant operation and safety

    SciTech Connect

    Scheveneels, G.

    1997-02-01

    The steam generator tubes of Doel 4 are affected by a multitude of corrosion phenomena. Some of them have been very difficult to manage because of their extremely fast evolution, non linear evolution behavior or difficult detectability and/or measurability. The exceptional corrosion behavior of the steam generator tubes has had its drawbacks on plant operation and safety. Extensive inspection and repair campaigns have been necessary and have largely increased outage times and radiation exposure to personnel. Although considerable effort was invested by the utility to control corrosion problems, non anticipated phenomena and/or evolution have jeopardized plant safety. The extensive plugging and repairs performed on the steam generators have necessitated continual review of the design basis safety studies and the adaptation of the protection system setpoints. The large asymmetric plugging has further complicated these reviews. During the years many preventive and recently also defence measures have been implemented by the utility to manage corrosion and to decrease the probability and consequences of single or multiple tube rupture. The present state of the Doel 4 steam generators remains troublesome and further examinations are performed to evaluate if continued operation until June `96, when the steam generators will be replaced, is justified.

  2. Transmit-receive eddy current probes for defect detection and sizing in steam generator tubes

    SciTech Connect

    Obrutsky, L.S.; Cecco, V.S.; Sullivan, S.P.

    1997-02-01

    Inspection of steam generator tubes in aging Nuclear Generating Stations is increasingly important. Defect detection and sizing, especially in defect prone areas such as the tubesheet, support plates and U-bend regions, are required to assess the fitness-for-service of the steam generators. Information about defect morphology is required to address operational integrity issues, i.e., risk of tube rupture, number of tubes at risk, consequential leakage. A major challenge continues to be the detection and sizing of circumferential cracks. Utilities around the world have experienced this type of tube failure. Conventional in-service inspection, performed with eddy current bobbin probes, is ineffectual in detecting circumferential cracks in tubing. It has been demonstrated in CANDU steam generators, with deformation, magnetite and copper deposits that multi-channel probes with transmit-receive eddy current coils are superior to those using surface impedance coils. Transmit-receive probes have strong directional properties, permitting probe optimization according to crack orientation. They are less sensitive to lift-off noise and magnetite deposits and possess good discrimination to internal defects. A single pass C3 array transmit-receive probe developed by AECL can detect and size circumferential stress corrosion cracks as shallow as 40% through-wall. Since its first trial in 1992, it has been used routinely for steam generator in-service inspection of four CANDU plants, preventing unscheduled shutdowns due to leaking steam generator tubes. More recently, a need has surfaced for simultaneous detection of both circumferential and axial cracks. The C5 probe was designed to address this concern. It combines transmit-receive array probe technology for equal sensitivity to axial and circumferential cracks with a bobbin probe for historical reference. This paper will discuss the operating principles of transmit-receive probes, along with inspection results.

  3. Eddy current signal deconvolution technique for the improvement of steam generator tubing burst pressure predictions.

    SciTech Connect

    Petri, M. C.; Wei, T. Y. C.; Kupperman, D. S.; Reifman, J.; Morman, J. A.

    2000-01-01

    Eddy current techniques are extremely sensitive to the presence of axial cracks in nuclear power plant steam generator tube walls, but they are equally sensitive to the presence of dents, fretting, support structures, corrosion products, and other artifacts. Eddy current signal interpretation is further complicated by cracking geometries more complex than a single axial crack. Although there has been limited success in classifying and sizing defects through artificial neural networks, the ability to predict tubing integrity has, so far, eluded modelers. In large part, this lack of success stems from an inability to distinguish crack signals from those arising from artifacts. We present here a new signal processing technique that deconvolves raw eddy current voltage signals into separate signal contributions from different sources, which allows signals associated with a dominant crack to be identified. The signal deconvolution technique, combined with artificial neural network modeling, significantly improves the prediction of tube burst pressure from bobbin-coil eddy current measurements of steam generator tubing.

  4. Heat and mass transfer in a steam-generating tube

    SciTech Connect

    Doroshchuk, V.Y.

    1990-01-01

    Reliable data on the distribution of phases of steam-water flow in a channel cross-section, on steam and water velocities, and on void fractions are almost non-existent. This paper reports that by assuming that u{sub 1}/u{sub 3} = square root {rho}{prime}/{rho}{prime}{prime} (where u{sub 1} and u{sub 3} are the velocities of steam in the flow core and of a liquid film near a wall, {rho}{prime}{prime} and {rho}{prime} are the densities of corresponding phases) and by using the limiting void fractions x{sup o}{sub lim}, the above-mentioned flow parameters can be determined analytically for the instant of onset of a heat transfer crisis of the second kind.

  5. Transient prediction of 19-tube once-through steam generator by RELAP5/MOD1

    SciTech Connect

    Hassan, Y.A.; Morgan, C.D.

    1982-01-01

    A simulation of Babcock and Wilcox's Alliance Research Center loss-of-feedwater of 19-tube model of once-through steam generator (OTSG) was performed with RELAP5/MOD1 and compared with the experimental data. Acceptable transient scenario was obtained when implementing Biasi and Macbeth critical heat flux correlations.

  6. Proceedings: 1983 Workshop on Primary-Side Stress Corrosion Cracking of PWR Steam Generator Tubing

    SciTech Connect

    1987-11-01

    Utility and vendor representatives from around the world met to share information on stress corrosion cracking of steam generator tubing from the primary side. In 32 presentations, speakers discussed in-plant experience with the phenomenon and related laboratory data. The workshop was the first to present results of remedial stress relief programs.

  7. Proceedings: 1985 Workshop on Primary-Side Stress Corrosion Cracking of PWR Steam Generator Tubing

    SciTech Connect

    1987-06-01

    To date, more than 30 PWRs have reported stress corrosion cracking of steam generator tubing from the primary water side. In 32 presentations, this report offers in-plant and laboratory data on the contributing factors, as well as discussing some promising remedial measures.

  8. Specific features of corrosion damage to heat-transfer tubes of steam generators used at nuclear power stations equipped with VVER-1000 reactors

    NASA Astrophysics Data System (ADS)

    Nemytov, D. S.; Tyapkov, V. F.

    2009-07-01

    Specific features of corrosion damage occurring to the heat-transfer tubes of steam generators used at nuclear power stations equipped with VVER-1000 reactors are considered. The results obtained from metallographic studies of flaws found in samples cut out from steam-generator tubes are analyzed. Regularities with which flaws of steam-generator tubes are distributed over the tube bundle volume are discussed. Approaches for assessing the technical state and remaining service life of steam-generator tubes are presented.

  9. Robotic camera for automatic localization of steam generator tubes in nuclear power stations

    NASA Astrophysics Data System (ADS)

    Cers, Philippe; Garnero, Marie-Agnes

    1994-11-01

    Maintenance of steam generators occupies a substantial proportion of scheduled shutdowns at nuclear power stations. Maintenance operations are broken down into a number of distinct phases; these are performed separately to ensure accountability for the work carried out at each stage, thereby guaranteeing the quality of the maintenance process as a whole. One of these phases, known as `marking,' consists in locating certain tubes in the steam generator tube plate and marking them using a suitable system. The list of tubes for marking may be determined on the basis of prior tests. Marked tubes will undergo subsequent operations as required, such as plugging for example. Clearly, the quality of the marking process will have a significant impact on all subsequent maintenance operations on tubes in the secondary bundle. Present-day marking tools make little use of automation, and over-reliance on human judgement means that the marking phase is liable to error. Moreover, depending on the number of tubes to mark, this phase can be long and fastidious. With these considerations in mind, the EDF Research Division has developed a display system for locating steam generator tubes, with the main purpose of facilitating marking operations. Following an initialization phase, this system (named LUCANER) provides the operator with a simple, reliable and fully automatic method for locating tubes in the tube plate. Besides reducing the risk of error, the system also reduces the time required for the marking phase. The system can also be used for complementary phases involving checks on markings, checks on plugging, etc. In a wider context, it provides visual inspection capabilities over a large part of the bowl.

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

    SciTech Connect

    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.

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

    SciTech Connect

    Comby, R.J.

    1997-02-01

    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.

  12. Evaluation and Reliability Enhancement for ET Data of VVER Steam Generator Tubes

    SciTech Connect

    Kadenko, Ihor; Sakhno, Nadiya; Yermolenko, Ruslan; Anderson, Michael T.; Taylor, Tom T.

    2003-12-01

    Currently, there are three remote automated eddy current inspection systems supplied in Ukraine under international cooperation agreements for examination of steam generators, collectors and steam generator tubing. Since 1966 the field experience in using eddy current testing (ET) for in-service inspection of VVER designed steam generators in Ukraine has shown several advantages over the previous inspection methodology which was an air bubble leakage technique. However, the field experience in using eddy current inspection technology has also shown that inspections are not always reliable. Some nuclear power plants have experienced unplanned shutdowns due to leaks in steam generator tubes, some of which were tested by ET prior to the leak. Therefore, eddy current inspection has shown significant improvement over previous testing techniques, field experience also shows the necessity to improve the ET inspection reliability as applied at Ukraine nuclear power plants. This paper presents the status of efforts by the Ukraine Nondestructive Certification and Training Facility (NDEF) Eddy Current laboratory to improve eddy current inspection in Ukraine.

  13. STEAM GENERATOR FOR NUCLEAR REACTOR

    DOEpatents

    Kinyon, B.W.; Whitman, G.D.

    1963-07-16

    The steam generator described for use in reactor powergenerating systems employs a series of concentric tubes providing annular passage of steam and water and includes a unique arrangement for separating the steam from the water. (AEC)

  14. Tritium permeation through steam generator tubing of helium-cooled ceramic breeder blankets

    SciTech Connect

    Fuetterer, M.; Raepsaet, X.; Proust, E.

    1994-12-31

    The potential sources of tritium contamination of the helium-coolant of ceramic breeder blankets have been evaluated in a previous paper for the specific case of the European BIT DEMO blanket. This evaluation associated with a rough assessment of the permeability to tritium of the tubing of helium-heated steam generators confirmed that the control of tritium losses to the steam circuit is a critical issue for this class of blanket requiring developments in three areas: (1) permeation barriers, (2) tritium recovery processes maintaining a very low concentration in tritiated species in the coolant, and (3) methods for controlling the chemistry of the coolant. Consequently, in order to define the specifications of these developments, a detailed evaluation of the permeability to tritium of helium-heated steam generators (SGs) was performed, which will be reported in this paper. This study includes the definition of the thermal-hydraulic operating conditions of the SGs through thermodynamic cycle calculations, and its thermal-hydraulic design. The obtained geometry, area and temperature profiles along the tubes are then used to estimate, based on relevant permeability data, the tritium permeation through the SG as a function of the composition in tritiated species of the coolant. The implications of these results, in terms of requirements for the considered tritium control methods, will also be discussed on the basis of expected limits in tritium release to the steam circuit.

  15. Proceedings: 1983 Workshop on Secondary-Side Stress Corrosion Cracking and Intergranular Corrosion of PWR Steam Generator Tubing

    SciTech Connect

    1986-03-01

    Participants in this international workshop discussed research investigating mechanisms and propagation rates of intergranular corrosion in PWR steam generators. Laboratory test results, which have been consistent with power plant experience, permitted preliminary definition of corrosion rates in alloy 600 tubing.

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

    SciTech Connect

    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.

  17. Risk assessment of severe accident-induced steam generator tube rupture

    SciTech Connect

    1998-03-01

    This report describes the basis, results, and related risk implications of an analysis performed by an ad hoc working group of the U.S. Nuclear Regulatory Commission (NRC) to assess the containment bypass potential attributable to steam generator tube rupture (SGTR) induced by severe accident conditions. The SGTR Severe Accident Working Group, comprised of staff members from the NRC`s Offices of Nuclear Reactor Regulation (NRR) and Nuclear Regulatory Research (RES), undertook the analysis beginning in December 1995 to support a proposed steam generator integrity rule. The work drew upon previous risk and thermal-hydraulic analyses of core damage sequences, with a focus on the Surry plant as a representative example. This analysis yielded new results, however, derived by predicting thermal-hydraulic conditions of selected severe accident scenarios using the SCDAP/RELAP5 computer code, flawed tube failure modeling, and tube failure probability estimates. These results, in terms of containment bypass probability, form the basis for the findings presented in this report. The representative calculation using Surry plant data indicates that some existing plants could be vulnerable to containment bypass resulting from tube failure during severe accidents. To specifically identify the population of plants that may pose a significant bypass risk would require more definitive analysis considering uncertainties in some assumptions and plant- and design-specific variables. 46 refs., 62 figs., 37 tabs.

  18. STEAM GENERATOR GROUP PROJECT

    SciTech Connect

    Clark, R. A.; Lewis, M

    1985-09-01

    This report is a summary of progress in the Surry Steam Generator Group Project for 1984. Information is presented on the analysis of two baseline eddy current inspections of the generator. Round robin series of tests using standard in-service inspection techniques are described along with some preliminary results. Observations are reported of degradation found on tubing specimens removed from the generator, and on support plates characterized in-situ. Residual stresses measured on a tubing specimen are reported. Two steam generator repair demonstrations are described; one for antivibration bar replacement, and one on tube repair methods. Chemical analyses are shown for sludge samples removed from above the tube sheet.

  19. Status of the steam generator tube circumferential ODSCC degradation experienced at the Doel 4 plant

    SciTech Connect

    Roussel, G.

    1997-02-01

    Since the 1991 outage, the Doel Unit 4 nuclear power plant is known to be affected by circumferential outside diameter intergranular stress corrosion cracking at the hot leg tube expansion transition. Extensive non destructive examination inspections have shown the number of tubes affected by this problem as well as the size of the cracks to have been increasing for the three cycles up to 1993. As a result of the high percentage of tubes found non acceptable for continued service after the 1993 in-service inspection, about 1,700 mechanical sleeves were installed in the steam generators. During the 1994 outage, all the tubes sleeved during the 1993 outage were considered as potentially cracked to some extent at the upper hydraulic transition and were therefore not acceptable for continued service. They were subsequently repaired by laser welding. Furthermore all the tubes not sleeved during the 1993 outage were considered as not acceptable for continued service and were repaired by installing laser welded sleeves. During the 1995 outage, some unexpected degradation phenomena were evidenced in the sleeved tubes. This paper summarizes the status of the circumferential ODSCC experienced in the SG tubes of the Doel 4 plant as well as the other connected degradation phenomena.

  20. Predictions of structural integrity of steam generator tubes under normal operating, accident, an severe accident conditions

    SciTech Connect

    Majumdar, S.

    1997-02-01

    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 was 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 was 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 was developed and validated by tests under various temperature and pressure loadings that can occur during postulated severe accidents.

  1. Code System for Calculating the Nonlinear Transient Behavior of a Natural Circulation U-Tube Steam Generator with Its Main Steam System.

    Energy Science and Technology Software Center (ESTSC)

    2000-04-20

    Version 00 The code is based on a non-linear theoretical model describing the steady-state and transient behavior of a vertical natural-circulation U-tube steam generator together with its main steam system. The steam generator is considered to consist of a heat exchange section, a top plenum, a down-comer region and a main steam system (with a sequence of relief and/or safety valves, isolation, bypass, turbine-trip and turbine-control valves and a steam turbine). Possible perturbations from outsidemore » can be: inlet water temperature, inlet water mass flow and system pressure on the primary side, feedwater temperature, feed-water mass flow and outlet steam mass flow disturbed by actions of the different valves within the main steam system on the secondary side.« less

  2. TMI-2 B-loop steam generator tube sheet loose debris examination and analysis

    SciTech Connect

    Hayner, G O; Hardt, T L

    1989-06-01

    The debris recovered from the upper tube sheet of the TMI-2-B-loop steam generator was analyzed in an effort to determine the concentration and distribution of the chemical and radiochemical species. The debris is of special interest because it is believed to have been transported from the core region sequence between 174 and 192 min after accident initiation when a B-loop reactor coolant pump was restarted. Characterization of five size fractions and 10 of the largest particles was accomplished by destructive (chemical, radiochemical, metallography, and SEM/EDS) and nondestructive (photographic examination and density) methods of analysis. 2 refs., 7 figs., 14 tabs.

  3. Pitting of steam-generator tubing alloys in solutions containing thiosulfate and sulfate or chloride.

    PubMed

    Zhang, William; Carcea, Anatolie G; Newman, Roger C

    2015-01-01

    The pitting of nuclear steam generator tubing alloys 600, 690 and 800 was studied at 60 °C using dilute thiosulfate solutions containing excess sulfate or (for Alloy 600) chloride. A potentiostatic scratch method was used. In sulfate solutions, all alloys pitted at low potentials, reflecting their lack of protective Mo. The alloys demonstrated the most severe pitting at a sulfate : thiosulfate concentration ratio of ∼40. Alloy 600 pitted worst at a chloride : thiosulfate ratio of ∼2000. The results are interpreted through the mutual electromigration of differently charged anions into a pit nucleus, and differences in the major alloy component. PMID:25898311

  4. Root causes of intergranular attack in an operating nuclear steam generator tube

    NASA Astrophysics Data System (ADS)

    Hur, Do Haeng; Lee, Deok Hyun; Choi, Myung Sik; Song, Myung Ho; Han, Jung Ho

    2008-04-01

    This paper reports the secondary side intergranular attack of an Alloy 600 tube, which was located within sludge piles in the hot-leg side of an operating nuclear steam generator. Carbide distribution along the grain boundaries and chromium depletion were analyzed using optical microscopy and transmission electron microscopy. Local crevice chemistry in contact with the defect was also assessed from the hideout return test data and oxide film analysis results using energy dispersive spectroscopy. The main causes of this defect are discussed based on the microstructure, local chemistry and operation temperature.

  5. Investigation of eddy current examination on OD fatigue crack for steam generator tubes

    NASA Astrophysics Data System (ADS)

    Kong, Yuying; Ding, Boyuan; Li, Ming; Liu, Jinhong; Chen, Huaidong; Meyendorf, Norbert G.

    2015-03-01

    The opening width of fatigue crack was very small, and conventional Bobbin probe was very difficult to detect it in steam generator tubes. Different sizes of 8 fatigue cracks were inspected using bobbin probe rotating probe. The analysis results showed that, bobbin probe was not sensitive for fatigue crack even for small through wall crack mixed with denting signal. On the other hand, the rotating probe was easily to detect all cracks. Finally, the OD phase to depth curve for fatigue crack using rotating probe was established and the results agreed very well with the true crack size.

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

    SciTech Connect

    Lozano, J.; Bollini, G.J.

    1997-02-01

    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.

  7. Analysis of pulsed eddy current data using regression models for steam generator tube support structure inspection

    NASA Astrophysics Data System (ADS)

    Buck, J. A.; Underhill, P. R.; Morelli, J.; Krause, T. W.

    2016-02-01

    Nuclear steam generators (SGs) are a critical component for ensuring safe and efficient operation of a reactor. Life management strategies are implemented in which SG tubes are regularly inspected by conventional eddy current testing (ECT) and ultrasonic testing (UT) technologies to size flaws, and safe operating life of SGs is predicted based on growth models. ECT, the more commonly used technique, due to the rapidity with which full SG tube wall inspection can be performed, is challenged when inspecting ferromagnetic support structure materials in the presence of magnetite sludge and multiple overlapping degradation modes. In this work, an emerging inspection method, pulsed eddy current (PEC), is being investigated to address some of these particular inspection conditions. Time-domain signals were collected by an 8 coil array PEC probe in which ferromagnetic drilled support hole diameter, depth of rectangular tube frets and 2D tube off-centering were varied. Data sets were analyzed with a modified principal components analysis (MPCA) to extract dominant signal features. Multiple linear regression models were applied to MPCA scores to size hole diameter as well as size rectangular outer diameter tube frets. Models were improved through exploratory factor analysis, which was applied to MPCA scores to refine selection for regression models inputs by removing nonessential information.

  8. Continuous-wave radar to detect defects within heat exchangers and steam generator tubes.

    SciTech Connect

    Nassersharif, Bahram (New Mexico State University, Las Cruces, NM); Caffey, Thurlow Washburn Howell; Jedlicka, Russell P.; Garcia, Gabe V. (New Mexico State University, Las Cruces, NM); Rochau, Gary Eugene

    2003-01-01

    A major cause of failures in heat exchangers and steam generators in nuclear power plants is degradation of the tubes within them. The tube failure is often caused by the development of cracks that begin on the outer surface of the tube and propagate both inwards and laterally. A new technique was researched for detection of defects using a continuous-wave radar method within metal tubing. The experimental program resulted in a completed product development schedule and the design of an experimental apparatus for studying handling of the probe and data acquisition. These tests were completed as far as the prototypical probe performance allowed. The prototype probe design did not have sufficient sensitivity to detect a defect signal using the defined radar technique and did not allow successful completion of all of the project milestones. The best results from the prototype probe could not detect a tube defect using the radar principle. Though a more precision probe may be possible, the cost of design and construction was beyond the scope of the project. This report describes the probe development and the status of the design at the termination of the project.

  9. Steam condensation and liquid hold-up in steam generator U-tubes during oscillatory natural circulation

    SciTech Connect

    De Santi, G.F.; Mayinger, F.

    1990-01-01

    In many accident scenarios, natural circulation is an important heat transport mechanism for long-term cooling of light water reactors. In the event of a small pipe break, with subsequent loss of primary cooling fluid loss-of-coolant accident (LOCA), or under abnormal operating conditions, early tripping of the main coolant pumps can be actuated. Primary fluid flow will then progress from forced to natural convection. Understanding of the flow regimes and heat-removal mechanisms in the steam generators during the entire transient is of primary importance to safety analysis. Flow oscillations during two-phase natural circulation experiments for pressurized water reactors (PWRs) with inverted U-tube steam generators occur at high pressure and at a primary inventory range between two-phase circulation and reflex heat removal. This paper deals with the oscillatory flow behavior that was observed in the LOBI-MOD2 facility during the transition period between two-phase natural circulation and reflex condensation.

  10. Steam generator performance degradation

    SciTech Connect

    Lovett, J.T.; Dow, B.L. )

    1991-09-01

    A survey was conducted to determine the range and severity of steam generator performance degradation effects experienced by PWRs in the United States. The survey results were tabulated and correlated with steam generator age and design. Operating experience at several PWRs was examined in detail. The operating experience at US PWRs was compared to that of PWRs in Japan and Germany. Possible causes for the performance degradation were postulated and evaluated. The sensitivity of steam generator output pressure to changes in various parameters (such as fouling factor, average reactor coolant temperature, and percentage of steam generator tubes plugged) was calculated. These calculations were used in the evaluation of possible causes of steam generator performance degradation. Several deposit exfoliation scenarios were evaluated in terms of the calculated effect on fouling factor trends and associated steam generator output pressure trends. 15 refs., 32 figs., 7 tabs.

  11. Performance demonstration tests for eddy current inspection of steam generator tubing

    SciTech Connect

    Kurtz, R.J.; Heasler, P.G.; Anderson, C.M.

    1996-05-01

    This report describes the methodology and results for development of performance demonstration tests for eddy current (ET) inspection of steam generator tubes. Statistical test design principles were used to develop the performance demonstration tests. Thresholds on ET system inspection performance were selected to ensure that field inspection systems would have a high probability of detecting and and correctly sizing tube degradation. The technical basis for the ET system performance thresholds is presented in detail. Statistical test design calculations for probability of detection and flaw sizing tests are described. A recommended performance demonstration test based on the design calculations is presented. A computer program for grading the probability of detection portion of the performance demonstration test is given.

  12. Thermal EOR: a review of insulated tubing and downhole steam generator materials evaluations

    SciTech Connect

    Aeschliman, D.P.; Moreno, J.B.; Marshall, B.W.

    1985-01-01

    A downhole steam generator post-field-test materials study program has been completed. For the particular combustor design that was tested, wall temperatures as high as 1500/sup 0/F (816/sup 0/C) were observed. Significant variations with respect to time and position were also observed. The CoCrAlY and FeCrAlY alloys appeared to suffer little chemical attack, but both were damaged by what we conclude to be thermally-induced stresses. This mechanism also appears to be a dominant component in the corrosive attack suffered by 310 SS. It was concluded that very thin combustor walls are required in this design to avoid the potentially damaging ''alternating plasticity'' and ''ratcheting'' regimes of cyclic thermal stress. A study of wellbore heat losses from steam injection wells equipped with insulated tubing and conventional (uninsulated) couplings was completed. This study has confirmed that overall wellbore heat loss is substantially higher than expected for good quality tubing with uninsulated couplings if the well bore is wet - an effect attributed to ''wellbore refluxing.'' The temperature at which wellbore refluxing occurs is the boiling point of the annulus water, and is consequently a controllable function of annulus pressure. Reduction of annulus pressure is one means of decreasing the heat loss associated with the refluxing process. Another method is the use of insulated couplings which provided a sufficient thermal resistance to prevent boiling, and hence refluxing, even when totally immersed. Finally, the observation that steady wellbore refluxing can occur without visible steam venting at the wellhead demonstrates that operators should not assume a dry, nonrefluxing annulus based solely on the observed absence of venting. 9 refs., 7 figs.

  13. On the probability of exceeding allowable leak rates through degraded steam generator tubes

    SciTech Connect

    Cizelj, L.; Sorsek, I.; Riesch-Oppermann, H.

    1997-02-01

    This paper discusses some possible ways of predicting the behavior of the total leak rate through the damaged steam generator tubes. This failure mode is of special concern in cases where most through-wall defects may remain In operation. A particular example is the application of alternate (bobbin coil voltage) plugging criterion to Outside Diameter Stress Corrosion Cracking at the tube support plate intersections. It is the authors aim to discuss some possible modeling options that could be applied to solve the problem formulated as: Estimate the probability that the sum of all individual leak rates through degraded tubes exceeds the predefined acceptable value. The probabilistic approach is of course aiming at reliable and computationaly bearable estimate of the failure probability. A closed form solution is given for a special case of exponentially distributed individual leak rates. Also, some possibilities for the use of computationaly efficient First and Second Order Reliability Methods (FORM and SORM) are discussed. The first numerical example compares the results of approximate methods with closed form results. SORM in particular shows acceptable agreement. The second numerical example considers a realistic case of NPP in Krsko, Slovenia.

  14. Remote field eddy current technique applied to the inspection of nonmagnetic steam generator tubes

    NASA Astrophysics Data System (ADS)

    Shin, Young-Kil; Chung, Tae-Eon; Lord, William

    2001-04-01

    As steam generator (SG) tubes have aged, new and subtle degradations have appeared. Most of them start growing from outside the tubes. Since outer diameter defects might not be detected by conventional eddy current testing due to skin effect phenomena, this paper studies the feasibility of using the remote field eddy current (RFEC) technique, which has shown equal sensitivity to inner diameter (ID) and outer diameter (OD) defects in ferromagnetic pipe inspection. Finite element modeling studies show that the operating frequency needs to be increased up to a few hundred kHz in order for RFEC effects to occur in the nonmagnetic SG tube. The proper distance between exciter and sensor coils is also found to be 1.5 OD, which is half of the distance used in ferromagnetic pipe inspection. The resulting defect signals show equal sensitivity to ID and OD defects. These results demonstrate superior capability of the proposed RFEC probe compared to the differential ECT probe in detecting OD defects.

  15. Multiloop integral system test (MIST): Test Group 34, Steam generator tube rupture

    SciTech Connect

    Gloudemans, J.R. . Nuclear Power Div.)

    1989-07-01

    The multiloop integral system test (MIST) is part of a multiphase program started in 1983 to address small-break loss-of-coolant accidents (SBLOCAs) specific to Babcock and Wilcox designed plants. MIST is sponsored by the US Nuclear Regulatory Commission, the Babcock Wilcox Owners Group, the Electric Power Research Institute, and Babcock and Wilcox. The unique features of the Babcock and Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral facilities to address the thermal-hydraulic SBLOCA questions. MIST and two other supporting facilities were specifically designed and constructed for this program, and an existing facility--the Once Through Integral System (OTIS)--was also used. Data from MIST and the other facilities will be used to benchmark the adequacy of system codes, such as RELAP5 and TRAC, for predicting abnormal plant transients. The MIST program is reported in 11 volumes. The program is summarized in Volume 1; Volumes 2 through 8 describes groups of tests by test type; Volume 9 presents inter-group comparisons; Volume 10 provides comparisons between the calculations of RELAP5/MOD2 and MIST observations, and Volume 11 presents the later Phase 4 tests. This Volume 6 pertains to Test Group 34, Steam Generator Tube Rupture. The specifications, conduct, observations, and results of these tests are described. 6 refs., 189 figs., 16 tabs.

  16. The Thermal Hydraulics of Tube Support Fouling in Nuclear Steam Generators

    SciTech Connect

    Rummens, Helena E.C.; Rogers, J.T.; Turner, C.W.

    2004-12-15

    It is hypothesized that the thermal-hydraulic environment plays a role in the fouling of tube supports in nuclear steam generators. Experiments were performed to simulate the thermal-hydraulic environment near various designs of supports. Pressure loss, local velocity, turbulence intensity, and local void fraction were measured to characterize the effect of the support. Fouling mechanisms specific to supports were inferred from these experimental data and from actual steam generator inspection results. An analytical model was developed to predict the rate of particulate deposition on the supports, to better understand the complex processes involved.This paper presents the following set of tools for assessing the fouling propensity of a given support design: (1) proposed fouling mechanisms, (2) criteria for support fouling propensity, (3) correlation of fouling with parameters such as mass flux and quality, (4) descriptions of experimental tools such as flow visualization and measurement of pressure-loss profiles, and (5) analytical tools.An important conclusion from this and our previous work is that the fouling propensity is greater with broached support plates, both trefoil and quatrefoil, than with lattice bar supports and formed bar supports, in which significant cross flows occur.

  17. Neural network inversion of synthetic eddy current testing signals from flaws in steam generator tubes

    NASA Astrophysics Data System (ADS)

    Song, S. J.; Kim, C. H.; Shin, Y. K.; Lee, H. B.; Park, Y. W.; Yim, C. J.

    2001-04-01

    This paper reports our recent endeavor to develop automated, systematic inversion tools by the novel combination of neural networks and finite element modeling for eddy current flaw characterization in steam generator tubes. Specifically, this paper describes 1) development of the finite element models that can simulate synthetic ECT signals from axisymmetric flaws with arbitrary cross-sections, 2) construction of databases with abundant flaw signals, 3) implementation of effective feature extraction software and proposition of feature selection criteria, and finally 4) development of inversion tools by use of two neural networks for flaw classification and sizing. In addition, this paper also presents the performance of the proposed inversion tools for solving two sample problems: classification of flaws with non-symmetric cross-sections, and classification and sizing of flaws with tip variation.

  18. Life Estimation of PWR Steam Generator U-Tubes Subjected to Foreign Object-Induced Fretting Wear

    SciTech Connect

    Jo, Jong Chull; Jhung, Myung Jo; Kim, Woong Sik; Kim, Hho Jung

    2005-10-15

    This paper presents an approach to the remaining life prediction of steam generator (SG) U-tubes, which are intact initially, subjected to fretting-wear degradation due to the interaction between a vibrating tube and a foreign object in operating nuclear power plants. The operating SG shell-side flow field conditions are obtained from a three-dimensional SG flow calculation using the ATHOS3 code. Modal analyses are performed for the finite element models of U-tubes to get the natural frequency, corresponding mode shape, and participation factor. The wear rate of a U-tube caused by a foreign object is calculated using the Archard formula, and the remaining life of the tube is predicted. Also discussed in this study are the effects of the tube modal characteristics, external flow velocity, and tube internal pressure on the estimated results of the remaining life of the tube.

  19. Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes ; Revised September 3, 2003

    SciTech Connect

    Gary E. Rochau and Thurlow W.H. Caffey, Sandia National Laboratories, Albuquerque, NM 87185-0740; Bahram Nassersharif and Gabe V. Garcia, Department of Mechanical Engineering, New Mexico State University, Las Cruces, NM 88003-8001; Russell P. Jedlicka, Klipsch School of Electrical and Computer Engineering, New Mexico State University, Las Cruces, NM 88003-8001

    2003-05-01

    OAK B204 Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes ; Revised September 3, 2003. A major cause of failures in heat exchangers and steam generators in nuclear power plants is degradation of the tubes within them. The tube failure is often caused by the development of cracks that begin on the outer surface of the tube and propagate both inwards and laterally. A new technique was researched for detection of defects using a continuous-wave radar method within metal tubing. The technique is 100% volumetric, and may find smaller defects, more rapidly, and less expensively than present methods. The project described in this report was a joint development effort between Sandia National Laboratories (SNL) and New Mexico State University (NMSU) funded by the US Department of Energy. The goal of the project was to research, design, and develop a new concept utilizing a continuous wave radar to detect defects inside metallic tubes and in particular nuclear plant steam generator tubing. The project was divided into four parallel tracks: computational modeling, experimental prototyping, thermo-mechanical design, and signal detection and analysis.

  20. SCC analysis of Alloy 600 tubes from a retired steam generator

    NASA Astrophysics Data System (ADS)

    Hwang, Seong Sik; Kim, Hong Pyo

    2013-09-01

    Steam generators (SG) equipped with Alloy 600 tubes of a Korean nuclear power plants were replaced with a new one having Alloy 690 tubes in 1998 after 20 years of operation. To set up a guide line for an examination of the other SG tubes, a metallographic examination of the defected tubes was carried out. A destructive analysis on 71 tubes was addressed, and a relation among the stress corrosion crack (SCC) defect location, defect depth, and location of the sludge pile was obtained. Tubes extracted from the retired SG were transferred to a hot laboratory. Detailed nondestructive analysis examinations were taken again at the laboratory, and the tubes were then destructively examined. The types and sizes of the cracks were characterized. The location and depth of the SCC were evaluated in terms of the location and height of the sludge. Most axial cracks were in the sludge pile, whereas the circumferential ones were around the top of the tube sheet (TTS) or below the TTS. Average defect depth of the axial cracks was deeper than that of the circumferential ones. Axial cracks at tube support plate (TSP) seem to be related with corrosion/sludge in crevice like at the TTS region. Circumferential cracks at TSP seem to be caused by tube denting at the upper part of the TSP. Tubes not having clear ECT signals for quantifying an ECT data-base. Tubes having no ECT signal. Tubes with a large ECT signal. Tubes with various types and sizes of flaws (primary water stress corrosion cracking (PWSCC), outside diameter stress corrosion cracking (ODSCC), Pit). Tubes with distinct PWSCC or ODSCC. Tubes were extracted from the RSG based on the field ECT with the criteria, and transferred to a hot laboratory at the Korea Atomic Energy Research Institute (KAERI) for destructive examination. A comprehensive ECT inspection was performed again at the hot laboratory to confirm the location of the cracks obtained from a field inspection. These exact locations of the defects were marked on the

  1. A simplified mathematical model of a U-tube steam generator under variable load conditions

    NASA Astrophysics Data System (ADS)

    Laskowski, Rafał; Lewandowski, Janusz

    2013-09-01

    A steam generator in a nuclear power plant with a light water reactor is a heat exchanger, in which the heat is being transferred from the primary to the secondary loop (it links the primary and secondary loops). When the power plant is running, the inlet parameters (temperatures and mass flow rates) on both sides of the steam generator can change. It is important to know how the changes of these parameters affect the steam generator performance. The complexity of the processes taking place in the steam generator makes it difficult to create a simulator reflecting its performance under changed conditions. In order to simplify the task, the steam generator was considered as a `black box' with the aim of examining how the changes of the inlet parameters affect the changes of the outlet ones. On the basis of the system (steam generator) response, a simple mathematical model of the steam generator under variable load conditions was proposed. In the proposed model, there are two dimensionless parameters and three constant coefficients. A linear relation between these dimensionless parameters was obtained. The correctness of the model was verified against the data obtained with a steam generator simulator for European Pressured Reactor and AP-600 reactors. A good agreement between the proposed model and the simulator data was achieved.

  2. Nonuniform steam generator U-tube flow distribution during natural circulation tests in ROSA-IV large scale test facility

    SciTech Connect

    Kukita, Y.; Nakamura, H.; Tasaka, K. ); Chauliac, C. )

    1988-08-01

    Natural circulation experiments were conducted in a large-scale (1/48 scale in volume) full-height simulator of a Westinghouse-type pressurized water reactor. This facility has two steam generators each containing 141 full-size U-tubes of 9 different heights. Transition of the natural circulation mode was observed in the experiments as the primary of side mass inventory was decreased. Three major circulation modes were observed: single-phase liquid natural circulation, two-phase natural circulation, and reflux condensation. For all these circulation modes, and during the transitions between the modes, the mass flow distribution among the steam generator U-tubes was significantly nonuniform. The longer U-tubes indicated reversed flow at higher primary side mass inventories and also tended to empty earlier than the shorter U-tubes when the primary side mass inventory was decreased.

  3. Coincident steam generator tube rupture and stuck-open safety relief valve carryover tests: MB-2 steam generator transient response test program

    SciTech Connect

    Garbett, K; Mendler, O J; Gardner, G C; Garnsey, R; Young, M Y

    1987-03-01

    In PWR steam generator tube rupture (SGTR) faults, a direct pathway for the release of radioactive fission products can exist if there is a coincident stuck-open safety relief valve (SORV) or if the safety relief valve is cycled. In addition to the release of fission products from the bulk steam generator water by moisture carryover, there exists the possibility that some primary coolant may be released without having first mixed with the bulk water - a process called primary coolant bypassing. The MB-2 Phase II test program was designed specifically to identify the processes for droplet carryover during SGTR faults and to provide data of sufficient accuracy for use in developing physical models and computer codes to describe activity release. The test program consisted of sixteen separate tests designed to cover a range of steady-state and transient fault conditions. These included a full SGTR/SORV transient simulation, two SGTR overfill tests, ten steady-state SGTR tests at water levels ranging from very low levels in the bundle up to those when the dryer was flooded, and three moisture carryover tests without SGTR. In these tests the influence of break location and the effect of bypassing the dryer were also studied. In a final test the behavior with respect to aerosol particles in a dry steam generator, appropriate to a severe accident fault, was investigated.

  4. The Streaming Potential Generated by Flow of Wet Steam in Capillary Tubes

    SciTech Connect

    Marsden, S.S. Jr.; Tyran, Craig K.

    1986-01-21

    For a constant pressure differential, the flow of wet steam generated electric potentials which increased with time and did not reach equilibrium values. These potentials were found to increase to values greater than 100 volts. The reason for this kind of potential build-up behavior was the presence of tiny flowing water slugs which were interspersed with electrically nonconductive steam vapor slugs. The measured electric potential for wet steam increased with pressure differential, but the relationship was not linear. The increase in potential with pressure drop was attributed both to an increase in fluid flow rate and changes in the wet steam quality.

  5. A pulsed eddy current probe for inspection of support plates from within Alloy-800 steam generator tubes

    SciTech Connect

    Krause, T. W.; Babbar, V. K.; Underhill, P. R.

    2014-02-18

    Support plate degradation and fouling in nuclear steam generators (SGs) can lead to SG tube corrosion and loss of efficiency. Inspection and monitoring of these conditions can be integrated with preventive maintenance programs, thereby advancing station-life management processes. A prototype pulsed eddy current (PEC) probe, targeting inspection issues associated with SG tubes in SS410 tube support plate structures, has been developed using commercial finite element (FE) software. FE modeling was used to identify appropriate driver and pickup coil configurations for optimum sensitivity to changes in gap and offset for Alloy-800 SG tubes passing through 25 mm thick SS410 support plates. Experimental measurements using a probe that was manufactured based on the modeled configuration, were used to confirm the sensitivity of differential PEC signals to changes in relative position of the tube within the tube support plate holes. Models investigated the effect of shift and tilt of tube with respect to hole centers. Near hole centers and for small shifts, modeled signal amplitudes from the differentially connected coil pairs were observed to change linearly with tube shift. This was in agreement with experimentally measured TEC coil response. The work paves the way for development of a system targeting the inspection and evaluation of support plate structures in steam generators.

  6. A pulsed eddy current probe for inspection of support plates from within Alloy-800 steam generator tubes

    NASA Astrophysics Data System (ADS)

    Krause, T. W.; Babbar, V. K.; Underhill, P. R.

    2014-02-01

    Support plate degradation and fouling in nuclear steam generators (SGs) can lead to SG tube corrosion and loss of efficiency. Inspection and monitoring of these conditions can be integrated with preventive maintenance programs, thereby advancing station-life management processes. A prototype pulsed eddy current (PEC) probe, targeting inspection issues associated with SG tubes in SS410 tube support plate structures, has been developed using commercial finite element (FE) software. FE modeling was used to identify appropriate driver and pickup coil configurations for optimum sensitivity to changes in gap and offset for Alloy-800 SG tubes passing through 25 mm thick SS410 support plates. Experimental measurements using a probe that was manufactured based on the modeled configuration, were used to confirm the sensitivity of differential PEC signals to changes in relative position of the tube within the tube support plate holes. Models investigated the effect of shift and tilt of tube with respect to hole centers. Near hole centers and for small shifts, modeled signal amplitudes from the differentially connected coil pairs were observed to change linearly with tube shift. This was in agreement with experimentally measured TEC coil response. The work paves the way for development of a system targeting the inspection and evaluation of support plate structures in steam generators.

  7. Assessment of the leak tightness integrity of the steam generator tubes affected by ODSCC at the tube support plates

    SciTech Connect

    Cuvelliez, Ch.; Roussel, G.

    1997-02-01

    An EPRI report gives a method for predicting a conservative value of the total primary-to-secondary leak rate which may occur during, a postulated steam generator depressurization accident such as a Main Steam Line Break (MSLB) in a steam generator with axial through-wall ODSCC at the TSP intersections. The Belgian utility defined an alternative method deviating somewhat from the EPRI method. When reviewing this proposed method, the Belgian safety authorities performed some calculations to investigate its conservatism. This led them to recommend some modifications to the EPRI method which should reduce its undue conservatism while maintaining the objective of conservatism in the offsite dose calculations.

  8. Comparison of a RELAP5/MOD2 posttest calculation to the data during the recovery portion of a semiscale single-tube steam generator tube rupture experiment

    SciTech Connect

    Chapman, J.C.

    1986-09-01

    This report discusses the comparisons of a RELAP5 posttest calculation of the recovery portion of the Semiscale Mod-2B test S-SG-1 to the test data. The posttest calculation was performed with the RELAP5/MOD2 cycle 36.02 code without updates. The recovery procedure that was calculated mainly consisted of secondary feed and steam using auxiliary feedwater injection and the atmospheric dump valve of the unaffected steam generator (the steam generator without the tube rupture). A second procedure was initiated after the trends of the secondary feed and steam procedure had been established, and this was to stop the safety injection that had been provided by two trains of both the charging and high pressure injection systems. The Semiscale Mod-2B configuration is a small scale (1/1705), nonnuclear, instrumented, model of a Westinghouse four-loop pressurized water reactor power plant. S-SG-1 was a single-tube, cold-side, steam generator tube rupture experiment. The comparison of the posttest calculation and data included comparing the general trends and the driving mechanisms of the responses, the phenomena, and the individual responses of the main parameters.

  9. Finite element modeling of wall-loss sizing in a steam generator tube using a pulsed eddy current probe

    NASA Astrophysics Data System (ADS)

    Babbar, V. K.; Lepine, B.; Buck, J.; Underhill, P. R.; Morelli, J.; Krause, T. W.

    2015-03-01

    Inspection of steam generator (SG) tubes by conventional eddy current may, in general, involve analysis of indications from volumetric wall loss, cracks, fouling and support-plate degradation; however, it may be difficult to size or quantify effects from support-to-tube gap and tube tilt, especially in the presence of support plates. Pulsed eddy current (PEC) technology is being developed to investigate such complex tube and flaw geometries. The present work employs finite element modeling to investigate the effectiveness of PEC in identifying and sizing the outer diameter wall-loss in SG tubes. The signals analyzed using a modified principal components analysis (PCA) method reveal the potential success of a PEC-PCA combination to produce scores that can be used to size the wall-loss in the presence of support plates. The modeling results are in good agreement with experimental observations.

  10. The thermalhydraulics of tube-support fouling in nuclear steam generators

    NASA Astrophysics Data System (ADS)

    Rummens, Helena Elisabeth Cornelia

    Nuclear steam generators (SGs) world-wide have experienced a wide variety of problems, of which a recent one has been fouling and blockage of the supports that restrain the SG primary-coolant tubes. Water chemistry and operating conditions are known to influence fouling, and it is hypothesized that the thermal and hydraulic environments near a support also play a role. The work presented here endeavours to show the effect of support design on this environment and hence on fouling. Experiments were performed to simulate the thermalhydraulic environment near various designs of tube supports. Air/water mixtures were useful in showing the hydraulic flow patterns, while Freon-11 vapour/liquid mixtures showed thermal effects. Measurements of pressure loss, local velocity, and local void fraction were also made to quantitatively characterize the effect of the support. A computer program, called TSFOUL, was coded to predict deposit thicknesses in and near a support. Larger codes used for such predictions in industry have been unable to predict blockage of supports, hence the need for support- specific models. TSFOUL has the same classic particle deposition models as in the larger codes, but considers additional factors such as stagnation zones and surfaces normal to the flow. The fouling mechanisms specific to supports were inferred from SG inspections and from experimental flow patterns, and measured values helped to make the models more quantitative. While limited by a lack of good validation data, TSFOUL was able to predict reasonable deposition patterns, and helped to understand the complex interaction between different mechanisms. The net product is a set of tools for assessing the fouling propensity of a given tube-support design: (1)proposed fouling mechanisms, (2)criteria for support fouling propensity, (3)correlation of fouling with mass flux and quality, (4)experimental tools such as flow visualization and measurement of pressure-loss profiles, and (5)analytical

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

    SciTech Connect

    Not Available

    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).

  12. Failure behavior of internally pressurized flawed and unflawed steam generator tubing at high temperatures -- Experiments and comparison with model predictions

    SciTech Connect

    Majumdar, S.; Shack, W.J.; Diercks, D.R.; Mruk, K.; Franklin, J.; Knoblich, L.

    1998-03-01

    This report summarizes experimental work performed at Argonne National Laboratory on the failure of internally pressurized steam generator tubing at high temperatures ({le} 700 C). A model was developed for predicting failure of flawed and unflawed steam generator tubes under internal pressure and temperature histories postulated to occur during severe accidents. The model was validated by failure tests on specimens with part-through-wall axial and circumferential flaws of various lengths and depths, conducted under various constant and ramped internal pressure and temperature conditions. The failure temperatures predicted by the model for two temperature and pressure histories, calculated for severe accidents initiated by a station blackout, agree very well with tests performed on both flawed and unflawed specimens.

  13. Substantiation of recommendations for ensuring the design service life of heat-transfer tubes used in a PGV-1000MKP steam generator

    NASA Astrophysics Data System (ADS)

    Popadchuk, V. S.; Trunov, N. B.; Brykov, S. I.; Zhukov, R. Yu.; Tupikov, R. A.; Seleznev, A. V.; Popkov, R. I.; Metal'Nikov, M. S.; Styazhkin, P. S.; Karzov, G. P.; Suvorov, S. A.

    2011-03-01

    We present the results obtained from tests and studies carried out on the model of tube bundles for a PGV-1000 horizontal steam generator that were conducted for experimentally substantiating the design service life of a steam generator tube bundle intended for use at new nuclear power stations equipped with a PGV-1000MKP steam generator. Measures taken to minimize the incipience and development of local corrosion damage to the heat-transfer tubes and ensure their design service life are substantiated and confirmed.

  14. Wear behavior of 2-1/4 Cr-1 Mo tubing against alloy 718 tube-support material in sodium-cooled steam generators

    SciTech Connect

    Wilson, W L

    1983-05-01

    A series of prototypic steam generator 2-{1/4} Cr-1 Mo tube/alloy 718 tube support plate wear tests were conducted in direct support of the Westinghouse Nuclear Components Division -- Breeder Reactor Components Project Large Scale steam Generator design. The initial objective was to verify the acceptable wear behavior of softer, over-aged'' alloy 718 support plate material. For all interfaces under all test conditions, resultant wear damage was adhesive in nature with varying amounts of 2-{1/4} Cr-1 Mo tube material being adhesively transferred to the alloy 718 tube supports. Maximum tube wear depths exceeded the initially established design allowable limit of 127 {mu}m (.005 in.) at 17 of the 18 interfaces tested. A decrease in contact stresses produced acceptable tube wear depths below a readjusted maximum design allowable value of 381 {mu}m (.015 in.). Additional conservatisms associated with the simulation of a 40-year lifetime of rubbing in a one-week laboratory test provided further confidence that the 381 {mu}m maximum tube wear allowance would not be exceeded in service. Softer, over-aged'' alloy 718 material was found to produce slightly less wear damage on 2-{1/4} Cr-1 Mo tubing than fully age hardened material. Also, air formed oxide films on the alloy 718 reduced initial tube wear and delayed the onset of adhesive surface damage. However, at high surface stress levels, these films were not sufficiently stable to provide adequate long term protection from adhesive wear. The results of the present work and those of previous test programs suggest that the successful in-sodium tribological performance of 2-{1/4} Cr-1 Mo/alloy 718 rubbing couples is dependent upon the presence of lubricative surface films, such as oxides and/or surface reaction or deposition products. 11 refs., 13 figs., 4 tabs.

  15. CRBRP modular steam generator tube-to-tubesheet and shell-closure welding

    SciTech Connect

    Viri, D.P.

    1982-01-01

    The original Modular Steam Generator (MSG), whiand inh was designed, built, and tested by the Energy Systems Group (ESG) of Rockwell International, was a departure from conventional boilers or heat exchangers. The design was a hockeystick concept - the upper section of the generator is curved 90/sup 0/. Factors affecting operating parameters were considered and incorporated in the original MSG design. The MSG was fully instrumented and functionally tested at the Energy Technology Engineering Center at Rockwell. The MSG steamed continuously for over 4000 h, and at the conclusion of the 9000-h test cycle, it was systematically dismantled and examined for wear to critical components. This paper explains the solutions to several manufacturing challenges presented by the unique design of the MSG.

  16. Steady-state and transient prediction of a 19-tube once-through steam generator using RELAP5/MOD1

    SciTech Connect

    Hassan, Y.A.; Morgan, C.D.

    1983-01-01

    Comparisons of the predictions of RELAP5/MOD1 to data obtained from a 19-tube model of a once-through steam generator (OTSG) were performed. The initial results were not satisfactory since the predicted outlet steam temperature was much too low. This discrepancy was traced to the inappropriate use of the modified Zuber critical heat flux (CHF) correlation for the conditions occurring during integral economizer OTSG operation. A study of available low-flow CHF correlations was performed that showed that either the Macbeth or Biasi correlations used in conjunction with RELAP5/MOD1 would produce good agreement with both the steadystate and transient data for the integral economizertype OTSG. The Macbeth correlation was the best for the OTSG with a recirculation path; however, it was not entirely satisfactory due to a slight delay in its prediction of CHF. A loss-of-feedwater transient was modeled using the Macbeth CHF correlation and compared to experimental data with satisfactory results.

  17. A study of natural circulation in the evaporator of a horizontal-tube heat recovery steam generator

    NASA Astrophysics Data System (ADS)

    Roslyakov, P. V.; Pleshanov, K. A.; Sterkhov, K. V.

    2014-07-01

    Results obtained from investigations of stable natural circulation in an intricate circulation circuit with a horizontal layout of the tubes of evaporating surface having a negative useful head are presented. The possibility of making a shift from using multiple forced circulation organized by means of a circulation pump to natural circulation in vertical heat recovery steam generator is estimated. Criteria for characterizing the performance reliability and efficiency of a horizontal evaporator with negative useful head are proposed. The influence of various design solutions on circulation robustness is considered. With due regard of the optimal parameters, the most efficient and least costly methods are proposed for achieving more stable circulation in a vertical heat recovery steam generator when a shift is made from multiple forced to natural circulation. A procedure for calculating the circulation parameters and an algorithm for checking evaporator performance reliability are developed, and recommendations for the design of heat recovery steam generator, nonheated parts of natural circulation circuit, and evaporating surface are suggested.

  18. Investigation of Frequency Mixing Techniques for Eddy Current Testing of Steam Generator Tubes in Nuclear Power Plants

    NASA Astrophysics Data System (ADS)

    Jung, H. J.; Kong, Y. B.; Song, S.-J.; Kim, C.-H.; Choi, Y. H.; Kang, S.-C.; Song, M. H.

    2007-03-01

    In eddy current testing (ECT) of steam generator tubes in nuclear power plants, it is very important to extract flaw signals from the signals compound by flaws and supporting structures. To perform such an important task, the multifrequency ECT methods are widely adopted since they have a well-known capability of extracting the flaw signal from the compound signals. Therefore, various frequency mixing algorithms have been proposed up to now. In the present work, two different frequency mixing algorithms, a time-domain optimization method and a discrete cosine transform (DCT) based optimization method, are investigated using experimental signals captured from a ASME standard tube. In this paper, we discuss the basic principles and the performances of these two frequency mixing techniques.

  19. Passivity degradation of nuclear steam generator tubing alloy induced by Pb contamination at high temperature

    NASA Astrophysics Data System (ADS)

    Lu, B. T.; Luo, J. L.; Lu, Y. C.

    2012-10-01

    Effects of Pb contamination on the passivity of a Ni-based alloy (UNS N06690) in a simulated crevice chemistry of steam generator with near-neutral pH at 300 °C are elucidated using electrochemical measurements and surface analysis techniques. The experimental observations reveal that Pb impurity can enter anodic film, which results in substantial changes in the films structure via hindering the dehydration during the passivation and retarding the formation of spinel oxides. The presence of Pb-contamination can also increase hydrogen content in anodic film. Finally, the mechanism of passivity degradation induced by Pb contamination is described on the basis of the experimental data and established theory.

  20. Experimental Verification of Model-Based ECT Signal Interpretation for Quantitative Flaw Characterization in Steam Generator Tubes

    NASA Astrophysics Data System (ADS)

    Song, Sung-Jin; Kim, Young H.; Kim, Eui-Lae; Chung, Tae-Eon; Yim, Chang-Jae

    2003-03-01

    The model-based inversion tools for eddy current signals have been developed by the novel combination of neural networks and finite element modeling for quantitative flaw characterization in steam generator tubes. In the present work, interpretation of experimental eddy current signals was carried out in order to validate the developed inversion tools. A database was constructed using the synthetic flaw signals generated by the finite element modeling. The hybrid neural networks of a PNN classifier and BPNN size estimators were trained using the synthetic signals. Experimental eddy current signals were obtained from axisymmetric artificial flaws. Interpretations of flaws were carried out by feeding experimental signals into the neural networks. The results of interpretations were excellent, so that the developed inversion tools would be applicable to the interpretation of experimental eddy current signals.

  1. Evaluation of a main steam line break with induced, multiple tube ruptures: A comparison of NUREG 1477 (Draft) and transient methodologies Palo Verde Nuclear Generating Station

    SciTech Connect

    Parrish, K.R.

    1995-09-01

    This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2.

  2. Eddy-current inspection for steam generator tubing program. Annual progress report for period ending December 31, 1979

    SciTech Connect

    Dodd, C.V.; Deeds, W.E.; McClung, R.W.

    1980-07-01

    Eddy-current methods provide the best in-service inspection of steam generator tubing, but present techniques can produce ambiguity because of the many independent variables that affect the signals. The current development program has used mathematical models and developed or modified computer programs to design optimum probes, instrumentation, and techniques for multifrequency, multiproperty examinations. Interactive calculations and experimental measurements have been made with the use of modular eddy-current instrumentation and a minicomputer. These establish the coefficients for the complex equations that define the values of the desired properties (and the attainable accuracy) despite changes in other significant variables. The computer programs for calculating the accuracy with which various properties can be measured indicate that the tubing wall thickness and the defect size can be measured much more accurately than is currently required, even when other properties are varying. Our experimental measurements have confirmed these results, although more testing is needed for all the different combinations of cases and different types of defects. To facilitate the extensive laboratory scanning of the matrix of specimens that are necessary to develop algorithms for detection and analysis for all the possible combinations of positions of flaws, tube supports, and probe coils, we have designed, constructed, and begun operation of a computer-controlled automatic positioner. We have demonstrated the ability to overcome the large signals produced by the edge of the tube supports. An advanced microcomputer has been designed, constructed, and installed in the instrumentation to control the examination and provide real-time calculations of the desired properties for display recording during the scanning of the tube.

  3. Identification and classification of dynamic event tree scenarios via possibilistic clustering: application to a steam generator tube rupture event.

    PubMed

    Mercurio, D; Podofillini, L; Zio, E; Dang, V N

    2009-11-01

    This paper illustrates a method to identify and classify scenarios generated in a dynamic event tree (DET) analysis. Identification and classification are carried out by means of an evolutionary possibilistic fuzzy C-means clustering algorithm which takes into account not only the final system states but also the timing of the events and the process evolution. An application is considered with regards to the scenarios generated following a steam generator tube rupture in a nuclear power plant. The scenarios are generated by the accident dynamic simulator (ADS), coupled to a RELAP code that simulates the thermo-hydraulic behavior of the plant and to an operators' crew model, which simulates their cognitive and procedures-guided responses. A set of 60 scenarios has been generated by the ADS DET tool. The classification approach has grouped the 60 scenarios into 4 classes of dominant scenarios, one of which was not anticipated a priori but was "discovered" by the classifier. The proposed approach may be considered as a first effort towards the application of identification and classification approaches to scenarios post-processing for real-scale dynamic safety assessments. PMID:19819366

  4. Steam generator support system

    DOEpatents

    Moldenhauer, James E.

    1987-01-01

    A support system for connection to an outer surface of a J-shaped steam generator for use with a nuclear reactor or other liquid metal cooled power source. The J-shaped steam generator is mounted with the bent portion at the bottom. An arrangement of elongated rod members provides both horizontal and vertical support for the steam generator. The rod members are interconnected to the steam generator assembly and a support structure in a manner which provides for thermal distortion of the steam generator without the transfer of bending moments to the support structure and in a like manner substantially minimizes forces being transferred between the support structure and the steam generator as a result of seismic disturbances.

  5. Steam generator support system

    DOEpatents

    Moldenhauer, J.E.

    1987-08-25

    A support system for connection to an outer surface of a J-shaped steam generator for use with a nuclear reactor or other liquid metal cooled power source is disclosed. The J-shaped steam generator is mounted with the bent portion at the bottom. An arrangement of elongated rod members provides both horizontal and vertical support for the steam generator. The rod members are interconnected to the steam generator assembly and a support structure in a manner which provides for thermal distortion of the steam generator without the transfer of bending moments to the support structure and in a like manner substantially minimizes forces being transferred between the support structure and the steam generator as a result of seismic disturbances. 4 figs.

  6. Correlation of secondary-side IGA/SCC degradation of recirculating steam generator tubing with the on-line addition of boric acid

    SciTech Connect

    Partridge, M.J.; Zemitis, W.S.; Gorman, J.A. )

    1992-08-01

    A survey of field data indicates that the on-line addition of boric acid can reduce the rate of intergranular attack and stress corrosion cracking (IGA/SCC) within the hot leg tube support crevices for some PWR steam generators. However, the beneficial effect was not seen at all surveyed plants. 68 refs., 12 tabs., 12 refs.

  7. Computational fluid dynamics (CFD) simulations of aerosol in a U-shaped steam generator tube

    NASA Astrophysics Data System (ADS)

    Longmire, Pamela

    scenario evaluated but ranged from 1.61 to 3.2. At the outlet, the computed AMMD (1.9 mum) had GSD between 1.12 and 2.76. Decontamination factors (DF), computed based on deposition from trajectory calculations, were just over 3.5 for the bend and 4.4 at the outlet. Computed DFs were consistent with expert elicitation cited in NUREG-1150 for aerosol retention in steam generators.

  8. CRBRP steam-generator design evolution

    SciTech Connect

    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.

  9. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    SciTech Connect

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  10. Evaluation on double-wall-tube residual stress distribution of sodium-heated steam generator by neutron diffraction and numerical analysis

    SciTech Connect

    Kisohara, N.; Suzuki, H.; Akita, K.; Kasahara, N.

    2012-07-01

    A double-wall-tube is nominated for the steam generator heat transfer tube of future sodium fast reactors (SFRs) in Japan, to decrease the possibility of sodium/water reaction. The double-wall-tube consists of an inner tube and an outer tube, and they are mechanically contacted to keep the heat transfer of the interface between the inner and outer tubes by their residual stress. During long term SG operation, the contact stress at the interface gradually falls down due to stress relaxation. This phenomenon might increase the thermal resistance of the interface and degrade the tube heat transfer performance. The contact stress relaxation can be predicted by numerical analysis, and the analysis requires the data of the initial residual stress distributions in the tubes. However, unclear initial residual stress distributions prevent precious relaxation evaluation. In order to resolve this issue, a neutron diffraction method was employed to reveal the tri-axial (radius, hoop and longitudinal) initial residual stress distributions in the double-wall-tube. Strain gauges also were used to evaluate the contact stress. The measurement results were analyzed using a JAEA's structural computer code to determine the initial residual stress distributions. Based on the stress distributions, the structural computer code has predicted the transition of the relaxation and the decrease of the contact stress. The radial and longitudinal temperature distributions in the tubes were input to the structural analysis model. Since the radial thermal expansion difference between the inner (colder) and outer (hotter) tube reduces the contact stress and the tube inside steam pressure contributes to increasing it, the analytical model also took these effects into consideration. It has been conduced that the inner and outer tubes are contacted with sufficient stresses during the plant life time, and that effective heat transfer degradation dose not occur in the double-wall-tube SG. (authors)

  11. RPV steam generator pressure boundary

    SciTech Connect

    Strosnider, J.

    1996-03-01

    As the types of SG tube degradation affecting PWR SGs has changed, and improvements in tube inspection and repair technology have occurred, current SG regulatory requirements and guidance have become increasingly out of date. This regulatory situation has been dealt with on a plant-specific basis, however to resolve this problem in the long term, the NRC has begun development of a performance-based rule. As currently structured, the proposed steam generator rule would require licensees to implement SG programs that monitor the condition of the steam generator tubes against accepted performance criteria to provide reasonable assurance that the steam generator tubes remain capable of performing their intended safety functions. Currently the staff is developing three performance criteria that will ensure the tubes can continue to perform their safety function and therefore satisfy the SG rule requirements. The staff, in developing the criteria, is striving to ensure that the performance criteria have the two key attributes of being (1) measurable (enabling the tube condition to be {open_quotes}measured{close_quotes} against the criteria) and (2) tolerable (ensuring that failures to meet the criteria do not result in unacceptable consequences). A general description of the criteria are: (1) Structural integrity criteria: Ensures that the structural integrity of the SG tubes is maintained for the operating cycle consistent with the margins intended by the ASME Code. (2) Leakage integrity criteria: Ensures that postulated accident leakages and the associated dose releases are limited relative to 10 CFR Part 50 guidelines and 10 CFR Part 50 Appendix A GDC 19. (3) Operational leakage criteria: Ensures that the operating unit will be shut down as a defense-in depth measure when operational SG tube leakage exceeds established leakage limits.

  12. Corrosion aspects of Ni-Cr-Fe based and Ni-Cu based steam generator tube materials

    NASA Astrophysics Data System (ADS)

    Dutta, R. S.

    2009-09-01

    This paper reviews corrosion related issues of Ni-Cr-Fe based (in a general sense) and Ni-Cu based steam generator tube materials for nuclear power plants those have been dealt with for last more than four decades along with some updated information on corrosion research. The materials include austenitic stainless steels (SSs), Alloy 600, Monel 400, Alloy 800 and Alloy 690. Compatibility related issues of these alloys are briefly discussed along with the alloy chemistry and microstructure. For austenitic SSs, stress corrosion cracking (SCC) behaviour in high temperature aqueous environments is discussed. For Alloy 600, intergranular cracking in high temperature water including hydrogen-induced intergranular cracking is highlighted along with the interactions of material in various environments. In case of Monel 400, intergranular corrosion and pitting corrosion at ambient temperature and SCC behaviour at elevated temperature are briefly described. For Alloy 800, the discussion covers SCC behaviour, surface characterization and microstructural aspects of pitting, whereas hydrogen-related issues are also highlighted for Alloy 690.

  13. High-resolution comparison of primary- and secondary-side intergranular degradation in alloy 600 steam generator tubing

    SciTech Connect

    Bruemmer, Stephen M.; Guertsman, Valery Y.; Thomas, Larry E.

    2000-01-01

    Abstract Intergranular (IG) attack and stress-corrosion cracks in alloy 600 tubing removed from the PWR steam generator #1 at Ringhals 2 have been characterized by analytical transmission electron microscopy (ATEM). Comparisons are made between environmentally induced cracks initiated on the primary-water ID surface versus those initiated on the secondary-water OD surface. General SCC crack morphologies were quite similar with branched IG cracking extending to approximately 50% through wall. Corrosion products in the open cracks were quite different with hydrated nickel phosphate seen filling the secondary-side crack, while the crack wall oxide in the primary-side crack was a Cr and Fe-rich spinel. Both samples revealed narrow (~10-nm wide), deeply penetrated, oxidized zones along most grain boundaries that intersect the open cracks. The local structures and chemistries in these corrosion-affected zones were examined by high-resolution TEM imaging, electron diffraction and fine-probe compositional analysis. These porous IG penetrations were nearly identical in appearance for both the primary- and secondary-side examples and contained Cr-rich oxides (Cr2O3 on the primary side and spinel plus Cr2O3 on the secondary side). Similarities between corrosion-induced structures for primary- and secondary-side cracking may indicate that the same degradation mechanism is operating in both cases. However, controlled experiments are needed where specific mechanisms can be properly distinguished.

  14. Evaluation of sampling plans for in-service inspection of steam generator tubes. Volume 2, Comprehensive analytical and Monte Carlo simulation results for several sampling plans

    SciTech Connect

    Kurtz, R.J.; Heasler, P.G.; Baird, D.B.

    1994-02-01

    This report summarizes the results of three previous studies to evaluate and compare the effectiveness of sampling plans for steam generator tube inspections. An analytical evaluation and Monte Carlo simulation techniques were the methods used to evaluate sampling plan performance. To test the performance of candidate sampling plans under a variety of conditions, ranges of inspection system reliability were considered along with different distributions of tube degradation. Results from the eddy current reliability studies performed with the retired-from-service Surry 2A steam generator were utilized to guide the selection of appropriate probability of detection and flaw sizing models for use in the analysis. Different distributions of tube degradation were selected to span the range of conditions that might exist in operating steam generators. The principal means of evaluating sampling performance was to determine the effectiveness of the sampling plan for detecting and plugging defective tubes. A summary of key results from the eddy current reliability studies is presented. The analytical and Monte Carlo simulation analyses are discussed along with a synopsis of key results and conclusions.

  15. CORCO downhole steam generator

    SciTech Connect

    Rintoul, B.

    1982-03-01

    The opening of a new frontier in steaming moved forward in Jan. 1982 when a CORCO (Chemical Oil Recovery Co.) generator described as the first commercial down-hole steam generator went into operation in Kern County's Devils Den field, 60 miles northwest of Bakersfield, CA. A major reason for selecting the down-hole generator for the Devils Den field is that along with steam the unit puts away flue gas resulting from combustion. There is no pressure to speak of in the escudo, and it is hoped that the inert gas will build up bottom-hole pressure to assist in oil recovery. Another reason is that the down-hole generator, rated for 7 million btu/hr, makes it possible to tailor steam injection to the well's requirements. The advantages and disadvantages of the CORCO generator are described, along with its application in the Kern River field.

  16. Steam generator hand hole shielding.

    PubMed

    Cox, W E

    2000-05-01

    Seabrook Station is an 1198 MWE Pressurized Water Reactor (PWR) that began commercial operation in 1990. Expensive and dose intensive Steam Generator Replacement Projects among PWR operators have led to an increase in steam generator preventative maintenance. Most of this preventative maintenance is performed through access ports in the shell of the steam generator just above the tube sheet known as secondary side hand holes. Secondary side work activities performed through the hand holes are typically performed without the shielding benefit of water in the secondary side of the steam generator. An increase in cleaning and inspection work scope has led to an increase in dose attributed to steam generator secondary side maintenance. This increased work scope and the station goal of maintaining personnel radiation dose ALARA led to the development of the shielding concept described in this article. This shield design saved an estimated 2.5 person-rem (25 person-Smv) the first time it was deployed and is expected to save an additional 50 person-rem (500 person-mSv) over the remaining life of the plant. PMID:10770158

  17. Proceedings: 1984 Workshop on Secondary-Side Stress Corrosion Cracking and Intergranular Corrosion of PWR Steam Generator Tubing

    SciTech Connect

    1986-03-01

    During 1984, research investigating intergranular corrosion and stress corrosion cracking in PWR steam generators provided data to formulate a corrosion-product transport theory. In addition, the research showed that changing the pH of liquids in generator crevices will retard and sometimes arrest the corrosion process.

  18. Failure Pressure and Leak Rate of Steam Generator Tubes With Stress Corrosion Cracks

    SciTech Connect

    Majumdar, S.; Kasza, K.; Park, J.Y.; Bakhitiari, S.

    2002-07-01

    This paper illustrates the use of an 'equivalent rectangular crack' approach to predict leak rates through laboratory generated stress corrosion cracks. A comparison between predicted and observed test data on rupture and leak rate from laboratory generated stress corrosion cracks are provided. Specimen flaws were sized by post-test fractography in addition to pre-test advanced eddy current technique. The test failure pressures and leak rates are shown to be closer to those predicted on the basis of fractography than on NDE. However, the predictions based on NDE results are encouraging, particularly because they have the potential to determine a more detailed geometry of ligamentous cracks from which more accurate predictions of failure pressure and leak rate can be made in the future. (authors)

  19. Application of nano-sized TiO2 as an inhibitor of stress corrosion cracking in the steam generator tube materials.

    PubMed

    Kim, Kyung Mo; Lee, Eun Hee; Kim, Uh Chul; Choi, Byung Seon

    2010-01-01

    Several chemicals were studied to suppress the damage due to a stress corrosion cracking (SCC) of the steam generator (SG) tubes in nuclear power plants. SCC tests were carried out to investigate the performance of TiO2 on several types of SG tube materials. The SCC tests were conducted by using an m-RUB specimen in a 10% NaOH solution at a temperature of 315 degrees C. The test with the addition of TiO2 showed a decrease in the SCC rate for the SG tubing materials. In order to improve the inhibition property in a crevice of TiO2, a sonochemical technique was applied to reduce the size of the TiO2 particle. From the SCC tests with the RUB specimen, the SG tube materials showed an enhanced cracking resistance with the addition of nano-sized TiO2 and the surface property was also changed. PMID:20352829

  20. Candu 6 severe core damage accident consequence analysis for steam generator tube rupture scenario using MAAP4-CANDU V4.0.5A: preliminary results

    SciTech Connect

    Petoukhov, S.M.; Awadh, B.; Mathew, P.M.

    2006-07-01

    This paper describes the preliminary results of the consequence analysis for a generic AECL CANDU 6 station, when it undergoes a postulated, low probability Steam Generator multiple Tube Rupture (SGTR) severe accident with assumed unavailability of several critical plant safety systems. The Modular Accident Analysis Program for CANDU (MAAP4-CANDU) code was used for this analysis. The SGTR accident is assumed to begin with the guillotine rupture of 10 steam generator tubes in one steam generator in Primary Heat Transport System (PHTS) loop 1. For the reference case, the following systems were assumed unavailable: moderator and shield cooling, emergency core cooling, crash cool-down, and main and auxiliary feed water. Two additional cases were analyzed, one with the crash cool-down system available, and another with the crash cool-down and the auxiliary feed water systems available. The three scenarios considered in this study show that most of the initial fission product inventory would be retained within the containment by various fission product retention mechanisms. For the case where the crash cool-down system was credited but the auxiliary feed water systems were not credited, the total mass of volatile fission products released to the environment including stable and radioactive isotopes was about four times more than in the reference case, because fission products could be released directly from the PHTS to the environment through the Main Steam Safety Valves (MSSVs), bypassing the containment. For the case where the crash cool-down and auxiliary feed water systems were credited, the volatile fission product release to the environment was insignificant, because the fission product release was substantially mitigated by scrubbing in the water pool in the secondary side of the steam generator (SG). (authors)

  1. TRAC PF1/MOD1 calculations and data comparisons for mist feed and bleed and steam generator tube rupture experiments

    SciTech Connect

    Siebe, D.A.; Boyack, B.E.; Steiner, J.L.

    1988-01-01

    Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents, loss of feedwater and other transients in Babcock and Wilcox (BandW) plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 /times/ 4 (two hot legs and steam generators (SGs), four cold legs and reactor coolant pumps) representation of lowered-loop reactor system of the BandW design. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Two other integral experimental facilities are included in the IST program: test loops at the University of Maryland, College Park, and at SRI International (SRI-2). The objective of the IST tests is to generate high-quality experimental data to be used for assessing thermal-hydraulic safety computer codes. Efforts are under way at Los Alamos to assess TRAC-PF1/MOD1 against data from each of the IST facilities. Calculations and data comparisons for TRAC-PF1/MOD1 assessment are presented for two transients run in the MIST facility. These are MIST Test 330302, a feed and bleed test with delayed high-pressure injection; and Test 3404AA, an SG tube-rupture test with the affected SG isolated. Only MIST assessment results are presented in this paper. The TRAC-PF1/MOD1 calculations completed to date for MIST tests are in reasonable agreement with the data from these tests. Reasonable agreement is defined as meaning that major trends are predicted correctly, although TRAC values are frequently outside the range of data uncertainty. We believe that correct conclusions will be reached if the code is used in similar applications despite minor code/model deficiencies. 7 refs., 5 figs., 2 tabs.

  2. Steam generators regulatory practices and issues in Spain

    SciTech Connect

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

    1997-02-01

    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.

  3. Steam generator for liquid metal fast breeder reactor

    DOEpatents

    Gillett, James E.; Garner, Daniel C.; Wineman, Arthur L.; Robey, Robert M.

    1985-01-01

    Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

  4. Corrosion response of downhole steam generator assembly and instrumentation and supply line tubings at Long Beach field test

    SciTech Connect

    Weirick, L.J.

    1983-07-01

    Three families of metals were used to make the non-combustor components for Sandia's downhole steam generator assembly and supply and instrumentation lines. These three families were: first, plain carbon steel (API Grade J 55); second, austenitic stainless steels (316 and 310); and third, a nickel-base superalloy (Inconel 600). The metals in all three of these families were found to be deficient in their corrosion response. J 55 exhibited minimal to severe general corrosion (rusting). The austenitic stainless steels pitted and cracked. Inconel 600 showed both severe pitting and some intergranular attack. For the most part, these materials were found to be unsuitable for extended life in a downhole steam generator. It is recommended that Inconel 625 be used in future systems where a moderate strength material is specified and Inconel 718 be considered where a high strength material is necessary. 11 references, 45 figures, 5 tables.

  5. Aerodynamic heated steam generating apparatus

    SciTech Connect

    Kim, K.

    1986-08-12

    An aerodynamic heated steam generating apparatus is described which consists of: an aerodynamic heat immersion coil steam generator adapted to be located on the leading edge of an airframe of a hypersonic aircraft and being responsive to aerodynamic heating of water by a compression shock airstream to produce steam pressure; an expansion shock air-cooled condensor adapted to be located in the airframe rearward of and operatively coupled to the aerodynamic heat immersion coil steam generator to receive and condense the steam pressure; and an aerodynamic heated steam injector manifold adapted to distribute heated steam into the airstream flowing through an exterior generating channel of an air-breathing, ducted power plant.

  6. SCDAP/RELAP5 Evaluation of the Potential for Steam Generator Tube Ruptures as a Result of Severe Accidents in Operating PWRs

    SciTech Connect

    Knudson, Darrell Lee; Ghan, Larry Scott; Dobbe, Charles Albin

    1998-09-01

    Natural circulation flows can develop within a reactor coolant system (RCS) during certain severe reactor accidents, transferring decay energy from the core to other parts of the RCS. The associated heatup of RCS structures can lead to pressure boundary failures; with notable vulnerabilities in the pressurizer surge line, the hot leg nozzles, and the steam generator (SG) tubes. The potential for a steam generator tube rupture (SGTR) is of particular concern because fission products could be released to the environment through such a failure. The Nuclear Regulatory Commission (NRC) developed a program to address SG tube integrity issues in operating pressurized water reactors (PWRs) based on the possibility for environmental release. An extensive effort to evaluate the potential for accident-induced SGTRs using SCDAP/RELAP5 at the Idaho National Engineering and Environmental Laboratory (INEEL) was directed as one part of the NRC program. All SCDAP/RELAP5 calculations performed during the INEEL evaluation were based on station blackout accidents (and variations thereof) because those accidents are considered to be one of the more likely scenarios leading to natural circulation flows at temperatures and pressures that could threaten SG tube integrity (as well as the integrity of other vulnerable RCS pressure boundaries). Variations that were addressed included consideration of the effects of RCP seal leaks, intentional RCS depressurization through pressurizer PORVs, SG secondary depressurization, DC-HL bypass flows, U-tube SG sludge accumulation, and quenching of upper plenum stainless steel upon relocation to the lower head. Where available, experimental data was used to guide simulation of natural circulation flows. Independent reviews of the applicability of the natural circulation experimental data, the suitability of the code, and the adequacy of the modeling were completed and review recommendations were incorporated into the evaluation within budget and

  7. Large-eddy simulation of turbulence in steam generators

    SciTech Connect

    Bagwell, T.G.; Hassan, Y.A. ); Steininger D.A. )

    1989-11-01

    A major problem associated with steam generators is excessive tube vibration caused by turbulent-flow buffeting and fluid-elastic excitation. Vibration can lead to tube rupture or wear, necessitating tube plugging and reducing the availability of the steam generator. The fluid/structure interaction phenomenon that causes fluid-elastic tube excitation is unknown at present. The current investigation defines the spectral characteristics of turbulent flow entering the Westinghouse D4 steam generator tube bundles using the large-eddy simulation (LES) technique. Due to the recent availability of supercomputers, LES is being considered as a possible engineering design analysis tool. The information from this study will provide input for defining the temporally fluctuating forces on steam generator tube banks. The GUST code was used to analyze the water box of a Westinghouse model D4 steam generator.

  8. ADVANCED STEAM GENERATORS

    SciTech Connect

    Richards, Geo. A.; Casleton, Kent H.; Lewis, Robie E.; Rogers, William A.; Woike, Mark R.; Willis; Brian P.

    2001-11-06

    Concerns about climate change have encouraged significant interest in concepts for ultra-low or ''zero''-emissions power generation systems. In some proposed concepts, nitrogen is removed from the combustion air and replaced with another diluent such as carbon dioxide or steam. In this way, formation of nitrogen oxides is prevented, and the exhaust stream can be separated into concentrated CO{sub 2} and steam or water streams. The concentrated CO{sub 2} stream could then serve as input to a CO{sub 2} sequestration process or utilized in some other way. Some of these concepts are illustrated in Figure 1. This project is an investigation of one approach to ''zero'' emission power generation. Oxy-fuel combustion is used with steam as diluent in a power cycle proposed by Clean Energy Systems, Inc. (CES) [1,2]. In oxy-fuel combustion, air separation is used to produce nearly pure oxygen for combustion. In this particular concept, the combustion temperatures are moderated by steam as a diluent. An advantage of this technique is that water in the product stream can be condensed with relative ease, leaving a pure CO{sub 2} stream suitable for sequestration. Because most of the atmospheric nitrogen has been separated from the oxidant, the potential to form any NOx pollutant is very small. Trace quantities of any minor pollutants species that do form are captured with the CO{sub 2} or can be readily removed from the condensate. The result is a nearly zero-emission power plant. A sketch of the turbine system proposed by CES is shown in Figure 2. NETL is working with CES to develop a reheat combustor for this application. The reheat combustion application is unusual even among oxy-fuel combustion applications. Most often, oxy-fuel combustion is carried out with the intent of producing very high temperatures for heat transfer to a product. In the reheat case, incoming steam is mixed with the oxygen and natural gas fuel to control the temperature of the output stream to about

  9. Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

    SciTech Connect

    Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J.

    1996-09-01

    The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR.

  10. US PWR steam generator management: An overview

    SciTech Connect

    Welty, C.S. Jr.

    1997-02-01

    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 management{close_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 Corrosion{close_quotes}, and is provided as a supplement to that material.

  11. Revised evaluation of steam generator testing alternatives

    SciTech Connect

    1981-01-01

    A scoping evaluation was made of various facility alternatives for test of LMFBR prototype steam generators and models. Recommendations are given for modifications to EBR-II and SCTI (Sodium Components Test Installation) for prototype SG testing, and for few-tube model testing. (DLC)

  12. Downhole steam generator shows merit

    SciTech Connect

    Not Available

    1980-11-01

    Production from a 5-spot pattern in Kern River Field reached 25,000 bbl during a 5-month test of a down-hole steam generator-equivalent to the amount of oil expected if steam injection from the conventional source had been continued. The test evaluated the down-hole generator as a steam source relatively free of atmospheric pollutants. The biggest objection to steam recovery of heavy crude is the volume of combustion products vented to the atmosphere, and these frequently contain small amounts of sulfur compounds. One big advantage of generating steam down hole is elimination of heat losses in the injection well. The practical limit for conventional steam injection is in a reservoir approximately 2,500 ft deep; the down-hole generator should operate economically to 6,000 ft. The test proved the feasibility of the method, and cleared the way for a series of down-hole generator installation and retrieval tests.

  13. Pressing problems of managing the service life of tube bundles used in steam generators at nuclear power stations equipped with VVER reactors

    NASA Astrophysics Data System (ADS)

    Trunov, N. B.; Popadchuk, V. S.; Davidenko, S. E.; Zhukov, R. Yu.

    2010-05-01

    Optimal approaches for monitoring the state and blanking of the tube bundles of horizontal tube generators are considered, and pressing problems associated with managing their service life are discussed.

  14. Downhole steam generation: material studies

    SciTech Connect

    Beauchamp, E.K.; Weirick, L.J.; Muir, J.F.

    1982-01-01

    One enhanced oil recovery technique for extracting heavy crude from deep reservoirs by steam at the bottom of an injection well. Development of a downhole steam generator that will produce steam and inject it into formations at depths greater than 2500 feet is one objective of a Department of Energy/Sandia National Laboratories development effort - Project DEEP STEAM. Extensive material studies have been performed in support of Project DEEP STEAM; current efforts are devoted primarily to the selection and evaluation of materials for use in downhole steam generators. This paper presents observations of the performance of candidate metals and refractory ceramics (combustor liners) during tests of two prototypic, high pressure, diesel/air combustion, direct contact, downhole steam generators. The first downhole test of such a generator provides data on the performance of various metals (304L, 310 and 316S stainless steels and plain carbon steel) exposed for several weeks to a warm, aerated saltwater environment. A number of corrosion mechanisms acted to cause severely degraded perforance of some of the metals. Several refractory liner designs were evaluated during ground level tests of a generator having a ceramic-lined combustion chamber. Of the two refractories employed, alumina and silicon carbide, the alumina liners exhibited more serious surface degradation and corrosion.

  15. Laser removal of sludge from steam generators

    DOEpatents

    Nachbar, Henry D.

    1990-01-01

    A method of removing unwanted chemical deposits known as sludge from the metal surfaces of steam generators with laser energy is provided. Laser energy of a certain power density, of a critical wavelength and frequency, is intermittently focused on the sludge deposits to vaporize them so that the surfaces are cleaned without affecting the metal surface (sludge substrate). Fiberoptic tubes are utilized for laser beam transmission and beam direction. Fiberoptics are also utilized to monitor laser operation and sludge removal.

  16. Completion system for downhole steam generator

    SciTech Connect

    Vandevier, J.E.

    1989-05-30

    This patent describes an apparatus for providing electrical power to a downhole steam generator in a cased well. The method consists of: a packer supported on a string of tubing, the packer having means for sealing against casing in the well and at least one conduit extending longitudinally through the packer; a connector box mounted below the lower end of the packer, the connector box having a connector plate containing a plurality of passages; a plurality of feed through electrical connectors mounted in insulators in the passages in the connector plate; support means for mounting the steam generator below the connector box; an aperture located in the sidewall of the tubing immediately above the packer; an electrical cable extending from the surface alongside the tubing into the aperture and through the conduit into the connector box, the electrical cable having a plurality of electrical conductors, each of which ends in a terminal that is electrically connected to one of the electrical connectors; and electrical conductors extending between the steam generator and engaging a lower end of each electrical connector in the connection plate.

  17. International agreement report: Assessment study of RELAP-5 MOD-2 Cycle 36. 01 based on the DOEL-2 Steam Generator Tube Rupture incident of June 1979

    SciTech Connect

    Stubbe, E J

    1986-10-01

    This report presents a code assessment study based on a real plant transient that occurred at the DOEL 2 power plant in Belgium on June 25th 1979. DOEL 2 is a two-loop WESTINGHOUSE PWR plant of 392 MWe. A steam generator tube rupture occurred at the end of a heat-up phase which initiated a plant transient which required substantial operator involvement and presented many plant phenomena which are of interest for code assessment. While real plant transients are of special importance for code validation because of the elimination of code scaling uncertainties, they introduce however some uncertainties related to the specifications of the exact initial and boundary conditions which must be reconstructed from available on-line plant recordings and on-line computer diagnostics. Best estimate data have been reconstructed for an assessment study by means of the code RELAP5/MOD2/CYCLE 36.01. Because of inherent uncertainties in the plant data, the assessment work is focussed on phenomena whereby the comparison between plant data and computer data is based more on trends than on absolute values. Such approach is able to uncover basic code weaknesses and strengths which can contribute to a better understanding of the code potential.

  18. Direct steam generation in line-focus solar collectors

    NASA Astrophysics Data System (ADS)

    May, E. K.; Murphy, L. M.

    1983-01-01

    The performance benefits of the direct (in situ) generation of steam in the receiver tube of a line focus solar collector were assessed. Compared to existing technology using steam flash or unfired boiler systems, the in situ technique could produce 25% more steam at a reduced delivery cost. It is indicated that two phase flow instabilities, if present, can be readily controlled, and that the possibility of freezing is not an impediment to using water in cold climates.

  19. Characterization of PWR steam generator deposits

    SciTech Connect

    Varrin, R. Jr.

    1996-02-01

    Restoring the thermal performance of the steam generators often requires the utility to remove deposits by expensive chemical means. This work demonstrates that careful characterization of secondary side deposit samples can reveal their chemical and physical properties which in turn contribute to an overall assessment of the need for and extent of steam generator inspection and maintenance. More specifically, knowledge of deposit characteristics can contribute to: (1) determination of the source of corrosion products, (2) assessment of feedwater chemistry control strategies, (3) prediction of rates of tube degradation, and (4) evaluation of degraded heat transfer performance or flow instabilities. Despite the relationships between deposits and steam generator operation and performance, few utilities elect to perform the types of characterizations which are suitable for the determination of the specific chemical and physical nature of their particular deposits. One of the principal goals of this document is to encourage utilities to consider deposit characterization an integral part of an overall effort to assess and maintain the material condition of the steam generators at their plant. This document includes a review of the nature of deposits and relates deposit characteristics to a variety of secondary side phenomena including corrosion and fouling. Candidate techniques for revealing relevant deposit properties are provided so that inferences regarding the role of deposits in promoting or causing these phenomena at their plant can be developed.

  20. Design with constructal theory: Steam generators, turbines and heat exchangers

    NASA Astrophysics Data System (ADS)

    Kim, Yong Sung

    This dissertation shows that the architecture of steam generators, steam turbines and heat exchangers for power plants can be predicted on the basis of the constructal law. According to constructal theory, the flow architecture emerges such that it provides progressively greater access to its currents. Each chapter shows how constructal theory guides the generation of designs in pursuit of higher performance. Chapter two shows the tube diameters, the number of riser tubes, the water circulation rate and the rate of steam production are determined by maximizing the heat transfer rate from hot gases to riser tubes and minimizing the global flow resistance under the fixed volume constraint. Chapter three shows how the optimal spacing between adjacent tubes, the number of tubes for the downcomer and the riser and the location of the flow reversal for the continuous steam generator are determined by the intersection of asymptotes method, and by minimizing the flow resistance under the fixed volume constraints. Chapter four shows that the mass inventory for steam turbines can be distributed between high pressure and low pressure turbines such that the global performance of the power plant is maximal under the total mass constraint. Chapter five presents the more general configuration of a two-stream heat exchanger with forced convection of the hot side and natural circulation on the cold side. Chapter six demonstrates that segmenting a tube with condensation on the outer surface leads to a smaller thermal resistance, and generates design criteria for the performance of multi-tube designs.

  1. Development, construction, and use of pneumometric tubes for measurement of steam flow in the steam lines of PVG-1000 at NPP

    SciTech Connect

    Gorbunov, Yu. S.; Ageev, A. G.; Vasil'eva, R. V.; Korol'kov, B. M.

    2007-05-15

    A system for the direct measurement of steam flow in steam lines after a steam generator, which utilizes a special design of pneumometric tubes and a computing unit that accounts for variation in steam pressure, has been developed to improve the quality of water-level regulation in the steam generators of VVER-1000 power-generating units in the stationary and transitional modes. The advantage of the pneumometric tubes consists in their structural simplicity, high erosion resistance, and absence of irrevocable losses during measurement of steam flow. A similar measurement system is used at foreign NPP. The measurement system in question has been placed in experimental service at the No. 3 unit of the Balakovo NPP, and has demonstrated its worthiness. This measurement system can also be used to determine steam flow in the steam lines of NPP units with VVER-1000 and VVER-440 reactors, and PBMK-1000 power-generating units.

  2. NUCLEAR FLASH TYPE STEAM GENERATOR

    DOEpatents

    Johns, F.L.; Gronemeyer, E.C.; Dusbabek, M.R.

    1962-09-01

    A nuclear steam generating apparatus is designed so that steam may be generated from water heated directly by the nuclear heat source. The apparatus comprises a pair of pressure vessels mounted one within the other, the inner vessel containing a nuclear reactor heat source in the lower portion thereof to which water is pumped. A series of small ports are disposed in the upper portion of the inner vessel for jetting heated water under pressure outwardly into the atmosphere within the interior of the outer vessel, at which time part of the jetted water flashes into steam. The invention eliminates the necessity of any intermediate heat transfer medium and components ordinarily required for handling that medium. (AEC)

  3. The criterion for blanking-off heat-transfer tubes in the steam generators at VVER-based nuclear power plants based on the results of eddy-current examination

    NASA Astrophysics Data System (ADS)

    Lunin, V. P.; Zhdanov, A. G.; Chegodaev, V. V.; Stolyarov, A. A.

    2015-05-01

    The problem of defining the criterion for blanking off heat-transfer tubes in the steam generators at nuclear power plants on the basis of signals obtained from the standard multifrequency eddy-current examination is considered. The decision about blanking off one or another tube is presently made with reference to one parameter of the relevant signal at the working frequency, namely, with reference to its phase, which directly depends on the depth of the flaw being detected, i.e., a crack in the tube. The crack depth equal to 60% of the tube wall thickness is regarded to be the critical one, at which a decision about withdrawing such a tube out from operation (blanking off) must be taken. However, since mechanical tensile rupture tests of heat-transfer tubes show the possibility of their further use with such flaws, the secondary parameter of the signal, namely, its amplitude, must be used for determining the blanking-off criterion. The signals produced by the standard flow-type transducers in response to flaws in the form of a longitudinal crack having the depth and length within the limits permitted by the relevant regulations were calculated using 3D finite-element modeling. Based on the obtained results, the values of the eddy-current signal amplitude were determined, which, together with the signal phase value, form a new amplitude-phase criterion for blanking off heat-transfer tubes. For confirming the effectiveness of this technique, the algorithm for revealing the signal indications satisfying the proposed amplitude-phase criterion was tested on real signals obtained from operational eddy-current examination of the state of steam generator heat-transfer tubes carried out within the framework of planned preventive repair.

  4. Direct firing downhole steam generator

    SciTech Connect

    Binsley, R.L.; Wagner, W.R.; Wright, D.E.

    1982-06-29

    Direct firing downbole steam generator basically comprises an injector assembly axially connected with a combustion chamber. Downstream of the combustion chamber and oriented so as to receive its output is a heat exchanger wherein preheated water is injected into the heat exchanger through a plurality of one-way valves, vaporized and injected through a nozzle, packer and check valve into the well formation.

  5. Fast fluidized bed steam generator

    DOEpatents

    Bryers, Richard W.; Taylor, Thomas E.

    1980-01-01

    A steam generator in which a high-velocity, combustion-supporting gas is passed through a bed of particulate material to provide a fluidized bed having a dense-phase portion and an entrained-phase portion for the combustion of fuel material. A first set of heat transfer elements connected to a steam drum is vertically disposed above the dense-phase fluidized bed to form a first flow circuit for heat transfer fluid which is heated primarily by the entrained-phase fluidized bed. A second set of heat transfer elements connected to the steam drum and forming the wall structure of the furnace provides a second flow circuit for the heat transfer fluid, the lower portion of which is heated by the dense-phase fluidized bed and the upper portion by the entrained-phase fluidized bed.

  6. PREDICTION OF OXIDE SCALE EXFOLIATION IN STEAM TUBES

    SciTech Connect

    Sabau, Adrian S; Wright, Ian G

    2010-01-01

    Numerical simulation results are presented for the prediction of the likelihood of oxide scale exfoliation from superheater tubes. The scenarios considered involved alloys T22, TP347H, and TP347HFG subjected to a simplified operating cycle in a power plant generating supercritical steam. The states of stress and strain of the oxides grown in steam were based solely on modeling the various phenomena experienced by superheater tubes during boiler operation, current understanding of the oxidation behavior of each alloy in steam, and consideration of operating parameters such as heat flux, tube dimensions, and boiler duty cycle. Interpretation of the evolution of strain in these scales, and the approach to conditions where scale failure (hence exfoliation) is expected, makes use of the type of Exfoliation Diagrams that incorporate various cracking and exfoliation criteria appropriate for the system considered. In these diagrams, the strain accumulation with time in an oxide is represented by a strain trajectory derived from the net strain resulting from oxide growth, differences in coefficients of thermal expansion among the components, and relaxation due to creep. It was found that an oxide growing on a tube subjected to routine boiler load cycling conditions attained relatively low values of net strain, indicating that oxide failure would not be expected to occur during normal boiler operation. However, during a boiler shut-down event, strains sufficient to exceed the scale failure criteria were developed after times reasonably in accord with plant experience, with the scales on the ferritic steel failing in tension, and those on the austenitic steels in compression. The results presented illustrate that using this approach to track the state of strain in the oxide scale through all phases of boiler operation, including transitions from full-to-low load and shut-down events, offers the possibility of identifying the phase(s) of boiler operation during which oxide

  7. Downhole steam generator: field tests

    SciTech Connect

    Eson, R.L.

    1982-01-01

    Excessive air pollution and heat losses up to 32% in the surface lines and out the stacks of conventional generators are reasons why conventional steam generation is efficient. These problems are addressed and overcome through the use of a direct-fired down-hole steam generator (DSG). By performing the combustion process at high pressure, and then adding water, a mixture of carbon dioxide, nitrogen, and steam is discharged directly into the heavy oil reservoir. This study documents a series of field tests of a direct-fired DSG showing its ability to produce and inject high quality steam into heavy oil reservoirs without the need for expensive stack scrubbers to remove sulfur dioxide (SO/sub 2/), as well as sophisticated nitrogen oxides (NO/sub x/) control techniques. Results from the 6-in. diameter, 6-ft long, 7.1-mmBtu/hr DSG showed that corrosion can be controlled and production can be improved dramatically in actual field tests in California heavy oil reservoirs.

  8. New downhole steam generator tested

    SciTech Connect

    Bleakley, W.B.

    1981-07-01

    Completion of 2 field tests of a new-model down-hole steam generator paves the way for further evaluation and development of a system destined to increase California's heavy oil production. Current air pollution restrictions there prevent installation of conventional steam generators in several areas of interest to oil operators. The current series of tests, conducted by Chemical Oil Recovery Co. (CORCO) of Bakersfield, California, follows an earlier prototype operation conducted by Sandia National Laboratories in conjunction with the US Department of Energy. The CORCO tests were conducted on the surface with the generator's output going into Tenneco Oil Exploration and Production Co.'s overland-Riokern Well No. 80, located in the Kern River field 4 miles north of Bakersfield. The first test was concluded with just under 1000 bbl of steam injected, less than planned due to a higher-than-expected injection pressure. The unit operated at less than 25% capacity because of the air compressor limitation. Compressor output was only 285 psi, not enough to inject the desired volumes into the reservoir. Test data shows that injection amounted to 150 bpd of 90 to 95% quality steam at 225-psi wellhead pressure. After injection, the well was shut in for 3 days to allow soaking, then put on production. Initial production was 40 bopd at 175 F.

  9. Eddy-current steam generator data analysis performance. Final report

    SciTech Connect

    Harris, D.H.

    1993-06-01

    This study assessed the accuracy of eddy current, bobbin coil data analysis of steam generator tubes conducted under the structure of the PWR Steam Generator Examination Guidelines, Individual and team performance measures were obtained from independent analyses of data from 1619 locations in a sample of 199 steam generator tubes. The 92 reportable indications contained in the tube sample, including 64 repairable indications, were attributable to: wear at anti-vibration bars, intergranular attack/stress-corrosion cracking (IGA/SCC) within tube sheet crevice regions, primary-water stress-corrosion cracking (PWSCC) at tube roll transitions, or thinning at cold-leg tube supports. Analyses were conducted by 20 analysts, four each from five vendors of eddy current steam generator examination services. In accordance with the guidelines, site orientation was provided with plant-specific guidelines; preanalysis practice was completed on plant-specific data; analysts were qualified by performance testing; and independent primary-secondary analyses were conducted with resolution of discrepancies (team analyses). Measures of analysis performance included percentages of indications correctly reported, percentages of false reports, and relative operating characteristic (ROC) curves. ROC curves presented comprehensive pictures of analysis accuracy generalizable beyond the specific conditions of this study. They also provided single-value measures of analysis accuracy. Conclusions and recommendations were provided relative to analysis accuracy, effect of primary-secondary analyses, analyses of tube sheet crevice regions, establishment of reporting criteria, improvement of examination guidelines, and needed research.

  10. Prediction of localized flow velocities and turbulence in a PWR steam generator: Final report

    SciTech Connect

    Stuhmiller, J.H.

    1988-05-01

    The Steam Generator Project Office (SGPO) of the Steam Generator Owners Group and Electric Power Research Institute has developed a methodology for prediction of steam generator tube buffeting and associated material wear. Turbulent buffeting of steam generator tubes causes low amplitude vibratory response which results in fretting wear at support locations. Concerns raised at the Zion Nuclear Power Plant regarding the useful life of their steam generators prompted this study, in which the SGPO methodology is applied to analysis of the Westinghouse Model 51 steam generator. The specific intent of this project was to calculate turbulent buffeting forces within the tube bank of an operating Model 51 steam generator as a first step in the overall SGPO tube vibration and wear prediction strategy. Attention is focused on flow in the vicinity of anti-vibration bars (U-bend region) and on the flow that leaves the downcomer to impact against peripheral tubes. Other projects utilized the buffeting forces calculated here to determine tube vibratory response, tube-support plate impact statistics, and material wear rates. Besides successfully calculating hydraulic buffeting loads within the tube bank, the present project has enhanced the SGPO methodology and has identified hitherto unnoticed flow phenomena that occur in the steam generator. Experiments have also been carried out to validate numerical computations of the steam generator flow field.

  11. MINET validation study using steam generator test data

    SciTech Connect

    Van Tuyle, G.J.; Guppy, J.G.

    1984-01-01

    Three steam generator transient test cases that were simulated using the MINET computer code are described, with computed results compared against experimental data. The MINET calculations closely agreed with the experiment for both the once-through and the U-tube steam generator test cases. The effort is part of an ongoing effort to validate the MINET computer code for thermal-hydraulic plant systems transient analysis, and strongly supports the validity of the MINET models.

  12. Steam drive recovery method utilizing a downhole steam generator

    SciTech Connect

    Snavely, E. S.; Hopkins, D. N.

    1984-09-18

    Viscous oil is recovered from a subterranean, viscous oil-containing formation by a steam flooding technique wherein steam is generated in a downhole steam generator located in an injection well by spontaneous combustion of a pressurized mixture of a water-soluble fuel such as sugars and alcohols dissolved in water and substantially pure oxygen. The generated mixture of steam and combustion gases pass through the formation, displacing oil and reducing the oil's viscosity and the mobilized oil is produced from the formation via a spaced-apart production well.

  13. Steam drive oil recovery method utilizing a downhole steam generator

    SciTech Connect

    Nopkins, D. N.; Snavely, E. S.

    1984-10-23

    Viscous oil is recovered from a subterranean, viscous oil-containing formation by a steam flooding technique wherein steam is generated in a downhole steam generator located in an injection well by spontaneous combustion of a pressurized mixture of a water-soluble fuel such as sugars and alcohols dissolved in water or a stable hydrocarbon fuel-in-water emulsion and substantially pure oxygen. The generated mixture of steam and combustion gases pass through the formation, displacing oil and reducing the oil's viscosity and the mobilized oil is produced from the formation via a spaced-apart production well.

  14. Bore tube assembly for steam cooling a turbine rotor

    SciTech Connect

    DeStefano, Thomas Daniel; Wilson, Ian David

    2002-01-01

    An axial bore tube assembly for a turbine is provided to supply cooling steam to hot gas components of the turbine wheels and return the spent cooling steam. A pair of inner and outer tubes define a steam supply passage concentric about an inner return passage. The forward ends of the tubes communicate with an end cap assembly having sets of peripheral holes communicating with first and second sets of radial tubes whereby cooling steam from the concentric passage is supplied through the end cap holes to radial tubes for cooling the buckets and return steam from the buckets is provided through the second set of radial tubes through a second set of openings of the end cap into the coaxial return passage. A radial-to-axial flow transitioning device, including anti-swirling vanes is provided in the end cap. A strut ring adjacent the aft end of the bore tube assembly permits axial and radial thermal expansion of the inner tube relative to the outer tube.

  15. Direct firing downhole steam generator

    SciTech Connect

    Wagner, W.R.; Wright, D.E.; Binsley, R.L.

    1982-06-29

    A direct firing down-hole steam generator is composed of an injector assembly axially connected with a combustion chamber. Downstream of the combustion chamber and oriented so as to receive its output is a heat exchanger where preheated water is injected into the heat exchanger through a number of one-way valves. The heated water is vaporized and injected through a nozzle, packer, and check valve into the well formation. 9 claims.

  16. Corrosion on steam-side heat exchanger tubes

    SciTech Connect

    Arevalo, A.; Esparza, P.; Bas, C.G.; Morales, J.; Gonzalez, S.; Sanchez, S.R. de

    1996-01-01

    Two corrosion cases occurred on the steam side of tubes in a heat exchanger in a central power station which used seawater as a coolant. The first case is stress corrosion cracking in the presence of oxygen and ammonia with through-wall cracks that originated in dealloyed cavities. The second case describes high-pressure, high-temperature steam impinging the tubes, leading to deep erosion on the tube`s external side. Erosion-corrosion accelerated on the seawater side because heating decreased the critical flow rate and enhanced formation of deposits which were obstacles to the flow.

  17. Steam generator issues in the United States

    SciTech Connect

    Strosnider, J.R.

    1997-02-01

    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.

  18. Steam generators and related auxiliaries

    SciTech Connect

    Keller, D.L.

    1986-04-01

    The current capability of the power generation industry to supply steam generating equipment for large central fossil stations is much lower than that of several years ago. Volatile energy prices make it very difficult to predict long-term demand changes, but current conditions strongly suggest that demand forecasts and orders will increase from current levels. This combination of circumstances strongly suggest that, while not a certainty, the potential for material and equipment shortages is a very real possibility that belongs in any current assessment of the future of the industry.

  19. Nuclear steam-generator transplant total rises

    SciTech Connect

    Smock, R.

    1982-09-01

    Several utilities with pressurized water reactors (PWRs) are replacing leaking and corroded steam generators. Over half the PWRs face corrosion problems that will cost $50 million to $100 million per unit to correct. An alternative approach of installing new tube sleeves has only had one application. Corrosion prevention still eludes utilities, whose problems differ. Westinghouse units were the first to experience corrosion problems because they have almost all operated for a decade or more. Some advances in condenser and steam-generator technology should extend the component life of younger units, and some leaking PWR tubes can be plugged. Operating differences may explain why PWRs have operated for over 20 years on submarines using phosphate water chemistry, while the use of de-aerators in the secondary-systems of foreign PWRs may explain their better performance. Among the corrective steps recommended by Stone and Webster are tighter chemistry control, better plant layup practices, revamping secondary-system hardware, condensate polishing, and de-aerators. Research continues to find the long-term preventative. 2 tables. (DCK)

  20. PWR steam generator chemical cleaning. Phase II. Final report

    SciTech Connect

    Not Available

    1980-01-01

    Two techniques believed capable of chemically dissolving the corrosion products in the annuli between tubes and support plates were developed in laboratory work in Phase I of this project and were pilot tested in Indian Point Unit No. 1 steam generators. In Phase II, one of the techniques was shown to be inadequate on an actual sample taken from an Indian Point Unit No. 2 steam generator. The other technique was modified slightly, and it was demonstrated that the tube/support plate annulus could be chemically cleaned effectively.

  1. Alcohol LOX Steam Generator Test Experience

    NASA Astrophysics Data System (ADS)

    Schaefer, K.; Dommers, M.

    2004-10-01

    At the DLR test centre in Lampoldshausen there is a long experience in the development of rocket steam generators as a main subsystem for the altitude simulation. The rocket steam generators make it possible to supply the required quantities of steam at short notice with reduced investment and operating costs. The rocket steam generators are based on the combustion of liquid oxygen (LOX) and ethyl alcohol (ALC). The paper deals with the experience of the development of the steam generators and the operation at the altitude simulation P1.0 for satellite propulsion and P4.2 for altitude simulation of AESTUS upper stage engine.

  2. Materials development for a fast breeder reactor steam generator concept

    SciTech Connect

    Sessions, C.E.; Reynolds, S.D. Jr.; Hebbar, M.A.; Lewis, J.F.; Kiefer, J.H.

    1981-11-01

    The progress achieved since 1977 in the important area of materials and processes development of fast reactor steam generator development is summarized. The two distinguishing features of the proposed Westinghouse-Tampa steam generator concept are the convoluted shell expansion joint (CSEJ) and the double-wall tubing with a third fluid leak detection capability. A 2/one quarter/ Cr-1 Mo low alloy steel will be used for all important parts of the generator including the CSEJ and the tubes. Other areas in which progress was made include tube-to-tubesheet (T/TS) welding, post-weld heat treatment (PWHT), tube expansion, and development of materials specifications for prototype and future plant materials. 8 refs.

  3. Specific features of steam condensation inside tubes and channels

    NASA Astrophysics Data System (ADS)

    Mil'man, O. O.; Fedorov, V. A.; Kondrat'ev, A. V.; Ptakhin, A. V.

    2015-04-01

    The results of theoretical and experimental investigations revealed a dependence between the parameters of steam condensation process in tubes and channels on the coolant flow pattern: counter flow, forward flow, and cross flow. The loss of total steam pressure in case of counter flow is higher than it is in case of using a forward or cross flow pattern. During condensation, superheated steam retains its superheated state over the entire tube length if the steam quality at the outlet x 1 > 0, and the superheating temperature depends on the coolant flow pattern. A method for calculating the pressure loss to steam temperature ratio as a function of coolant flow pattern is developed, which is confirmed by experimental data. The ratio or its versions (where c and G are the specific heat capacity and flow rate of coolant) should be regarded as the main parameter in generalizing data.

  4. Machine vision calibration for a nuclear steam generator robot

    SciTech Connect

    Glass, S.W.; Fallon, J.B.; Reinholtz, C.F.; Abbott, A.L.

    1994-12-31

    Inspection and repair of pressurized water reactor steam generators are among the most costly and schedule-critical activities of a refueling outage. These. tasks are highly automated with robots and special tools. This paper describes a method of improving the calibration of a steam generator robot by adding a machine vision computer to the existing tool-head monocular video. The steam generators are heat exchangers containing several thousand tubes ranging from 20 to 40 m in length. Each tube is 1 to 2 cm in diameter with a wall thickness of {approximately} 1 mm. The tubes are welded into a thick tube sheet that caps a hemispherical or quarter-sphere plenum. Practically all work must be performed robotically because the plenum is a high-radiation area. A robotic arm with precise positioning capability must enter the plenum through a 40-cm passageway. Most arms are anchored to either the passageway or the tube sheet. The arm must identify each of the thousands of tubes by row and column number based on the measured robotic joint angles and the calibrated arm-to-tube sheet spatial transform.

  5. Mist/steam cooling in a 180{degree} tube bend

    SciTech Connect

    Guo, T.; Wang, T.; Gaddis, J.L.

    1999-07-01

    An experimental study on mist/steam cooling in a highly heated, horizontal 180{degree} tube bend has been performed. The mist/steam mixture is obtained by blending fine water droplets (3{approximately}15 microns) with the saturated steam at 1.5 bar. The test section consists of a thin wall ({approximately}0.9 mm), welded, circular, stainless steel 180-degree tube (20 mm ID) with a straight section downstream of the curved section, and is heated directly by a DC power supply. The experiment was conducted with steam Reynolds numbers ranging from 10,000 to 35,000, wall superheat up to 300 C, and droplet to steam mass ratio at about 2%. The results show that the heat transfer performance of steam can be significantly improved by adding mist into the main flow. Due to the effect of centrifugal force, the outer wall of the test section always exhibits a higher heat transfer enhancement than the inner wall. The highest enhancement occurs at a location on the outer wall about 45{degree} downstream of the inlet of the test section. Generally, only a small number of droplets can survive the 180{degree} turn and be present in the downstream straight section, as observed by a Phase Doppler Particle Analyzer (PDPA) system. The overall cooling enhancement of the mist/steam flow ranges from 40% to 300%. It increases as the main steam flow increases, but decreases as the wall heat flux increases.

  6. The case for endurance testing of sodium-heated steam generators

    SciTech Connect

    Onesto, A.T.; Zweig, H.R.; Gibbs, D.C. . Rocketdyne Division.); Carlson, R.D. ); Rodwell, E. ); Kakarala, C.R. )

    1993-08-01

    After operating pressurized water reactor (PWR) steam generators in U.S. nuclear plants during the past 33 years and plugging thousands of tubes and replacing numerous steam generators at immense costs, utility and steam generator designers are now confident that they can design, build, and operate PWR steam generators successfully. Deployment of liquid-metal fast breeder reactors (LMFBRs) will likely follow the same scenario if long-term testing is not performed and development completed prior to commercial deployment. A case is made for endurance testing of steam generators to be used in future LMFBRs.

  7. Development of a downhole steam generator system

    SciTech Connect

    Not Available

    1984-04-01

    This report describes the development of a downhole steam generator system for use in enhanced oil recovery. The system is composed of four major components: A state-of-the-art review indicated that advances in technology would be necessary in two areas (high pressure combustion and high temperature packer seals) in order to fabricate a field-worthy system. As a result, two tasks were undertaken which resulted in the development of a novel ceramic-lined combustor and a unique all-metal packer. These elements were incorporated into an overall system design. Key system components were built and tested in the laboratory. The program culminated in a successful simulated downhole test of the entire system, less tube string, at Sandia National Laboratories. 5 references, 41 figures, 9 tables.

  8. Some aspects of two-phase flow, heat transfer and dynamic instabilities in medium and high pressure steam generators

    NASA Astrophysics Data System (ADS)

    Unal, H. C.

    1981-03-01

    Experimental data for void fraction, incipient point of boiling, initial point of net vapor generation, bubble dynamics, dryout, two-phase flow pressure drop and density-wave oscillations were obtained in long, sodium heated steam generator tubes of different geometries for a wide range of operating conditions and at medium and high pressures. These data and data from literature taken in sodium and electrically heated steam generator tubes were correlated. Aspects of two-phase flow, heat transfer and density-wave oscillations in these steam generators disclosed include the distribution factor in small- and medium-size diameter steam generator tubes, the characteristic of the transitions at the incipient point of boiling and initial point of net vapor generation, bubble growth during subcooled nucleate flow boiling, the importance of the equivalent length for dryout in non-uniformly heated steam generator tubes and the mechanisms of density-wave oscillations in once-through steam generator tubes.

  9. Design of Steam Generator for 700 MWe IPHWR

    SciTech Connect

    John, Benny; Ghadge, S.G.

    2006-07-01

    The next stage in the Indian Nuclear Power programme consists of building 700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) units. This involves up-rating of all the plant equipment like Reactor, Steam Generators (SGs), turbo-generator, major pumps etc. The 434 MWth SG used in the current generation 540 MWe IPHWRs, is a mushroom type, inverted U tube, natural circulation steam generator. The challenge, for evolution of the 540 MWth SG design, was to keep the tube diameter, tube pitch and outer diameter of the steam generator sections identical to the 434 MWth SG as far as possible. Further, the temperature difference between the primary inlet and outlet temperatures from SG was also to be maintained equal to that of 434 MWth SG. The paper describes the thermal hydraulic studies carried out for arriving at an optimal process design of 540 MWth SG. The studies were carried out using the validated 1-D code developed in house. The paper covers the issues like, extraction of the extra 106 MWth power, maintenance of a good circulation ratio under all operating conditions, additional capacity requirements of steam separators and accommodation of internals in the given space. (authors)

  10. Steam Generator Group Project. Task 6. Channel head decontamination

    SciTech Connect

    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.

  11. Analysis of thermosyphoning in a steam generator model

    SciTech Connect

    Martinez, M.J.; Martinez, G.M.

    1991-10-01

    An analysis of thermosyphoning in a stream generator model is presented. The model considers the transient development of buoyancy-driven steam flow in the steam generator tubing, secondary side heat transfer and an inlet plenum mixing model. Numerical solutions are obtained for conditions intended to simulate the natural circulation phenomena in a 3-Loop pressurized water reactor in a loss-of-coolant accident scenario. The relation between the circulation rate and the heating rate is determined. The sensitivity of the model to various key parameters is examined. 16 refs.

  12. A PDE model of a waterwalls steam generation process.

    PubMed

    Delgadillo, Miguel A; Suárez, Dionisio A; Moreno, Jaime A

    2008-10-01

    This paper describes a model of a forced circulation waterwalls steam generator, derived from first principles. The distributed parameter criteria were applied to the heat transfer process and to the steam production inside the waterwalls. The model is capable of representing swell and shrink effects as well as the condensation-vaporization phenomena that take place inside the waterwall tubes, when large drum steam pressure variations are introduced. The swell and shrink effects are responsible for water displacement from the waterwalls to the drum and from the drum to the waterwalls. Open loop simulated test were produced with the steam pressure disturbance. Closed loop tests, including the models of the drum level and the combustion system and their control systems are presented. PMID:18692846

  13. Steam condensation inside a vertical tube with noncondensable gas

    SciTech Connect

    Araki, Hidefumi; Kataoka, Yoshiyuki; Murase, Michio

    1994-12-31

    Passive containment cooling systems are being studied extensively in order to enhance reactor simplicity. One promising concept is a system equipped with condensers submerged in pools located outside the primary containment vessel (PCV). Assuming a loss-of-coolant accident, steam flows into the condensers together with nitrogen, which fills the containment drywell. Then steam is condensed in the tubes, and the decay heat is released to the atmosphere by vaporization of the pool water, suppressing and pressure of the PCV below the design pressure. In the foregoing process, the noncondensable gas greatly lowers the heat transfer coefficient (HTC) inside the condenser tubes. Therefore the effect of noncondensable gases should be clarified to predict HTCs under such conditions. The objective of this study are to measure local HTCs inside a condenser tube and to develop their evaluation methods in the presence of noncondensable gas.

  14. Packer cooling system for a downhole steam generator assembly

    SciTech Connect

    Baugh, J.L.; Mooney, F.X.; Vandevier, J.E.

    1989-02-21

    An apparatus is described for providing electrical power to a downhole stream generator in a cased well, comprising in combination: a packer supported on a string of tubing; a connector box; an electrical cable extending from the surface alongside the tubing into the aperture and through the conduit into the connector box; a plurality of electrical conductors extending between the steam generator and engaging a lower end of each electrical connector in the connection plate; and cooling fluid passage means extending through the packer for circulating cooling fluid pumped down from the surface through the packer and back up the well to the surface. The patent also describes a method for installing an operating a steam generator in a well.

  15. Control system for fluid heated steam generator

    DOEpatents

    Boland, James F.; Koenig, John F.

    1985-01-01

    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.

  16. Control system for fluid heated steam generator

    DOEpatents

    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.

  17. Horizontal steam generators: Problems and prospects

    NASA Astrophysics Data System (ADS)

    Trunov, N. B.; Ryzhov, S. B.; Davidenko, S. E.

    2011-03-01

    Main results of the 40-year experience gained from operation of horizontal steam generators in VVER-type reactor installations used in Russia and many foreign countries are described. Existing unresolved problems are pointed out.

  18. Downhole steam generator subject of Sandia tests

    SciTech Connect

    Not Available

    1981-05-01

    The first field test of a down-hole steam generator developed to inject hot steam directly into deeply buried heavy oil reservoirs began in mid-April near Long Beach, CA. The 23-ft-long, 6-in.-diameter generator, developed by Sandia National Laboratories, will produce up to 800 cu ft of 500 F steam a minute (1.2 MW thermal) at the bottom of a 2500-ft well. Goals of the test are to demonstrate the feasibility of operating the generator at realistic depths and to determine its overall performance and environmental impact. Development of the generator is part of the US Department of Energy's Project Deep Steam to identify techniques for recovering heavy oil from deeply buried (greater than 2500 ft) reservoirs.

  19. 49 CFR 229.114 - Steam generator inspections and tests.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 4 2013-10-01 2013-10-01 false Steam generator inspections and tests. 229.114... Generators § 229.114 Steam generator inspections and tests. (a) Periodic steam generator inspection. Except as provided in § 229.33, each steam generator shall be inspected and tested in accordance...

  20. 49 CFR 229.114 - Steam generator inspections and tests.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 4 2014-10-01 2014-10-01 false Steam generator inspections and tests. 229.114... Generators § 229.114 Steam generator inspections and tests. (a) Periodic steam generator inspection. Except as provided in § 229.33, each steam generator shall be inspected and tested in accordance...

  1. 49 CFR 229.114 - Steam generator inspections and tests.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 4 2012-10-01 2012-10-01 false Steam generator inspections and tests. 229.114... Generators § 229.114 Steam generator inspections and tests. (a) Periodic steam generator inspection. Except as provided in § 229.33, each steam generator shall be inspected and tested in accordance...

  2. A tube-in-tube thermophotovoltaic generator

    SciTech Connect

    Ashcroft, J.; Campbell, B.; Depoy, D.

    1996-12-31

    A thermophotovoltaic device includes at least one thermal radiator tube, a cooling tube concentrically disposed within each thermal radiator tube and an array of thermophotovoltaic cells disposed on the exterior surface of the cooling tube. A shell having a first end and a second end surrounds the thermal radiator tube. Inner and outer tubesheets, each having an aperture corresponding to each cooling tube, are located at each end of the shell. The thermal radiator tube extends within the shell between the inner tubesheets. The cooling tube extends within the shell through the corresponding apertures of the two inner tubesheets to the corresponding apertures of the two outer tubesheets. A plurality of the thermal radiator tubes can be arranged in a staggered or an in-line configuration within the shell.

  3. Tube-in-tube thermophotovoltaic generator

    SciTech Connect

    Ashcroft, John; Campbell, Brian; DePoy, David

    1998-01-01

    A thermophotovoltaic device includes at least one thermal radiator tube, a cooling tube concentrically disposed within each thermal radiator tube and an array of thermophotovoltaic cells disposed on the exterior surface of the cooling tube. A shell having a first end and a second end surrounds the thermal radiator tube. Inner and outer tubesheets, each having an aperture corresponding to each cooling tube, are located at each end of the shell. The thermal radiator tube extends within the shell between the inner tubesheets. The cooling tube extends within the shell through the corresponding apertures of the two inner tubesheets to the corresponding apertures of the two outer tubesheets. A plurality of the thermal radiator tubes can be arranged in a staggered or an in-line configuration within the shell.

  4. Tube-in-tube thermophotovoltaic generator

    DOEpatents

    Ashcroft, J.; Campbell, B.; DePoy, D.

    1998-06-30

    A thermophotovoltaic device includes at least one thermal radiator tube, a cooling tube concentrically disposed within each thermal radiator tube and an array of thermophotovoltaic cells disposed on the exterior surface of the cooling tube. A shell having a first end and a second end surrounds the thermal radiator tube. Inner and outer tubesheets, each having an aperture corresponding to each cooling tube, are located at each end of the shell. The thermal radiator tube extends within the shell between the inner tubesheets. The cooling tube extends within the shell through the corresponding apertures of the two inner tubesheets to the corresponding apertures of the two outer tubesheets. A plurality of the thermal radiator tubes can be arranged in a staggered or an in-line configuration within the shell. 8 figs.

  5. Pulsed high-pressure (PHP) drain-down of steam generating system

    SciTech Connect

    Petrusek, R.A.

    1991-03-19

    This patent describes an improved method of draining down contained reactor-coolant water from the inverted vertical U-tubes of at least one 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 having a channel head at its lower end including a vertical dividing wall defining a primary water inlet side and a primary water outlet side of the generator, the steam generator having chemical volume control system means and residual heat removal system means, and the steam generator being 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, and the reactor being in communication with pressurizer means and comprising the steps of 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 tubesheet 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, the improvement in combination therewith for substantially increasing the effectiveness and efficiency of such water removal from the tubes. It includes determining the parameters effecting a first average volumetric rate of removal for a predetermined period of time, infra, of the reactor-coolant water from the inverted vertical U-tubes, the specific unit for the first average volumetric rate expressing properties identical with the properties expressed in a second average volumetric rate maintained in a later mentioned step.

  6. Sullair low pressure downhole steam generator system

    SciTech Connect

    Klingler, R.P.

    1982-01-01

    Scientists and engineers are continually searching for techniques to release more oil from known reservoirs to improve productivity and lessen dependence on new finds. Based on a record of success dating to the early 1960s, thermal methods, and in particular methodology for steam treating deep reservoirs, have become an area of intense activity. In the U.S. alone, it has been reported that ca 300,000 bopd was produced in 1981 by traditional surface steam methods. Of the thermal techniques emerging, downhole steam generation is of particular interest in this discussion. 11 references.

  7. Electric-arc steam plasma generator

    NASA Astrophysics Data System (ADS)

    Anshakov, A. S.; Urbakh, E. K.; Radko, S. I.; Urbakh, A. E.; Faleev, V. A.

    2015-01-01

    Investigation results on the arc plasmatorch for water-steam heating are presented. The construction arrangement of steam plasma generator with copper electrodes of the stepped geometry was firstly implemented. The energy characteristics of plasmatorch and erosion of electrodes reflect the features of their behavior at arc glow in the plasma-forming environment of steam. The results of numerical study of the thermal state of the composite copper-steel electrodes had a significant influence on optimization of anode water-cooling aimed at improvement of its operation life.

  8. Operational experiences of a downhole steam generator

    NASA Astrophysics Data System (ADS)

    Marshall, B. W.

    The US Department of Energy supported the development of downhole steam generators for enhanced oil recovery as a part of Project DEEP STEAM. A final step in the development program was to deploy a downhole steam generator in the field to demonstrate its reliable operation and to evaluate the effect of the combined steam/exhaust products effluent on the reservoir. Sandia National Laboratories entered into an agreement with the City of Long Beach to place two direct contact, high pressure combustors in the Wilmington Field in Long Beach, California. These units one downhole and the other on the surface, have now been operated for a few months and gas communication with the production wells measured. The operational experience of this field experiment are discussed.

  9. Operational experiences of a downhole steam generator

    SciTech Connect

    Marshall, B.W.

    1982-01-01

    The US Department of Energy has supported the development of downhole steam generators for enhanced oil recovery as a part of Project DEEP STEAM. A final step in the development program was to deploy a downhole steam generator (DHSG) in the field to demonstrate its reliable operation and to evaluate the effect of the combined steam/exhaust products effluent on the reservoir. Sandia National Laboratories entered into an agreement with the City of Long Beach to place two direct contact, high pressure combustors in the Wilmington Field in Long Beach, California. These units one downhole and the other on the surface, have now been operated for a few months and gas communication with the production wells measured. The operational experience of this field experiment are discussed.

  10. Modeling of a horizontal steam generator for the submerged nuclear power station concept

    SciTech Connect

    Palmrose, D.E.; Herring, J.S.

    1993-05-01

    A submerged nuclear power station has been proposed as an alternative power station with a relatively low environmental impact for use by both industrialized and developing countries. The station would be placed 10 m above the seabed at a depth of 30--100 m and a distance of 10--30 km from shore. The submerged nuclear power station would be manufactured and refueled in a central facility, thus gaining the economies of factoryfabrication and the flexibility of short-lead-time deployment. To minimize the size of the submerged hull, horizontal steam generators are proposed for the primary-to-secondary heat transfer, instead of the more traditional vertical steam generators. The horizontal steam generators for SNPS would be similar in design to the horizontal steam generators used in the N-Reactors except the tube orientation is horizontal (the tube`s inlet and outlet connection points on the tubesheet are at the same elevation). Previous RELAP5 input decks for horizontal steam generators have been either very simplistic (Loviisa PWR) or used a vertical tube orientation (N-Reactor). This paper will present the development and testing of a RELAP5 horizontal steam generator model, complete with a simple secondary water level control system, that accounts for the dynamic flow conditions which exist inside horizontal steam generators.

  11. Modifying steam generator corrosion behavior via chemical cleaning

    SciTech Connect

    Sweeney, K.; Neese, K.

    1994-12-31

    A steam generator chemical cleaning program was conducted in Palo Verde nuclear generating station (PVNGS) units 2 and 3 in 1994. This effort represented the first full-bundle chemical cleaning of a recirculating steam generator in the United States. The objectives of the process were: (1) to remove deposits in the upper bundle regions, which were identified by eddy-current analysis and linked to a free-span outside-diameter stress corrosion cracking (ODSCC) condition; (2) to remove tube scale deposits that interfere with heat transfer and may contain undesirable contaminants; (3) to remove deposits from the surface of the tube sheet and the flow distribution plate; and (4) to remove deposits from the drilled hole crevices in the FDP, which may be contributing to low recirculation ratios and upper bundle transition boiling. The Electric Power Research Institute/Steam Generator Owners` Group low-temperature process, modified to include {open_quotes}crevice cleaning{close_quotes} and {open_quotes}passivation{close_quotes} steps, was selected as the best method. Babock & Wilcox Nuclear Technologies was selected as the vendor.

  12. Downhole steam generator at Kern River

    SciTech Connect

    Rintoul, B.

    1980-05-01

    Testing of a prototype down-hole steam generator for use in enhanced oil recovery (EOR) operations has begun at a heavy oil reservoir in the Kern River oil field in California. Steam and combustion gases are directed into an 800-ft-deep reservoir through a standard surface steam delivery system, although the system is designed to function at depths to 4500 ft. Present steam injection techniques require one-third of the oil recovered to be used to fuel the injection system, and the boilers require scrubbers to control emissions to specifications. The down-hole system is expected to use only 2/3 as much fuel as the conventional systems and to have less impact on air quality.

  13. 49 CFR 229.105 - Steam generator number.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 4 2011-10-01 2011-10-01 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...

  14. 49 CFR 229.105 - Steam generator number.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 4 2012-10-01 2012-10-01 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...

  15. 49 CFR 229.105 - Steam generator number.

    Code of Federal Regulations, 2010 CFR

    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...

  16. 49 CFR 229.105 - Steam generator number.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 4 2013-10-01 2013-10-01 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...

  17. 49 CFR 229.105 - Steam generator number.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 4 2014-10-01 2014-10-01 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...

  18. Downhole steam generator for heavy oil recovery

    SciTech Connect

    Montgomery, M.C.

    1982-01-01

    The HI-KO steam generator (HI-KO DSG) is operated electronically and is designed to operate efficiently and cost effectively at any depth, temperature, and injection pressure with minimal labor and maintenance. The generator is transported easily and is free of pollutants, both above and below ground. Moreover, the well needs no special cement job to prepare it for steaming; nor are corrosive gases created in the liner. The HI-KO DSG operates at ca 100% efficiency. Being controlled from the surface, the btu's and temperature going into the formation are monitored at all times.

  19. COMMIX analysis of the sodium heated helical coil steam generator

    SciTech Connect

    Kakarala, C.R.; Burge, S.W.; Sha, W.T.

    1987-01-01

    This paper describes the COMMIX-HCSG computer program and compares predictions to data obtained from performance tests on a 76 MWt Helical Coil Steam Generator (HCSG) test unit. COMMIX-HCSG is a multi-dimensional thermal/hydraulic code that models both steady state and transient operation of an HCSG. The code solves a system of Navier-Stokes continuum equations that have been modified with a combination of volume and directional surface porosities and distributed resistances. This formulation properly accounts for the presence of tube bundle, supports, and baffles on the shell side of the steam generator. Turbulence models and heat transfer and pressure drop equations are used as applicable for the different regions including the upper plenum, the tube bundle, and the lower plenum of the HCSG. The data was obtained from performance tests conducted in early 1987 on the 76 MWt HCSG test unit at the Energy Technology Engineering Center (ETEC). The test unit contains over 700 instruments. HCSG development and tests are carried out as part of the Department of Energy program to develop reliable and economical liquid metal heated steam generators.

  20. Feasibility evaluation of a downhole steam generator

    SciTech Connect

    Wright, D.E.; Binsley, R.L.

    1982-04-01

    A discussion is presented of a downhole steam generator (DHSG); benefits and features are described. The work was done under the management of Sandia Laboratories as part of the U.S. DOE's Project Deep Steam. The effort began with a comparison of alternate ways of generating steam downhole. The most economical was found to be the indirect system, where heat from low-pressure (approximately 5-atm (500-kPal))combustion of fuel oil and air is transferred to high pressure water. A feasibility test unit was constructed and tested extensively; results are described. Comparative economic analyses indicate that the DHSG is competitive with conventional systems even at moderate depths. Below 2,500 ft (762 m) it is a clear choice. 3 refs.

  1. Mist/steam cooling in a heated horizontal tube -- Part 1: Experimental system

    SciTech Connect

    Guo, T.; Wang, T.; Gaddis, J.L.

    2000-04-01

    To improve the airfoil cooling significantly for the future generation of advanced turbine systems (ATS), a fundamental experimental program has been developed to study the heat transfer mechanisms of mist/steam cooling under highly superheated wall temperatures. The mist/steam mixture was obtained by blending fine water droplets (3 {approximately} 15 {micro}m in diameter) with the saturated steam at 1.5 bars. Two mist generation systems were tested by using the pressure atomizer and the steam-assisted pneumatic atomizer, respectively. The test section, heated directly by a DC power supply, consisted of a thin-walled ({approximately} 0.9 mm), circular stainless steel tube with an ID of 20 mm and a length of 203 mm. Droplet size and distribution were measured by a phase Doppler particle analyzer (PDPA) system through view ports grafted at the inlet and the outlet of the test section. Mist transportation and droplet dynamics were studied in addition to the heat transfer measurements. The experiment was conducted with steam Reynolds numbers ranging from 10,000 to 35,000, wall superheat up to 300 C, and droplet mass ratios ranging from 1 {approximately} 6%.

  2. Recent operating experiences with steam generators in Japanese NPPs

    SciTech Connect

    Yashima, Seiji

    1997-02-01

    In 1994, the Genkai-3 of Kyushu Electric Power Co., Inc. and the Ikata-3 of Shikoku Electric Power Co., Inc. started commercial operation, and now 22 PWR plants are being operated in Japan. Since the first PWR plant now 22 PWR plants are being operated in was started to operate, Japanese PWR plants have had an operating experience of approx. 280 reactor-years. During that period, many tube degradations have been experienced in steam generators (SGs). And, in 1991, the steam generator tube rupture (SGTR) occurred in the Mihama-2 of Kansai Electric Power Co., Inc. However, the occurrence of tube degradation of SGs has been decreased by the instructions of the MITI as regulatory authorities, efforts of Electric Utilities, and technical support from the SG manufacturers. Here the author describes the recent SGs in Japan about the following points. (1) Recent Operating Experiences (2) Lessons learned from Mihama-2 SGTR (3) SG replacement (4) Safety Regulations on SG (5) Research and development on SG.

  3. Steam Generator Group Project. Annual report, 1982

    SciTech Connect

    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.

  4. Radiological assessment of steam generator repair and replacement

    SciTech Connect

    Parkhurst, M.A.; Rathbun, L.A.; Murphy, D.W.

    1983-12-01

    Previous analyses of the radiological impact of removing and replacing corroded steam generators have been updated based on experience at Surry Units 1 and 2 and Turkey Point Units 3 and 4. The sleeving repairs of degraded tubes at San Onofre Unit 1, Point Beach Unit 2, and R.E. Ginna are also analyzed. Actual occupational doses incurred during application of the various technologies used in repairs have been included, along with radioactive waste quantities and constituents. Considerable progress has been made in improving radiation protection and reducing worker dose by the development of remotely controlled equipment and the implementation of dose reduction strategies that have been successful in previous repair operations.

  5. Dynamic stability experiments in sodium-heated steam generators. [LMFBR

    SciTech Connect

    France, D.M.; Roy, R.; Carlson, R.D.; Chiang, T.

    1984-01-01

    Seventy-two dynamic stability tests were performed in the sodium-heated boiling-water test facility at Argonne National Laboratory. A full-scale LMFBR steam generator tube was employed as the test section operating over the water parameter ranges of 6.9 to 15.9 MPa pressure and 170 to 800 kg/m/sup 2/.s mass flux. The stability thresholds from the test compared well to the predictions of a modified version of a correlation equation recently published by other investigators. Typical experimental data and the modified correlation equation are presented.

  6. Enhanced tubes for steam condensers. Volume 1, Summary of condensation and fouling; Volume 2, Detailed study of steam condensation

    SciTech Connect

    Webb, R.L.; Chamra, L.; Jaber, H.

    1992-02-01

    Electric utility steam condensers typically use plain tubes made of titanium, stainless steel, or copper alloys. Approximately two-thirds of the total thermal resistance is on the water side of the plain tube. This program seeks to conceive and develop a tube geometry that has special enhancement geometries on the tube (water) side and the steam (shell) side. This ``enhanced`` tube geometry, will provide increased heat transfer coefficients. The enhanced tubes will allow the steam to condense at a lower temperature. The reduced condensing temperature will reduce the turbine heat rate, and increase the plant peak load capability. Water side fouling and fouling control is a very important consideration affecting the choice of the tube side enhancement. Hence, we have consciously considered fouling potential in our selection of the tube side surface geometry. Using appropriate correlations and theoretical models, we have designed condensation and water side surface geometries that will provide high performance and be cleanable using sponge ball cleaning. Commercial tube manufacturers have made the required tube geometries for test purposes. The heat transfer test program includes measurement of the condensation and water side heat transfer coefficients. Fouling tests are being run to measure the waterside fouling resistance, and to the test the ability of the sponge ball cleaning system to clean the tubes.

  7. Solar steam generation by heat localization.

    PubMed

    Ghasemi, Hadi; Ni, George; Marconnet, Amy Marie; Loomis, James; Yerci, Selcuk; Miljkovic, Nenad; Chen, Gang

    2014-01-01

    Currently, steam generation using solar energy is based on heating bulk liquid to high temperatures. This approach requires either costly high optical concentrations leading to heat loss by the hot bulk liquid and heated surfaces or vacuum. New solar receiver concepts such as porous volumetric receivers or nanofluids have been proposed to decrease these losses. Here we report development of an approach and corresponding material structure for solar steam generation while maintaining low optical concentration and keeping the bulk liquid at low temperature with no vacuum. We achieve solar thermal efficiency up to 85% at only 10 kW m(-2). This high performance results from four structure characteristics: absorbing in the solar spectrum, thermally insulating, hydrophilic and interconnected pores. The structure concentrates thermal energy and fluid flow where needed for phase change and minimizes dissipated energy. This new structure provides a novel approach to harvesting solar energy for a broad range of phase-change applications. PMID:25043613

  8. Evaluation of alternatives in downhole steam generation

    SciTech Connect

    Not Available

    1981-09-01

    The objective is to evaluate two alternative approaches, taken by Sandia and World Energy Systems in their development of downhole steam generators, in terms of the requirements for commercialization and the technical and economic goals which each one must reach in order to satisfy those requirements. The conclusions are as follows: (1) The high-pressure downhole steam generator being developed by Sandia has potential for commercialization for production of heavy oil in the next few years. (2) The critical performance parameter is oil yield and can be expressed in terms of bbl oil/million Btu of steam generated; the yield which is required for clear economic attractiveness of the high-pressure generator is 1.2 bbl/10/sup 6/ Btu which corresponds to 5.3 bbl steam/bbl oil. (3) The downhole hydrogen/oxygen burner being developed by World Energy Systems has potential for commercialization for production of resources which are now unproducible or uneconomical. (4) The critical performance provided in nearly 1100 pre- and post-training forms returned by the solarr and 201 stream sediment samples. Statistical and areal distributions of uranium and possible uranium-related variables are displayed. A generalized geologic map of the survey area is provided, and pertinent geologic factors which may be of significance in evaluating the potential for uranium mineralization are briefly discussed. Ground water data indicate that high uranium values occur almost exclusively in the western portion of the quadrangle along the eastern portion of the Williston Basin. These high uranium values occur primarily in Pleistocene delta deposits and in glacial outwash and till. Groundwater in this area is geographically associated with high values of calcium, magnesium, manganese, potassium, selenium, strontium, sulfate, and total alkalinity. Stream sediment data indicate high uranium value the relative concentration of Sm/sup 2 +/ and Sm/sup 3 +/ ions changes with the change of composition.

  9. Determination of steam wetness in the steam-generating equipment of nuclear power plants

    NASA Astrophysics Data System (ADS)

    Gorburov, V. I.; Gorburov, D. V.; Kuz'min, A. V.

    2012-05-01

    Calculation and experimental methods for determining steam wetness in horizontal steam generators for nuclear power stations equipped with VVER reactors, namely, the classic salt technique and calculations based on operating parameters are discussed considered and compared.

  10. Analytic prediction of complex unsteady flow fields in preheat PWR steam generators: Final report

    SciTech Connect

    Stuhmiller, J.H.; Masiello, P.J.; Kan, K.K.; Chilukuri, R.

    1988-05-01

    Turbulent buffeting and associated tube vibration may cause unacceptable levels of fretting wear within steam generator tube banks. Fretting wear occurs when the vibrating tubes impact against support plates and anti-vibration bars. The goal of this work is to investigate the feasibility of predicting stem generator tube buffeting and vibration in turbulent crossflow using computer models that avoid empiricisms. The value of such a technquie lies in the ability to verify its predictions with separate effect tests, that are more controlled and readily available, and in the greater reliability of its predictions in situations that have not been tested experimentally.

  11. New treatment concept for steam generators technical aspects

    SciTech Connect

    Lindstrom, A.; Wirendal, B.O.; Lindberg, M.

    2007-07-01

    The project that will be described is a co-operation development project (SAGA) between Studsvik and the Ringhals NPP. The objective for this development project was, to show that it is possible to perform effective waste treatment of a Steam Generator(SG), to minimize the volume that in the end will have to be finally disposed of and to recycle as much of the metals as possible. Another objective for the project was to do this in a safe way and without a large dose load to the personnel. The treatment concept contains the whole chain of activities from loading of the steam generator at Ringhals NPP onto the special vessel M/S Sigyn, and the transportation of the SG from Ringhals NPP on the west coast of Sweden to Studsvik on the east coast, to the recycling of the metals and the packing of waste in final packages suitable for disposal. The volume for a final repository before treatment was about 400 m3 for the SG and after treatment the volume for final disposal is < 35 m{sup 3} which gives a volume reduction factor of about 11. The amount of material from the steam generator that has clearance for free release is 75-80 % of the weight. A project is started to analyse the experience from the project above and to come forward with recommendations for how to lower the dose exposure, minimize the secondary waste for final disposal and to decrease the treatment time. Some actions are already taken: - A new larger treatment facility is built at Studsvik, > 1000 m{sup 2}, planned to be operational in April 2007. - Investments in a larger band saw. - Improvements of the blasting equipment. - Improvements of the method of segmentation of the tube bundle. - Improvements of the method of volume reduction for the tube bundle. (authors)

  12. 45. William E. Barrett, Photographer, August 1975. EARLY STEAM GENERATING ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    45. William E. Barrett, Photographer, August 1975. EARLY STEAM GENERATING UNIT USED TO PRODUCE ELECTRICITY FOR MANUFACTURING OPERATIONS AND FOR THE TOWN OF RAINELLE. STEAM ENGINE IS A HAMILTON CORLISS. - Meadow River Lumber Company, Highway 60, Rainelle, Greenbrier County, WV

  13. Structure interaction due to thermal bowing of shrouds in steam generator of gas-cooled reactor

    SciTech Connect

    Woo, H.H.

    1981-01-01

    The design of the gas-cooled reactor steam generators includes a tube bundle support plate system which restrains and supports the helical tubes in the steam generator. The support system consists of an array of radially oriented, perforated plates through which the helical tube coils are wound. These support plates have tabs on their edges which fit into vertical slots in the inner and outer shrouds. When the helical tube bundle and support plates are installed in the steam generator, they most likely cannot fit evenly between the inner and outer shrouds. This imperfection leads to different gaps between two extreme sides of the tube bundle and the shrouds. With different gaps through the tube bundle height, the helium flow experiences different cooling effects from the tube bundle. Hence, the temperature distribution in the shrouds will be non-uniform circumferentially since their surrounding helium flow temperatures are varied. These non-uniform temperatures in the shrouds result in the phenomenon of thermal bowing of shrouds.

  14. Proceedings: 2003 Steam Generator Secondary Side Management Conference

    SciTech Connect

    2003-06-01

    With more utilities replacing steam generators and applying for (and receiving) license renewal and uprates, it is imperative that we coordinate our efforts for improved steam generator management. This report contains the work presented at EPRI's 2003 Steam Generator Secondary Side Management Conference, where 35 papers were presented on current issues, research, and utility experiences involving corrosion product generation and transport, deposit control and mitigation, deposit consolidation and removal, and short- and long-term strategic planning.

  15. Comparative evaluation of surface and downhole steam-generation techniques

    SciTech Connect

    Hart, C.

    1982-01-01

    It has long been recognized that the application of heat to reservoirs containing high API gravity oils can substantially improve recovery. Although steam injection is currently the principal thermal recovery method, heat transmission losses associated with delivery of the steam from the surface generators to the oil-bearing formation has limited conventional steam injection to shallow reservoirs. The objective of the Department of Energy's Project DEEP STEAM is to develop the technology required to economically produce heavy oil from deep reservoirs. The tasks included in this effort are the development and evaluation of thermally efficient delivery systems and downhole steam generation systems. This paper compares the technical and economic performance of conventional surface steam drives, which are strongly influenced by heat losses, with (a) thermally efficient delivery (through insulated strings) of surface generated steam, (b) low pressure combustion downhole steam generation, (c) high pressure combustion downhole steam generation using air as the oxygen source, and (d) high pressure combustion downhole steam generation substituting pure oxygen for air. The selection of a preferred technology based upon either total efficiency or cost is found to be strongly influenced by reservoir depth, steam mass flow rate, and sandface steam quality. Therefore, a parametric analysis has been performed which examines varying depths, injection rates and steam qualities. Results indicate that the technologies are not readily distinguishable for low injectivity reservoirs in which conventional steam drives are feasible. However, high injection rates produce a notable cost difference between high pressure combustion systems and the other technologies. Issues that must be addressed before gaining further insight into the economic viability of downhole steam generation are discussed.

  16. Steam generators secondary side chemical cleaning at Point Lepreau using the Siemens high temperature process

    SciTech Connect

    Verma, K.; MacNeil, C.; Odar, S.; Kuhnke, K.

    1997-02-01

    This paper describes the chemical cleaning of the four steam generators at the Point Lepreau facility, which was accomplished as a part of a normal service outage. The steam generators had been in service for twelve years. Sludge samples showed the main elements were Fe, P and Na, with minor amounts of Ca, Mg, Mn, Cr, Zn, Cl, Cu, Ni, Ti, Si, and Pb, 90% in the form of Magnetite, substantial phosphate, and trace amounts of silicates. The steam generators were experiencing partial blockage of broached holes in the TSPs, and corrosion on tube ODs in the form of pitting and wastage. In addition heat transfer was clearly deteriorating. More than 1000 kg of magnetite and 124 kg of salts were removed from the four steam generators.

  17. Comparative evaluation of surface and downhold steam generation techniques

    SciTech Connect

    Hart, C.M.

    1982-01-01

    The objective of the Department of Energy's Project DEEP STEAM is to develop the technology required to economically produce heavy oil from deep reservoirs. The tasks included in this effort are the development and evaluation of thermally efficient delivery systems and downhole steam generation systems. This paper compares the technical and economic performance of conventional surface steam drives, which are strongly influenced by heat losses, with (a) thermally efficient delivery (through insulated strings) of surface generated steam, (b) low pressure combustion downhole steam generatjion, (c) high pressure combustion downhole steam generation using air as the oxygen source, and (d) high pressure combustion downhole steam generation substituting pure oxygen for air. The selection of a preferred technology based upon either total efficiency or cost is found to be strongly influenced by reservoir depth, steam mass flow rate, and sandface steam quality. Therefore, a parametric analysis has been performed which examines varying depths, injection rates and steam qualities. Results indicate that the technologies are not readily distinguishable for low injectivity reservoirs in which conventional steam drives are feasible. However, high injection rates produce a notable cost difference between high pressure combustion systems and the other technologies. Issues that must be addressed before gaining further insight into the economic viability of downhole steam generatjion are discussed. (JMT)

  18. Comparative evaluation of surface and downhole steam-generation techniques

    NASA Astrophysics Data System (ADS)

    Hart, C.

    The application of heat to reservoirs containing high API gravity oils can substantially improve recovery. Although steam injection is currently the principal thermal recovery method, heat transmission losses associated with delivery of the steam from the surface generators to the oil bearing formation has limited conventional steam injection to shallow reservoirs. The objective of the Department of Energy's Project DEEP STEAM is to develop the technology required to economically produce heavy oil from deep reservoirs. The tasks included in this effort are the development and evaluation of thermally efficient delivery systems and downhole steam generation systems. The technical and economic performance of conventional surface steam drives, which are strongly influenced by heat losses are compared. The selection of a preferred technology based upon either total efficiency or cost is found to be strongly influenced by reservoir depth, steam mass flow rate, and sandface steam quality.

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

    NASA Astrophysics Data System (ADS)

    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.

  20. Performance benefits of the direct generation of steam in line-focus solar collectors

    NASA Astrophysics Data System (ADS)

    May, E. K.; Murphy, L. M.

    1983-05-01

    The performance benefits of the direct (in situ) generation of steam in the receiver tube of a line-focus solar collector are assessed in this paper. Compared to existing technology using steam-flash or unfired boiler systems, the in situ technique could reduce the delivered cost of steam in excess of 25 percent. The analysis indicates that two-phase flow instabilities, if present, can be readily controlled, and that the possibility of freezing is not an impediment to the use of water in cold climates.

  1. Economizer recirculation for low-load stability in heat recovery steam generator

    SciTech Connect

    Cuscino, R.T.; Shade, R.L. Jr.

    1986-04-15

    An economizer system is described for heating feedwater in a heat recovery steam generator which consists of: at least first and second economizer tube planes; each of the economizer tube planes including a plurality of generally parallel tubes; the tubes being generally vertically disposed; each of the economizer tube planes including a top header and a bottom header; all of the plurality of tubes in each economizer tube plane being connected in parallel to their top and bottom headers whereby parallel feedwater flow through the plurality of tubes between the top and bottom headers is enabled; one of the top and bottom headers being an inlet header; a second of the top and bottom headers being an outlet header; a boiler feed pump; the boiler feed pump being effective for applying a flow of feedwater to the inlet header; means for serially interconnecting the economizer tube planes; the means for serially interconnecting including means for flowing the feedwater upward and downward in tubes of alternating ones of the economizer tube planes between the inlet header and the outlet header; means for conveying heated feedwater from the outlet header to a using process; means for recirculating at least a portion of the heated feedwater from the outlet header to an inlet of the boiler feed pump; and the means for recirculating including means for relating the portion to a steam load in the using process whereby an increased flow is produced through all of the economizer tube planes at values of the steam load below a predetermined value and a condition permitting initiation of reverse flow in any of the tubes is substantially reduced.

  2. Evaluation of eddy-current procedures for measuring wear scars in preheat steam generators

    SciTech Connect

    Brown, S.D.

    1985-04-01

    Tests show that flat wear scar procedures will provide more accurate measurements of the depth of wear scars in steam generator tubes if they are supplemented by two new techniques. Used together, these methods can detect as little as 5% increase in scar depth.

  3. Review of the data bases for making decisions regarding Trojan steam generator replacement options

    SciTech Connect

    Johnson, A.B. Jr.; Gilbert, E.R.

    1992-03-01

    The central focus for this assessment has been to compare the corrosion behavior of two steam generator (SG) tube materials: Inconel 600 TT and Inconel 690 TT from (a) SG operating experience, and (b) laboratory data. The scope and results of the comparisons are summarized in this section. They provide the basis for projecting SG longevity.

  4. Vapor generator steam drum spray head

    DOEpatents

    Fasnacht, Jr., Floyd A.

    1978-07-18

    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. Remote installation of PWR steam generator nozzle dams

    SciTech Connect

    Ashton, A.

    1994-12-31

    Funded by a grant from the US Department of Energy, prototype equipment for remote installation of steam generator nozzle dams has been developed. The new nozzle dam design eliminates the need for personnel entry into the steam generator bowl and does not require installation of a sophisticated robotic arm. An innovative new bolt allows simple remote attachment to the existing nozzle seal ring in Westinghouse-type steam generators. Installation is performed manually away from the high radiation emanating from the steam generator manway.

  6. Modeling local chemistry in PWR steam generator crevices

    SciTech Connect

    Millett, P.J.

    1997-02-01

    Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledge of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.

  7. Comments on US LMFBR steam generator base technology

    SciTech Connect

    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.

  8. Modeling and field studies of fouling in once-through steam generators

    SciTech Connect

    Thompson, R.; Gaudreau, T.

    1995-12-31

    Efforts of the past 10 years to minimize fouling of the Crystal River-3 once-through steam generators are reviewed. The major focus has been on improving at-temperature pH control in the secondary cycle. Various concentrations of different pH control agents were tested in the field for hundreds of days to determine their effect on steam generator fouling. High concentrations of morpholine (50--100 ppm) in the feedwater were found to apparently produce de-fouling of the steam generators without an associated decrease in feedwater iron concentration as compared to that at lower levels of morpholine. Computer modeling of the pH(t) within the OTSG for the various chemistries tested indicates that the pH can change significantly with elevation within the steam generator by varying the pH control agent or its concentration. It is postulated that these variations in pH may change the surface charge of the tubes, tube support plates, and/or corrosion product particles in solution, to favor either deposition or repulsion of the particles, and thereby producing conditions that either favor fouling or de-fouling of the OTSG. Crystal River-3 experience indicates that corrosion product deposition and release processes inside the steam generator can be chemically manipulated to favor release, and thereby maximize plant performance, and delay or avoid costly hydraulic or chemical cleanings.

  9. Corrosion Processes of the CANDU Steam Generator Materials in the Presence of Silicon Compounds

    SciTech Connect

    Lucan, Dumitra; Fulger, Manuela; Velciu, Lucian; Lucan, Georgiana; Jinescu, Gheorghita

    2006-07-01

    The feedwater that enters the steam generators (SG) under normal operating conditions is extremely pure but, however, it contains low levels (generally in the {mu}g/l concentration range) of impurities such as iron, chloride, sulphate, silicate, etc. When water is converted into steam and exits the steam generator, the non-volatile impurities are left behind. As a result of their concentration, the bulk steam generator water is considerably higher than the one in the feedwater. Nevertheless, the concentrations of corrosive impurities are in general sufficiently low so that the bulk water is not significantly aggressive towards steam generator materials. The impurities and corrosion products existing in the steam generator concentrate in the porous deposits on the steam generator tubesheet. The chemical reactions that take place between the components of concentrated solutions generate an aggressive environment. The presence of this environment and of the tubesheet crevices lead to localized corrosion and thus the same tubes cannot ensure the heat transfer between the fluids of the primary and secondary circuits. Thus, it becomes necessary the understanding of the corrosion process that develops into SG secondary side. The purpose of this paper is the assessment of corrosion behavior of the tubes materials (Incoloy-800) at the normal secondary circuit parameters (temperature = 2600 deg C, pressure = 5.1 MPa). The testing environment was demineralized water containing silicon compounds, at a pH=9.5 regulated with morpholine and cyclohexyl-amine (all volatile treatment - AVT). The paper presents the results of metallographic examinations as well as the results of electrochemical measurements. (authors)

  10. Generator of steam plasma for gasification of solid fuels

    NASA Astrophysics Data System (ADS)

    An'shakov, A. S.; Urbakh, E. K.; Rad'ko, S. I.; Urbakh, A. E.; Faleev, V. A.

    2013-12-01

    A structural design of an electric-arc steam plasma torch (plasmatron) with copper tubular electrodes has been proposed and implemented. Operational parameters are determined for the stable generation of steam plasma. Experimental data are presented on the energy characteristics of the plasma generator with the capacity up to 100 kW.