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Sample records for candu steam generator

  1. CANDU steam generator life management: laboratory data and plant experience

    Energy Technology Data Exchange (ETDEWEB)

    Tapping, R.L. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Nickerson, J.H.; Subash, N. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Wright, M.D

    2001-10-01

    As CANDU reactors enter middle age, and the potential value of the plants in a deregulated market is realized, life management and life extension issues become increasingly important. An accurate assessment of critical components, such as the CANDU 6 steam generators (SGs), is crucial for successful life extension, and in this context, material issues are a key factor. For example, service experience with Alloy 900 tubing indicates very low levels of degradation within CANDU SGs; the same is also noted worldwide. With little field data for extrapolation, life management and life extension decisions for the tube bundles rely heavily on laboratory data. Similarly, other components of the SGs, in particular the secondary side internals, have only limited inspection data upon which to base a condition assessment. However, in this case there are also relatively little laboratory data. Decisions on life management and life extension are further complicated--not only is inspection access often restricted, but repair or replacement options for internal components are, by definition, also limited. The application of CANDU SG life management and life extension requires a judicious blend of in-service data, laboratory research and development (R and D) and materials and engineering judgment. For instance, the available laboratory corrosion and fretting wear data for Alloy 800 SG tubing have been compared with plant experience (with all types of tubing), and with crevice chemistry simulations, in order to provide an appropriate inspection guide for a 50-year SG life. A similar approach has been taken with other SG components, where the emphasis has been on known degradation mechanisms worldwide. This paper provides an outline of the CANDU SG life management program, including the results to date, a summary of the supporting R and D program showing the integration with condition assessment and life management activities, and the approach taken to life extension for a typical

  2. Characteristics of U-tube assembly design for CANDU 6 type steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jun Su; Jeong, Seung Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    Since the first operation of nuclear steam generator early 1960s, its performance requirements have been met but the steam generator problems have been met but the steam generator problems have been major cause of reducing the operational reliability, plant safety and availability. U-tube assembly of steam generator forms the primary system pressure boundary of the plant and have experienced several types of tube degradation problems. Tube failure and leakage resulting from the degradation will cause radioactive contamination of secondary system by the primary coolant, and this may lead to unplanned plant outages and costly repair operations such as tube plugging or steam generator replacement. For the case of steam generators for heavy water reactors, e.g. Wolsong 2, 3, and 4 NPP, a high cost of heavy water will be imposed additionally. During the plant operation, steam generator tubes can potentially be subject to adverse environmental conditions which will cause damages to U-tube assembly. Types of the damage depend upon the combined effects of design factors, materials and chemical environment of steam generator, and they are the pure water stress corrosion cracking, intergranular attack, pitting, wastage, denting, fretting and fatigue, etc. In this report, a comprehensive review of major design factors of recirculating steam generators has been performed against the potential tube damages. Then the design characteristics of CANDU-type Wolsong steam generator were investigated in detail, including tube material, thermalhydraulic aspects, tube-to-tubesheet joint, tube supports, water chemistry and sludge management. 9 tabs., 18 figs., 38 refs. (Author) .new.

  3. Steam generator tube failures

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Beckmann, G.; Gilli, P.V.; Fritz, K.; Lippitsch, J.

    1975-12-02

    A steam generator is disclosed which is particularly adapted to be used in nuclear power plants. A casing is provided with an inlet and outlet to receive and discharge a primary heating fluid from which heat is to be extracted. A pair of tube plates extend across the interior of the casing at the region of the inlet and outlet thereof, and a plurality of tubes extend along the interior of the casing and are connected in parallel between the tube plates with all of the tubes having open ends communicating with the inlet and outlet of the casing so that the primary heating fluid will flow through the interior of the tubes while a fluid in the casing at the exterior of the tubes will extract heat from the primary fluid. The casing has between the tubes at the region of the inlet a superheating chamber and at the region of the outlet a preheating chamber and between the latter chambers an evaporating chamber, the casing receiving water through an inlet at the preheating chamber and discharging superheated steam through an outlet at the superheating chamber. A separator communicates with the evaporating chamber to receive a mixture of steam and water therefrom for separating the steam from the water and for delivering the separated steam to the superheating chamber.

  5. Steam Drum Design for Direct Steam Generation

    OpenAIRE

    Willwerth, Lisa; Müller, Svenja; Krüger, Joachim; Succo, Manuel; Feldhoff, Jan Fabian; Tiedemann, Jörg; Pandian, Juvaraj; Krüger, Dirk; Hennecke, Klaus

    2016-01-01

    For the direct steam generation in solar fields, the recirculation concept has been demonstrated in several installations. Water masses in the solar field vary during transient phases, such as passing clouds. The volume of the steam drum can serve as a buffer during such transients by taking in excess water and providing water storage. The saturated steam mass flow to the superheating section or the consumer can be maintained almost constant during short transients; therefore the steam drum p...

  6. STEAM GENERATOR GROUP PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Safety design of next generation SUI of CANDU stations

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe St. N., Oshawa, L1H 7K4 ON (Canada); Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe St. N., Oshawa, L1H 7K4 ON (Canada)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Review of current SUI technologies and challenges. Black-Right-Pointing-Pointer Propose a new type of SUI detectors. Black-Right-Pointing-Pointer Propose a new SUI system architecture and layout. Black-Right-Pointing-Pointer Propose implementation procedure for SUI with reduced risks. - Abstract: Due to the age and operating experience of Nuclear Power Plants, equipment ageing and obsolescence has become one of the main challenges that need to be resolved for all systems, structures and components in order to ensure a safe and reliable production of energy. This paper summarizes the research into a methodology for modernization of Start-Up Instrumentation (SUI), both in-core and Control Room equipment, using a new generation of detectors and cables in order to manage obsolescence. The main objective of this research is to develop a new systematic approach to SUI installation/replacement procedure development and optimization. Although some additional features, such as real-time data monitoring and storage/archiving solutions for SUI systems are also examined to take full advantage of today's digital technology, the objectives of this study do not include detailed parametrical studies of detector or system performance. Instead, a number of technological, operational and maintenance issues associated with Start-Up Instrumentation systems at Nuclear Power Plants (NPPs) will be identified and a structured approach for developing a replacement/installation procedure that can be standardized and used across all of the domestic CANDU (Canadian Deuterium Uranium) stations is proposed.

  8. Proceedings of the fourth international conference on CANDU maintenance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    These proceedings record the information presented at the 4th International Conference on CANDU Maintenance held November 16-18,1997 in Toronto, Canada. The papers for these proceedings were prepared on component maintenance, human performance, steam generator leak detection, fuel channel inspections, rotating equipment maintenance, surveillance programs, inspection techniques, valve maintenance, steam generator repairs and performance, reactor aging management and preventative maintenance.

  9. Solar steam generation: Steam by thermal concentration

    Science.gov (United States)

    Shang, Wen; Deng, Tao

    2016-09-01

    The solar-driven generation of water steam at 100 °C under one sun normally requires the use of optical concentrators to provide the necessary energy flux. Now, thermal concentration is used to raise the vapour temperature to 100 °C without the need for costly optical concentrators.

  10. Steam generator tube integrity program

    Energy Technology Data Exchange (ETDEWEB)

    Dierks, D.R.; Shack, W.J. [Argonne National Laboratory, IL (United States); 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.

  11. Advanced technologies on steam generators

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  12. Parametric study for horizontal steam generator modelling

    Energy Technology Data Exchange (ETDEWEB)

    Ovtcharova, I. [Energoproekt, Sofia (Bulgaria)

    1995-12-31

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

  13. Probabilistic integrity assessment of CANDU pressure tube for the consideration of flaw generation time

    Energy Technology Data Exchange (ETDEWEB)

    Kwak, Sang Log; Kim, Young Jin [Sungkyunkwan Univ., Seoul (Korea, Republic of); Lee, Joon Seong [Kyonggi Univ., Seoul (Korea, Republic of); Park, Youn Won [KINS, Taejon (Korea, Republic of)

    2001-07-01

    This paper describes a Probabilistic Fracture Mechanics (PFM) analysis based on Monte Carlo (MC) simulation. In the analysis of CANDU pressure tube, it is necessary to perform the PFM analyses based on statistical consideration of flaw generation time. A depth and an aspect ratio of initial semi-elliptical surface crack, a fracture toughness value, Delayed Hydride Cracking (DHC) velocity, and flaw generation time are assumed to be probabilistic variables. In all the analyses, degradation of fracture toughness due to neutron irradiation is considered. Also, the failure criteria considered are plastic collapse, unstable fracture and crack penetration. For the crack growth by DHC, the failure probability was evaluated in due consideration of flaw generation time.

  14. RPV steam generator pressure boundary

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Steam generator tubing NDE performance

    Energy Technology Data Exchange (ETDEWEB)

    Henry, G. [Electric Power Research Institute, Charlotte, NC (United States); Welty, C.S. Jr. [Electric Power Research Institute, Palo Alto, CA (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Obrutsky, L.S.; Cecco, V.S.; Sullivan, S.P. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    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.

  17. Third international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

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

  18. Strategic maintenance plan for Cernavoda steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Cicerone, T. [CNE-PROD, Cernavoda (Romania); Dhar, D.; VandenBerg, J.P. [Babcock and Wilcox (Canada)

    2002-07-01

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

  19. Hard sludge removal in steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Santibanez, M.; Stoss, J.

    2013-07-01

    One of the majors problems during the life of Nuclear power plants is the efficiency lost in steam generator due to, among issues, the plugging and therefore useless, of tubes which presented possibility of cracking in the future. The hard sludge produced in the steam generators secondary side and deposited on the tube sheet or around the tubes as collar shape are one of the main agent causing this problem, so their elimination is considered a major topic in order to keep the steam generators in an optimum condition along the whole plant life. AREVA is aware of this global problem, therefore a process and tools have been continuously developed since 1995 in order to eliminate the hard deposits in a effective way, with no damage to steam generator's components and adaptable for the different steam generators models existing in the market.

  20. SARAPAN—A Simulated-Annealing-Based Tool to Generate Random Patterned-Channel-Age in CANDU Fuel Management Analyses

    Directory of Open Access Journals (Sweden)

    Doddy Kastanya

    2017-02-01

    Full Text Available In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the *SIMULATE module of the Reactor Fueling Simulation Program (RFSP code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the *INSTANTAN module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the *INSTANTAN module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

  1. Reliability of steam generator tubing

    Energy Technology Data Exchange (ETDEWEB)

    Kadokami, E. [Mitsubishi Heavy Industries Ltd., Hyogo-ku (Japan)

    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.

  2. Failure analysis of retired steam generator tubings

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Barton, R.A. [Ontario Hydro Technologies, Toronto, ON (Canada); Moran, T.E. [Framatome Technologies Inc., Lynchburg, VA (United States); Renaud, E. [Babcock and Wilcox Industries Ltd., Cambridge, ON (Canada)

    1997-07-01

    Degradation of steam generator (SG) tubing by localized corrosion is a widespread problem in the nuclear industry that can lead to costly forced out-ages, unit de-rating, SG replacement or even the permanent shutdown of a reactor. In response to the onset of SG tubing degradation at Ontario Hydro's Pickering Nuclear Generating Station (PNGS) Unit 5, and the determined unsuitability of conventional repair methods (mechanically expanded or welded sleeves) for Alloy 400, an alternative repair technology was developed. Electrosleeve is a non-intrusive, low-temperature process that involves the electrodeposition of a nanocrystalline nickel microalloy forming a continuously bonded, structural layer over the internal diameter of the degraded region. This technology is designed to provide a long-term pressure boundary repair, fully restoring the structural integrity of the damaged region to its original state. This paper describes the Electrosleeve process for SG tubing repair and the unique properties of the advanced sleeve material. The successful installation of Electrosleeves that have been in service for more than three years in Alloy 400 SG tubing at the Pickering-5 CANDU unit, the more recent extension of the technology to Alloy 600 and its demonstration in a U.S. pressurized water reactor (PWR), is presented. A number of PWR operators have requested plant operating technical specification changes to permit Electrosleeve SG tube repair. Licensing of the Electrosleeve by the U.S. Nuclear Regulatory Commission (NRC) is expected imminently. (author)

  4. Field measurements of beta ray energy spectra in CANDU nuclear generating stations

    Energy Technology Data Exchange (ETDEWEB)

    Horowitz, Y.S. (Ben-Gurion Univ. of the Negev, Beersheba (Israel). Dept. of Physics); Hirning, C.R. (Ontario Hydro, Whitby, ON (Canada)); Yuen, P.S.; Aikens, M.S. (AECL Research, Chalk River, ON (Canada). Chalk River Labs.)

    1994-01-01

    Field measurements of beta ray energy spectra have been carried out at various locations in CANDU nuclear generating stations operated by Ontario Hydro. The beta ray energy spectrometer consists of a 5 cm diameter x 2 cm thick BC-404 plastic scintillator situated behind a 100 [mu]m thick, totally depleted, silicon detector. Photon events are rejected by requiring a coincidence between the two detectors. The spectrometer is capable of measuring electron energies from 125 keV to 3.5 MeV. Beta ray energy spectra have been measured for uncontaminated and contaminated fueling machine components, fueling machine swipes and a reactor containment vault. The degree of protection afforded by various articles of protective clothing has also been investigated for the various fueling machine components. Monte Carlo calculations have been used to estimate beta factors for 100 mg.cm[sup -2] and 240 mg.cm[sup -2] LiF-TLD chips, which are used as 'skin-and 'extremity' dosemeters in the Ontario Hydro Radiation Dosimetry Programme. (Author).

  5. US PWR steam generator management: An overview

    Energy Technology Data Exchange (ETDEWEB)

    Welty, C.S. Jr. [Electric Power Research Institute, Palo Alto, CA (United States)

    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.

  6. Basic research and industrialization of CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  7. Fourth international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  8. Steam generator tubesheet waterlancing at Bruce B

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  9. Active acoustic leak detection for LMFBR steam generators. Pt. 6. Applicability to practical steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Kazuo; Kumagai, Hiromichi; Kinoshita, Izumi [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

    1998-03-01

    It is necessary to develop a reliable water leak detection system for steam generators of liquid metal reactors in order to prevent the expansion of damage and to maintain the structural integrity of the steam generators. The concept of the active acoustic method is to detect the change of the ultrasonic field due to the hydrogen gas bubbles generated by a sodium-water reaction. This method has the potential for improved detection performance compared with conventional passive methods, from the viewpoint of sensitivity, response time and tolerance against the background noise. A feasibility study of the active acoustic leak detection system is being carried out. This report predicts the performance of the active acoustic method in the practical steam generators from the results of the large scale in-water experiments. The results shows that the active acoustic system can detect a 10 g/s leak within a few seconds in large-scale steam generators. (author)

  10. Heat Recovery Steam Generator by Using Cogeneration

    Directory of Open Access Journals (Sweden)

    P.Vivek, P. Vijaya kumar

    2014-01-01

    Full Text Available A heat recovery steam generator or HRSG is an energy recovery heat exchanger that recovers heat from a hot gas stream. It produces steam that can be used in a process (cogeneration or used to drive a steam turbine (combined cycle. It has been working with open and closed cycle. Both of cycles are used to increase the performance and also power on the cogeneration plant. If we are using closed cycle technology, we can recycle the waste heat from the turbine. in cogeneration plant, mostly they are using open cycle technology. additional, by using closed cycle technology, we can use the waste heat that converts into useful amount of work. In this paper, the exhaust gas will be sent by using proper outlet from cogen unit, we are using only waste heat that produce from turbine.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-10-01

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

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

    Science.gov (United States)

    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.

  13. Solar steam generation by heat localization.

    Science.gov (United States)

    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.

  14. Modelling the horizontal steam generator with APROS

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  15. Impact of aging and material structure on CANDU plant performance

    Energy Technology Data Exchange (ETDEWEB)

    Nadeau, E.; Ballyk, J.; Ghalavand, N. [Candu Energy Inc., Mississauga, Ontario (Canada)

    2011-07-01

    In-service behaviour of pressure tubes is a key factor in the assessment of safety margins during plant operation. Pressure tube deformation (diametral expansion) affects fuel bundle dry out characteristics resulting in reduced margin to trip for some events. Pressure tube aging mechanisms also erode design margins on fuel channels or interfacing reactor components. The degradation mechanisms of interest are primarily deformation, loss of fracture resistance and hydrogen ingress. CANDU (CANada Deuterium Uranium, a registered trademark of the Atomic Energy of Canada Limited used under exclusive licence by Candu Energy Inc.) owners and operators need to maximize plant capacity factor and meet or exceed the reactor design life targets while maintaining safety margins. The degradation of pressure tube material and geometry are characterized through a program of inspection, material surveillance and assessment and need to be managed to optimize plant performance. Candu is improving pressure tubes installed in new build and life extension projects. Improvements include changes designed to reduce or mitigate the impact of pressure tube elongation and diametral expansion rates, improvement of pressure tube fracture properties, and reduction of the implications of hydrogen ingress. In addition, Candu provides an extensive array of engineering services designed to assess the condition of pressure tubes and address the impact of pressure tube degradation on safety margins and plant performance. These services include periodic and in-service inspection and material surveillance of pressure tubes and deterministic and probabilistic assessment of pressure tube fitness for service to applicable standards. Activities designed to mitigate the impact of pressure tube deformation on safety margins include steam generator cleaning, which improves trip margins, and trip design assessment to optimize reactor trip set points restoring safety and operating margins. This paper provides an

  16. Data analysis for steam generator tubing samples

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. Corrosion Evaluation and Corrosion Control of Steam Generators

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-06-15

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

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

    Science.gov (United States)

    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

    Science.gov (United States)

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

    2016-03-01

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

  20. Advanced Eddy current NDE steam generator tubing.

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Repair technology for steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

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

  2. Thermal hydraulic studies in steam generator test facility

    Energy Technology Data Exchange (ETDEWEB)

    Vinod, V.; Suresh Kumar, V.A.; Noushad, I.B.; Ellappan, T.R.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G. [Engineering Development Group Indira Gandhi Centre for Atomic Research, Kalpakkam-603102 (India)

    2005-07-01

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

  3. Validation of the THIRST steam generator thermalhydraulic code against the CLOTAIRE phase II experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Pietralik, J.M.; Campagna, A.O.; Frisina, V.C

    1999-04-01

    Steam generator thermalhydraulic codes are frequently used to calculate both global and local parameters inside a stern generator. The global parameters include heat transfer output, recirculation ratio, outlet temperatures, and pressure drops for operating and abnormal conditions. The local parameters are used in further analyses of flow-induced vibration, fretting wear, sludge deposition, and flow-accelerated corrosion. For these purposes, detailed, 3-dimensional 2-phase flow and heat transfer parameters are needed. To make the predictions more accurate and reliable, the codes need to be validated in geometries representative of real conditions. One such study is an international co-operative experimental program called CLOTAIRE, which is based in France. The CANDU Owners Group(COG) participated in the first two phases of the program. The results of the validation of Phase 1 were presented at the 1994 Steam Generator and Heat Exchanger Conference, and the results of the validation of Phase II are the subject of this report. THIRST is a thermalhydraulic, finite-volume code used to predict flow and heat transfer in steam generators. The local results of CLOTAIRE Phase II were used to validate the code. The results consist of the measurements of void fraction and axial gas-phase velocity in the U-bend region. The measurements were done using bi-optical probes. A comparison of global results indicates that the THIRST predictions, with the Chisholm void fraction model, are within 2% to 3% of the experimental results. Using THIRST with the homogeneous void fraction model, the global results were less accurate but still gave very good predictions; the greatest error was 10% for the separator pressure drop. Comparisons of the local predictions for void fraction and axial gas-phase velocity show good agreement. The Chisholm void fraction model generally gives better agreement with the experimental data, whereas the homogeneous model tends to overpredict the void fraction

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

    Energy Technology Data Exchange (ETDEWEB)

    Cepcek, S. [Nuclear Regulatory Authority of the Slovak Republic, Trnava (Slovakia)

    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.

  5. Steam generator corrosion 2007; Dampferzeugerkorrosion 2007

    Energy Technology Data Exchange (ETDEWEB)

    Born, M. (ed.)

    2007-07-01

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

  6. Study of Scaling Development on Tube Surfaces of Water Steam Loop in Steam Generator of CEFR

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Lu; LIU; Fu-chen; LUO; De-kang; WU; Qiang; ZHANG; Huan-qi

    2012-01-01

    <正>The steam generator worked as pressure boundary of Na-H2O loop in China Experimental FastReactor (CEFR), which was quite important for nuclear reactor safety. Once the tubes separating the water from steam leak because of corrosion by scaling, Na-H2O reaction would lead to severe accident. So it’s critically important to study how the scaling develops on the water-steam sides.

  7. Steam generator issues in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Strosnider, J.R. [NRC, Washington, DC (United States)

    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.

  8. Steam generator tube inspection in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Fukui, Shigetaka [Japan Power Engineering and Inspection Corp., Tokyo (Japan)

    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.

  9. Modelling of a Coil Steam Generator for CSP applications

    DEFF Research Database (Denmark)

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

    2014-01-01

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

  10. Study on Technology Solutions of CEFR Steam Generator

    Institute of Scientific and Technical Information of China (English)

    WU; Zhi-guang; YU; Hua-jin; LIAO; Zi-yu; ZHANG; Zhen-xing

    2012-01-01

    <正>The technology solutions of CFR1000 steam generator were researched which were compared and analyze with foreign fast reactor steam generator technology solutions. The comparative analysis included the integral/modular structure, the number of modules per loop, structure types, the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  12. LMFBR steam generator systems development program progress report

    Energy Technology Data Exchange (ETDEWEB)

    None

    The intent of this program is to investigate methods of producing 2-1/4 Cr-1Mo duplex tubing to meet the structural, thermal/hydraulic and leak detection design requirements of the duplex tube leak detection concept for application on the Demonstration Plant and/or prototype steam generator. The leak detection concept as envisioned for LMFBR steam generator application will be analyzed regarding response to credible leak situations. The results of testing will be used for this analysis. The third fluid system will be conceptually designed including the two plena design adaptations being considered and the advantages and disadvantages of each will be assessed. The test program for the single-tube steam generator model will be developed in accordance with the technical and schedular objectives of the LMFBR duplex tube steam generator development program. A conceptual steam generator configuration will be established for use as a reference in the on-going feasibility studies and Demo Plant system development.

  13. Steam generated conversion coating on aluminium alloys

    DEFF Research Database (Denmark)

    Din, Rameez Ud; Jellesen, Morten Stendahl; Ambat, Rajan

    pressure steam produced by an autoclave at a temperature of 107 – 121 °C and pressure of 15 -17 psi for 10 minutes to produce a thin coating of aluminium oxide. The aim of this study is to understand the effect of high pressure steam with and without different chemical additives on surface morphology...

  14. Steam generator tube integrity flaw acceptance criteria

    Energy Technology Data Exchange (ETDEWEB)

    Cochet, B. [FRAMATOME, Paris la Defense (France)

    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.

  15. Temperature Fluctuation Characteristics Analysis for Steam Generator of Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    ZHU; Li-na; WU; Zhi-guang

    2015-01-01

    In the case of boiling heat transfer deterioration,temperature fluctuating may accelerate the corrosion of heat transfer tubes and can also lead to thermal stress on the tubes.In this paper,dryout-induced temperature fluctuation for the fast reactor steam generator is investigated.The impacts of water flow rate,sodium inlet temperature and the outlet steam

  16. Backup and Ultimate Heat Sinks in CANDU Reactors For Prolonged SBO Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Brown, M. J. [Atomic Energy of Canada Limited, Ontario (Canada)

    2013-10-15

    In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ∼2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

  17. Modelling of Steam Generating Paraboloidal dish Solar Thermal Power System

    Energy Technology Data Exchange (ETDEWEB)

    Siangsukone, P.; Lovegrove, K.

    2006-07-01

    The Australian National University (ANU) has a 400m2 Paraboloidal dish solar concentrator system, informally named the Big Dish that produces superheated steam via a receiver mounted monotube boiler connected to 50kWe steam engine for electricity generation. This paper describes an investigation of the system and its components modelled using the TRNYSYS transient system simulation package. The system was modelled in the context of performance assessment for multiple dishes, central generation Rankine cycle power plants. Five new custom components; paraboloidal dish collector, steam cavity receiver, steam line or feedwater line, steam engine, and pressure drop calculator, were developed for the TRNSYS deck file constructed for this study. Validation tests were performed by comparing with the latest experimental results measured with a 1-minute time step and good agreement, with errors less than 10%, has been found. (Author)

  18. Combined gas/steam turbine power plants with coal fired steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Krueger, H.J.; Weirich, P.H. [ABB Kraftwerke AG, Mannheim (Germany)

    1994-12-31

    The combination of coal fired steam power plants with natural gas fired gas turbines results in an essential efficiency increase, up to 50%, requiring a portion of around one third of the fuel heat input in form of natural gas. There are two basic types of circuit arrangements in this category: in a topping process the gas turbine is connected to the steam generator on the gas side, and in a compound cycle power plant gas turbine and steam circuit are connected to each other on the water/steam side via a heat recovery steam generator. If comparable design parameters are applied slightly higher plant efficiencies can be obtained with the topping process. With respect to a higher power plant availability it is possible to operate both types of circuit arrangement without gas turbine. The specific investment cost of such combined cycle power plants is lower than that of corresponding steam power plants. Hence, they can represent economical solutions as far as the price ratio between natural gas and coal is not extremely high. In ecological respects, the advantage of this combination is a reduction of the specific CO{sub 2} emission by around 20-25%, compared with pure steam power plants. 1 ref., 9 figs., 2 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  20. Estimating probable flaw distributions in PWR steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Gorman, J.A.; Turner, A.P.L. [Dominion Engineering, Inc., McLean, VA (United States)

    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.

  1. MINET validation study using steam generator test data

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Automated Diagnosis and Classification of Steam Generator Tube Defects

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Development of data management system for steam generator inspection

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yong Moo; Im, Chang Jae; Lee, Yoon Sang; Kang, Soon Joo; An, Jong Kwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    The data communications environment for transferring Nuclear Power Plant Steam Generator Eddy Current testing data was investigated and after connecting LAN to Hinet-F network, the remote data transfer with the speed of 56 kbps was tested successfully. Data management system for Steam Generator Eddy current testing was also developed by using HP-UX, RMB (Rock Mountain Basic) 21 figs, 13 tabs, 5 refs. (Author).

  4. Measurements of beta ray spectra in CANDU nuclear generating stations using a silicon detector coincidence telescope

    Energy Technology Data Exchange (ETDEWEB)

    Horowitz, Y.S.; Weizman, Y. [Ben-Gurion Univ. of the Negev, Beersheba (Israel). Dept. of Physics; Hirning, C.R. [Ontario Hydro, Whitby, ON (Canada). Health Physics Dept.

    1996-12-31

    The measurement of beta ray spectra at various work locations inside nuclear generating stations operated by Ontario Hydro is described. The measurements were carried out using an advanced coincidence telescope spectrometer using silicon detectors only. The spectrometer is capable of measuring electron energies over the range 60 keV- 2500 keV with close to 100% coincidence efficiency. Photon rejection is carried out by requiring a coincidence between either two or three silicon detectors. Monte Carlo calculations were then used to estimate beta correction factors for the LiF:Mg,Ti elements used in the Ontario Hydro thermoluminescence dosemeters. Averaging over all the measured beta correction factors for the `skin` chip (100 mg.cm{sup -2}) results in a value of 2.73 {+-} 0.77 and for the extremity dosemeter (240 mg.cm{sup -2}) an average value of 4.42 {+-} 1.17 is obtained. These values are 57% and 120% greater, respectively, than the current values used by Ontario Hydro. In addition, beta correction factors for nine representative spectra were calculated for 40 mg.cm{sup -2} chips and 20 mg.cm{sup -2} chips and the results demonstrate the benefits of decreased dosemeter thickness. The average value of the beta correction factor, as well as the spread in the beta correction factor, decreases dramatically from 4.8 {+-} 2.1 (240 mg.cm{sup -2}) to 1.29 ``1.2`` +-`` 0.1 (20 mg.cm{sup -2}). (author).

  5. Nonlinear H-infinity control of nuclear steam generators

    Science.gov (United States)

    Ramalho, Fernando Pinto

    Motivated by the fact that problems related to the control of steam generators are responsible for a significant amount of downtime in nuclear power plants, this thesis investigates the applicability of linear and nonlinear Hinfinity theory to the control of nuclear steam generators. A nonlinear model based on mass, energy, and momentum balances was developed for a U-tube steam generator, with the water level and steam quality at the exit of the riser considered as state variables. In this model the steam flow to the turbines and the heat flow from the primary to the secondary side are represented as disturbances affecting the system, while the feedwater flow is used to compensate for changes in the water level. The performance specifications for the feedback loop are encoded using weight functions incorporated into an augmented plant, and the control problem is formulated to minimize the effects of disturbances on the controlled variables. The solution of the optimization problem is reduced to the solution of a set of differential equations, which, in the linear case, is equivalent to the solution of Riccati equations. The linear Hinfinity controller and filter were obtained for the U-tube steam generator with and without weight functions, and simulations for a 50 s ramp transient resulting in 50% decrease in the heat and steam flows were performed over 300 s. The use of weights provided less variation in the water level, and an excellent noise rejection capability was observed. For the nonlinear Hinfinity formulation a finite-difference method was used to solve the state and costate equations numerically for optimal feedwater flow minimizing water level variations. The combined solution of the state equation in the forward direction and the costate equations in the backward direction converged in 10 iteractions. The nonlinear controller results in less variation in the water level than the corresponding linear Hinfinity controller, demonstrating the feasibility

  6. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

    1995-10-01

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  7. Severe accident analysis of a station blackout accident using MAAP-CANDU for the Point Lepreau station refurbishment project level 2 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Petoukhov, S.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station, using the MAAP-CANDU code to simulate the progression of severe core damage accidents and fission product releases. Five representative severe accidents were selected: Station Blackout, Small Loss-of-Coolant, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State. Analysis results for the reference station blackout accident are discussed in this paper. (author)

  8. Localization of CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Alizadeh, Ala

    2010-09-15

    The CANDU pressurized heavy water reactor's principal design features suit it particularly well for technology transfer and localization. When the first commercial CANDU reactors of 540 MWe entered service in 1971, Canada's population of less than 24 million supported a 'medium' level of industrial development, lacking the heavy industrial capabilities of larger countries like the USA, Japan and Europe. A key motivation for Canada in developing the CANDU design was to ensure that Canada would have the autonomous capacity to build and operate nuclear power reactors without depending on foreign sources for key components or enriched fuel.

  9. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States). Dept. of Mechanical Engineering

    1995-12-31

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines; however there is practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

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

    Energy Technology Data Exchange (ETDEWEB)

    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. Hydrogen-based power generation from bioethanol steam reforming

    Science.gov (United States)

    Tasnadi-Asztalos, Zs.; Cormos, C. C.; Agachi, P. S.

    2015-12-01

    This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO2 emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint.

  12. Hydrogen-based power generation from bioethanol steam reforming

    Energy Technology Data Exchange (ETDEWEB)

    Tasnadi-Asztalos, Zs., E-mail: tazsolt@chem.ubbcluj.ro; Cormos, C. C., E-mail: cormos@chem.ubbcluj.ro; Agachi, P. S. [Babes-Bolyai University, Faculty of Chemistry and Chemical Engineering, 11 Arany Janos, Postal code: 400028, Cluj-Napoca (Romania)

    2015-12-23

    This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO{sub 2} emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint.

  13. Experience on management of CANDU spent fuel in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H.-Y.; Choi, B.-I.; Yoon, J.-H.; Seo, U.-S. [Korea Hydro and Nuclear Power Co. Ltd., Nuclear Environment Technology Inst. (KHNP/NETEC), Yusung-Gu, Daejeon (Korea, Republic of)

    2002-07-01

    In Korea, national policy on the management of spent fuel from both PWR and CANDU reactors demands that all the spent fuel be kept within reactor site in until 2016 the time spent fuel interim storage facility might open. Based on the end of 2001, KHNP has 4 CANDU reactors in operation generating approximately 5,000 bundles of spent fuels per each unit annually. The generation, accumulation, and management of CANDU spent fuel by KHNP in Korea are reviewed. CANDU spent fuel storage technology including pool storage in fuel building, concrete silo storage, and on going project for consolidating storage adapting modular vault type MACSTOR concept are outlined. Especially current joint development of storage of CANDU spent fuel for improving land usage is addressed. The explanation of the new consolidated dry storage system includes description of the storage facility, its safety evaluations, and final implementation. Finally future movement on management of spent fuel in Korea is also briefly introduced. (author)

  14. Research of laser cleaning technology for steam generator tubing

    Science.gov (United States)

    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.

  15. Subcooled choked flow through steam generator tube cracks

    Science.gov (United States)

    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.

  16. PMK-2. Experimental study on steam generator behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Ezsoel, G.; Szabados, L.; Trosztel, I. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1995-12-31

    The PMK-2 is a full pressure scaled-down model of the Paks Nuclear Power Plant, with a 1:2070 scaling ratio for the volume and power. It has a steam generator model which is a vertical section of the horizontal steam generator. The model has hot and cold collectors similarly to the steam generators of the plant. The heat transfer tubes are horizontal tubes. There are 82 rows of tubes and the elevations, as well as the heat transfer surface distribution is the same as in the plant. The elevation of the feed water supply is similar to that of the plant. To study the temperature distribution in both the primary and the secondary side several thermocouples are built in, in addition to the overall instrumentation of the loop which has again a high number of measurement channels. Paper gives a description and results of SPE-4, with special respect to the steam generator behaviour in both steady state and transient conditions. Axial distribution of coolant and feedwater temperatures are given for the primary and the secondary side of hot and cold collectors and the temperature distribution in the centre of steam generator. (orig.).

  17. Future CANDU nuclear power plant design requirements document executive summary

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Usmani, S.A. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    The future CANDU Requirements Document (FCRED) describes a clear and complete statement of utility requirements for the next generation of CANDU nuclear power plants including those in Korea. The requirements are based on proven technology of PHWR experience and are intended to be consistent with those specified in the current international requirement documents. Furthermore, these integrated set of design requirements, incorporate utility input to the extent currently available and assure a simple, robust and more forgiving design that enhances the performance and safety. The FCRED addresses the entire plant, including the nuclear steam supply system and the balance of the plant, up to the interface with the utility grid at the distribution side of the circuit breakers which connect the switchyard to the transmission lines. Requirements for processing of low level radioactive waste at the plant site and spent fuel storage requirements are included in the FCRED. Off-site waste disposal is beyond the scope of the FCRED. 2 tabs., 1 fig. (Author) .new.

  18. Physical and statistical models for steam generator clogging diagnosis

    CERN Document Server

    Girard, Sylvain

    2014-01-01

    Clogging of steam generators in nuclear power plants is a highly sensitive issue in terms of performance and safety and this book proposes a completely novel methodology for diagnosing this phenomenon. It demonstrates real-life industrial applications of this approach to French steam generators and applies the approach to operational data gathered from French nuclear power plants. The book presents a detailed review of in situ diagnosis techniques and assesses existing methodologies for clogging diagnosis, whilst examining their limitations. It also addresses numerical modelling of the dynamic

  19. Health and safety impact of steam generator tube degradation

    Energy Technology Data Exchange (ETDEWEB)

    Marston T. [PLG, Inc., Newport Beach, CA (United States)

    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.

  20. Evaluation of steam generator WWER 440 tube integrity criteria

    Energy Technology Data Exchange (ETDEWEB)

    Splichal, K.; Otruba, J.; Burda, J. [Nuclear Research Institute Rez plc. (Czechoslovakia)

    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.

  1. Regulation, pollution and heterogeneity in Japanese steam power generation companies

    Energy Technology Data Exchange (ETDEWEB)

    Barros, Carlos Pestana [Instituto Superior de Economia e Gestao, Technical University of Lisbon Rua Miguel Lupi, Lisbon (Portugal); Managi, Shunsuke [Faculty of Business Administration, Yokohama National University, 79-4, Tokiwadai, Hodogaya-ku, Yokohama 240-8501 (Japan)

    2009-08-15

    In this paper, the random stochastic frontier model is used to estimate the technical efficiency of Japanese steam power generation companies taking into regulation and pollution. The companies are ranked according to their productivity for the period 1976-2003 and homogenous and heterogeneous variables in the cost function are disentangled. Policy implication is derived. (author)

  2. Steam generators regulatory practices and issues in Spain

    Energy Technology Data Exchange (ETDEWEB)

    Mendoza, C.; Castelao, C.; Ruiz-Colino, J.; Figueras, J.M. [CSN, Madrid (Spain)

    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. Effect of liquid waste discharges from steam generating facilities

    Energy Technology Data Exchange (ETDEWEB)

    McGuire, H.E. Jr.

    1977-09-01

    This report contains a summary of the effects of liquid waste discharges from steam electric generating facilities on the environment. Also included is a simplified model for use in approximately determining the effects of these discharges. Four basic fuels are used in steam electric power plants: three fossil fuels--coal, natural gas, and oil; and uranium--presently the basic fuel of nuclear power. Coal and uranium are expected to be the major fuels in future years. The following power plant effluents are considered: heat, chlorine, copper, total dissolved solids, suspended solids, pH, oil and grease, iron, zinc, chrome, phosphorus, and trace radionuclides.

  4. Modeling a Helical-coil Steam Generator in RELAP5-3D for the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    Nathan V. Hoffer; Piyush Sabharwall; Nolan A. Anderson

    2011-01-01

    Options for the primary heat transport loop heat exchangers for the Next Generation Nuclear Plant are currently being evaluated. A helical-coil steam generator is one heat exchanger design under consideration. Safety is an integral part of the helical-coil steam generator evaluation. Transient analysis plays a key role in evaluation of the steam generators safety. Using RELAP5-3D to model the helical-coil steam generator, a loss of pressure in the primary side of the steam generator is simulated. This report details the development of the steam generator model, the loss of pressure transient, and the response of the steam generator primary and secondary systems to the loss of primary pressure. Back ground on High Temperature Gas-cooled reactors, steam generators, the Next Generation Nuclear Plant is provided to increase the readers understanding of the material presented.

  5. SGTR Project: Separate Effect Studies for Vertical Steam Generators

    Energy Technology Data Exchange (ETDEWEB)

    Peyres, V.; Polo, J.; Herranz, L. E.

    2003-07-01

    The SGTR project has been carried out within the fifth EURATOM Framework Programme (Contract No FIKS-CT-1999-0007). Its main objective was to provide an experimental database and to develop and/or verify models to support definition of accident management measures in the hypothetical case of a Steam Generator tube Rupture (SGTR) sequence. The project addressed both vertical and horizontal steam generator designs. This report summarises the main results obtained in the intermediate scale experimentation that addressed Western type steam generators. The specific goal of this test programme was to investigate aerosol retention in the break stage of the secondary side of a water-empty steam generator. The test matrix consisted of 12 tests that explored the influence of variables such as break type and orientation and inlet gas flow rate. This work was performed in the PECA facility of the Laboratory for Analysis of Safety Systems (LASS). Aerosol retention at the break stage of a dry steam generator was observed to be low and non-uniform. Neither break type nor orientation affected results significantly whenever gas flowrates exceeded about 100 kg/h. However, deposition patterns guillotine breaks and fish mouth ones showed remarkable differences. For flowrates above 100 kg/hm the higher the gas flow velocity, the lower the total mass depleted on tube bundle surfaces; however, at lower flowrates this trend was not maintained. An attempt to measure gas injection velocity at the break exit by Particle Image Velocity (PIV) was done but data were highly uncertain. (Author) 2 refs.

  6. Ultrasonic Cleaning of Nuclear Steam Generator by Micro Bubble

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woo Tae [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of); Kim, Sang Tae; Yoon, Sang Jung [Sae-An Engineering Co., Seoul (Korea, Republic of)

    2012-05-15

    In this paper, we present ultrasonic cleaning technology for a nuclear steam generator using micro bubble. We could extend the boundary of ultrasonic cleaning by using micro bubbles in water. Ultrasonic energy measured was increased about 5 times after the generation of micro bubbles in water. Furthermore, ultrasound energy was measured to be strong enough to create cavitation even though the ultrasound sensor was about 2 meters away from the ultrasonic transducer

  7. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Science.gov (United States)

    2010-10-01

    ...)). (a) Pressure vessels in which steam is generated are classed as “Unfired Steam Boilers” except as required otherwise by paragraph (b) of this section. Unfired steam boilers must be fitted with an efficient... § 54.15-15. Unfired steam boilers must be constructed in accordance with this part other than when...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, S.; Querol, A.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino de Vera s/n, 46022, Valencia (Spain)

    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)

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

    Science.gov (United States)

    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.

  11. CANDU, building the future

    Energy Technology Data Exchange (ETDEWEB)

    Stern, F. [Stern Laboratories (Canada)

    1997-07-01

    The CEO of Stern Laboratories delivered a speech on the problems and challenges facing the nuclear industry. The CANDU system is looked at as the practical choice for the future of our energy source. The people of the industry must be utilized and respected to deliver to the best of their ability.

  12. Design of fault tolerant control system for steam generator using

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Ki; Seo, Mi Ro [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A controller and sensor fault tolerant system for a steam generator is designed with fuzzy logic. A structure of the proposed fault tolerant redundant system is composed of a supervisor and two fuzzy weighting modulators. A supervisor alternatively checks a controller and a sensor induced performances to identify which part, a controller or a sensor, is faulty. In order to analyze controller induced performance both an error and a change in error of the system output are chosen as fuzzy variables. The fuzzy logic for a sensor induced performance uses two variables : a deviation between two sensor outputs and its frequency. Fuzzy weighting modulator generates an output signal compensated for faulty input signal. Simulations show that the proposed fault tolerant control scheme for a steam generator regulates well water level by suppressing fault effect of either controllers or sensors. Therefore through duplicating sensors and controllers with the proposed fault tolerant scheme, both a reliability of a steam generator control and sensor system and that of a power plant increase even more. 2 refs., 9 figs., 1 tab. (Author)

  13. Two Phase Flow Stability in the HTR-10 Steam Generator

    Institute of Scientific and Technical Information of China (English)

    居怀明; 左开芬; 刘志勇; 徐元辉

    2001-01-01

    A 10 MW High Temperature Gas Cooled Reactor (HTR-10) designed bythe Institute of Nuclear Energy Technology (INET) is now being constructed. The steam generator (SG) in the HTR-10 is one of the most important components for reactor safety. The thermal-hydraulic performance of the SG was investigated. A full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 thermal hydraulic experimental system in detail. The SGTM-10 simulates the actual thermal and structural parameters of the HTR-10. The SGTM-10 includes three separated loops: the primary helium loop, the secondary water loop, and the tertiary cooling water loop. Two parallel tubes are arranged in the test assembly. The main experimental equipment is shown in the paper. Expermental results are given illustrating the effects of the outlet pressures, the heating power, and the inlet subcooling.

  14. Analysis of flow instabilities in forced-convection steam generator

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Because of the practical importance of two-phase instabilities, substantial efforts have been made to date to understand the physical phenomena governing such instabilities and to develop computational tools to model the dynamics. The purpose of this study is to present a numerical model for the analysis of flow-induced instabilities in forced-convection steam generator. The model is based on the assumption of homogeneous two-phase flow and thermodynamic equilibrium of the phases. The thermal capacity of the heater wall has been included in the analysis. The model is used to analyze the flow instabilities in the steam generator and to study the effects of system pressure, mass flux, inlet temperature and inlet/outlet restriction, gap size, the ratio of do /di, and the ratio of qi/qo on the system behavior.

  15. PWR steam generator chemical cleaning, Phase I. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Rothstein, S.

    1978-07-01

    United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI.

  16. Numerical investigation of mass transfer in the flow path of the experimental model of the PGV-1500 steam generator's steam receiving section with two steam nozzles

    Science.gov (United States)

    Golibrodo, L. A.; Krutikov, A. A.; Nadinskii, Yu. N.; Nikolaeva, A. V.; Skibin, A. P.; Sotskov, V. V.

    2014-10-01

    The hydrodynamics of working medium in the steam volume model implemented in the experimental setup constructed at the Leipunskii Institute for Physics and Power Engineering was simulated for verifying the procedure of calculating the velocity field in the steam space of steam generators used as part of the reactor plants constructed on the basis of water-cooled water-moderated power-generating reactors (VVER). The numerical calculation was implemented in the environment of the STAR-CCM+ software system with its cross verification in the STAR-CD and ANSYS CFX software systems. The performed numerical investigation served as a basis for substantiating the selection of the computation code and parameters for constructing the computer model of the steam receiving device of the PGV-1500 steam generator experimental model, such as the quantization scheme, turbulence model, and mesh model.

  17. Investigation of thermal storage and steam generator issues

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    A review and evaluation of steam generator and thermal storage tank designs for commercial nitrate salt technology showed that the potential exists to procure both on a competitive basis from a number of qualified vendors. The report outlines the criteria for review and the results of the review, which was intended only to assess the feasibility of each design, not to make a comparison or select the best concept.

  18. Modelling studies of horizontal steam generator PGV-1000 with Cathare

    Energy Technology Data Exchange (ETDEWEB)

    Karppinen, I. [VTT Energy, Espoo (Finland)

    1995-12-31

    To perform thermal-hydraulic studies applied to nuclear power plants equipped with VVER, a program of qualification and assessment of the CATHARE computer code is in progress at the Institute of Protection and Nuclear Safety (IPSN). In this paper studies of modelling horizontal steam generator of VVER-1000 with the CATHARE computer code are presented. Steady state results are compared with measured data from the fifth unit of Novovoronezh nuclear power plant. (orig.). 10 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Shoemaker, P.; Cislo, D. [AREVA NP Inc., Lynchburg, Virginia (United States)]. E-mail: paul.shoemaker@areva.com

    2006-07-01

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

  20. Thermal-hydraulic experiments for the PCHE type steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Shin, C. W.; No, H. C. [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Printed circuit heat exchanger (PCHE) manufactured by HEATRIC is a compact type of the mini-channel heat exchanger. The PCHE is manufactured by diffusion bonding of the chemically-etched plates, and has high heat transfer rate due to a large surface. Therefore, the size of heat exchanger can be reduced by 1/5 - 1/6 and PCHE can be operated under high pressure, high temperature and multi-phase flow. Under such merits, it is used as heat exchanger with various purposes of gas cycle and water cycle. Recently, it is newly suggested as an application of a steam generator. IRIS of MIT and FASES of KAIST conceptually adopted PCHE as a steam generator. When using boiling condition of micro-channel, flow instability is one of the critical issues. Instability may cause unstable mass flow rate, sudden temperature change and system control failure. However instability tests of micro channels using water are very limited because the previous studies were focused on a single tube or other fluid instead of water. In KAIST, we construct the test facility to study the thermal hydraulics and fluid dynamics of the heat exchanger, especially occurrence of instability. By inducing the pressure drop of inlet water, amplitude of oscillation declined by 90%. Finally, the throttling effect was experimentally confirmed that PCHE could be utilized as a steam generator.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  3. Estimation of Aging Effects on LOHS for CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Yong Ki; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    To evaluate the Wolsong Unit 1's capacity to respond to large-scale natural disaster exceeding design, the loss of heat sink(LOHS) accident accompanied by loss of all electric power is simulated as a beyond design basis accident. This analysis is considered the aging effects of plant as the consequences of LOHS accident. Various components of primary heat transport system(PHTS) get aged and some of the important aging effects of CANDU reactor are pressure tube(PT) diametral creep, steam generator(SG) U-tube fouling, increased feeder roughness, and feeder orifice degradation. These effects result in higher inlet header temperatures, reduced flows in some fuel channels, and higher void fraction in fuel channel outlets. Fresh and aged models are established for the analysis where fresh model is the circuit model simulating the conditions at retubing and aged model corresponds to the model reflecting the aged condition at 11 EFPY after retubing. CATHENA computer code[1] is used for the analysis of the system behavior under LOHS condition. The LOHS accident is analyzed for fresh and aged models using CATHENA thermal hydraulic computer code. The decay heat removal is one of the most important factors for mitigation of this accident. The major aging effect on decay heat removal is the reduction of heat transfer efficiency by steam generator. Thus, the channel failure time cannot be conservatively estimated if aged model is applied for the analysis of this accident.

  4. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  5. Process Technology Development of Ni Electroplating in Steam Generator Tube

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joung Soo; Kim, H. P.; Lim, Y. S.; Kim, S. S.; Hwang, S. S.; Yi, Y. S.; Kim, D. J.; Jeong, M. K.

    2009-11-15

    Operating nuclear power steam generator tubing material, Alloy 600, having superior resistance to corrosion has many experiences of damage by various corrosion mechanisms during long term operation period. In this research project, a new Ni electroplating technology to be applied to repair the damaged steam generator tubes has been developed. In this technology development, the optimum conditions for variables affecting the Ni electroplating process, optimum process conditions for maximum adhesion forces at interface between were established. The various mechanical properties (RT and HT tensile, fatigue, creep, burst, etc.) and corrosion properties (general corrosion, pitting, crevice corrosion, stress corrosion cracking, boric acid corrosion, doped steam) of the Ni plated layers made at the established optimum conditions have been evaluated and confirmed to satisfy the specifications. In addition, a new ECT probe developed at KAERI enable to detect defects from magnetic materials was confirmed to be used for Ni electroplated Alloy 600 tubes at the field. For the application of this developed technology to operating plants, a mock-up electroplating system has been designed and manufactured, and set up at Doosan Heavy Industry Co. and also its performance test has been done. At same time, the anode probe has been modified and improved to be used with the established mock-up system without any problem

  6. Failure Analysis in Steam Generators: A Study Case

    Directory of Open Access Journals (Sweden)

    Alberto Eduardo Calvo González

    2016-04-01

    Full Text Available The use of crude oil as alternative to fuel steam power stations was justifi ed by economic reasons. Thischange of fuel required to modify the operational procedures, due fundamentally to its lower heating valueand its high density and viscosity. Nevertheless the changes made, the use of the crude oil caused aquick deterioration of the thermal exchange surfaces causing not planned forced outages. The statisticalanalysis of the steam generator forced outages went to reheater number two. Therefore the scope of thiswork is the study of the happened failures in reheater number two, of 433,536 kg/hr, 13,4 MPa, and 525 Ctemperature of superheated and reheated steam generator used to move 125 MW turbines. In the carriedout failure analysis, based on nondestructive evaluation methods, towered with metallographic test, waspossible to clarify the root cause, as well as establish the worsening growth rate that allowed establishingthe sequence of tests to avoid a possible not planned outage. In turn it was possible to choose the kind ofsteel should be used in that reheater to give the defi nitive solution of the problem. While steel substitutionis not carried out it sho uld stay the régime of tests and proposed assays. This experience can serve frombase to the design of a maintenance system based on the condition.

  7. An experimental study of a VVER reactor's steam generator model operating in the condensing mode

    Science.gov (United States)

    Morozov, A. V.; Remizov, O. V.

    2012-05-01

    Results obtained from an experimental study of a VVER reactor's steam generator model operating in the condensing mode are presented. The obtained empirical dependence for calculating the power of heat exchangers operating in the steam condensation mode is presented.

  8. Alternate tube plugging criteria for steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Cueto-Felgueroso, C.; Aparicio, C.B. [Tecnatom, S.A., Madrid (Spain)

    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. Quadratic controller syntheses for the steam generator water level

    Energy Technology Data Exchange (ETDEWEB)

    Arzelier, D.; Daafouz, J.; Bernussou, J.; Garcia, G

    1998-06-01

    The steam generator water level, (SGWL), control problem in the pressurized water reactor of a nuclear power plant is considered from robust control techniques point of view. The plant is a time-varying system with a non minimum phase behavior and an unstable open-loop response. The time-varying nature of the plant due to change in operating power is taken into account by including slowly time-varying uncertainty in the model. A linear Time-Invariant, (LTI) guaranteed cost quadratic stabilizing controller is designed in order to address some of the particular issues arising for such a control problem. (author) 17 refs.

  10. Radiological assessment of steam generator repair and replacement

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Steam generator chemical cleaning at the Palo Verde Nuclear Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Jevec, J.M. [Babcock and Wilcox, Alliance, OH (United States). R and D Division; Knollmeyer, P.M. [B and W Nuclear Technologies, Lynchburg, VA (United States); Paramithas, P. [Palo Verde Nuclear Generating Station, Tonopah, AZ (United States)

    1995-09-01

    The secondary side of the Palo Verde Units 2 and 3 steam generators were chemically cleaned in 1994. The primary purpose of the chemical cleaning was to remove deposits bridging between adjacent tubes and also to remove bulk tube and tubesheet deposits. A secondary objective was to remove deposits from the flow distribution plate-to-tube crevice. The chemical cleaning consisted of a magnetite dissolution step, a separate step aimed at removing deposits in the flow distribution plate crevices, and a final step to remove residual copper and passivate the carbon steel surfaces of the steam generator. Corrosion monitoring was employed during the cleaning to ensure that the cleaning resulted in corrosion to steam generator materials of construction that was below the predetermined chemical cleaning corrosion allowances. The process application, removal efficiency, and corrosion results are presented in this paper.

  12. Analysis of once-through steam generator instability

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Han Ok; Kang, Hyung Suk; Cho, Bong Hyun; Yoon, Ju Hyeon [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-03-01

    KAERI is carrying out a development of the design for a new type of integral reactor named SMART (System-integrated Modular Advanced Reactor). Two models, the frequency domain-linear model and the time domain-nonlinear model, are developed for the analysis of once-through helical steam generator flow instability. The linear model is used for easy determination of critical point with constant heat flux condition. The nonlinear model is for the analysis of oscillation characteristics beyond the critical point as well as determination of the point with real primary boundary conditions. The developed linear model is utilized to evaluate the effect of several nondimensional parameters on flow stability for the wide range of input conditions. The results from the developed nonlinear model are compared with the existing experimental data including steady state values and critical conditions. The calculated lengths of each region and pressure drops in the steady show almost same trends with Nariai's experimental results. Two developed models can be utilized to analyze the steam generator flow instabilities and to design the inlet orifices are to prevent flow instabilities. (author). 118 refs., 32 figs., 1 tab.

  13. Flow distribution in the inlet plenum of steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Khadamakar, H.P. [Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400 019 (India); Patwardhan, A.W., E-mail: aw.patwardhan@ictmumbai.edu.in [Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400 019 (India); Padmakumar, G.; Vaidyanathan, G. [Experimental Thermal Hydraulics Section, Separation Technology and Hydraulics Division, Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2011-10-15

    Highlights: > Various flow distribution devices have been studied to make the flow distribution uniform in axial as well as tangential direction. > Experiments were performed using Ultrasonic Velocity Profiler (UVP) and Particle Image Velocimetry (PIV). > CFD modeling has been carried out to give more insights. > Various flow distribution devices have been compared. - Abstract: The flow distribution in a 1/5th and 1/8th scale models of inlet plenum of steam generator (SG) has been studied by a combination of experiments and Computational Fluid Dynamics (CFD) simulations. The distribution of liquid sodium in the inlet plenum of the SG strongly affects the thermal as well as mechanical performance of the steam generator. Various flow distribution devices have been used to make the flow distribution uniform in axial as well as tangential direction in the window region. Experiments have been conducted to measure the radial velocity distribution using Ultrasonic Velocity Profiler (UVP) and Particle Image Velocimetry (PIV) under a variety of conditions. CFD modeling has been carried out for various configurations to give more insight into the flow distribution phenomena. The various flow distribution devices have been compared on the basis of a non-uniformity index parameter.

  14. Recent operating experiences with steam generators in Japanese NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Yashima, Seiji [Japan Power Engineering and Inspection Corp., Tokyo (Japan)

    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.

  15. Actual operation and regulatory activities on steam generator replacement in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Saeki, Hitoshi [Kyushu Electric Power Co., Inc., Fukyoka (Japan)

    1997-02-01

    This paper summarizes the operating reactors in Japan, and the status of the steam generators in these plants. It reviews plans for replacement of existing steam generators, and then goes into more detail on the planning and regulatory steps which must be addressed in the process of accomplishing this maintenance. The paper also reviews the typical steps involved in the process of removal and replacement of steam generators.

  16. Steam generator degradation: Current mitigation strategies for controlling corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Millett, P. [Electric Power Research Institute, Palo Alto, CA (United States)

    1997-02-01

    Steam Generator degradation has caused substantial losses of power generation, resulted in large repair and maintenance costs, and contributed to significant personnel radiation exposures in Pressurized Water Reactors (PWRs) operating throughout the world. EPRI has just published the revised Steam Generator Reference Book, which reviews all of the major forms of SG degradation. This paper discusses the types of SG degradation that have been experienced with emphasis on the mitigation strategies that have been developed and implemented in the field. SG degradation is presented from a world wide perspective as all countries operating PWRs have been effected to one degree or another. The paper is written from a US. perspective where the utility industry is currently undergoing tremendous change as a result of deregulation of the electricity marketplace. Competitive pressures are causing utilities to strive to reduce Operations and Maintenance (O&M) and capital costs. SG corrosion is a major contributor to the O&M costs of PWR plants, and therefore US utilities are evaluating and implementing the most cost effective solutions to their corrosion problems. Mitigation strategies developed over the past few years reflect a trend towards plant specific solutions to SG corrosion problems. Since SG degradation is in most cases an economic problem and not a safety problem, utilities can focus their mitigation strategies on their unique financial situation. Accordingly, the focus of R&D has shifted from the development of more expensive, prescriptive solutions (e.g. reduced impurity limits) to corrosion problems to providing the utilities with a number of cost effective mitigation options (e.g. molar ratio control, boric acid treatment).

  17. Severe accident analysis of a small LOCA accident using MAAP-CANDU support level 2 PSA for the Point Lepreau station refurbishment project

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Brown, M.J.; Mathew, P.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP4-CANDU code was used to calculate the progression of postulated severe core damage accidents and fission product releases. Five representative severe core damage accidents were selected: Station Blackout, Small Loss-of-Coolant Accident, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State Accident. Analysis results for only the reference Small LOCA Accident scenario (which is a very low probability event) are discussed in this paper. (author)

  18. Computerized operating cost model for industrial steam generation

    Energy Technology Data Exchange (ETDEWEB)

    Powers, T.D.

    1983-02-01

    Pending EPA regulations, establishing revised emission levels for industrial boilers are perceived to have an effect on the relative costs of steam production technologies. To aid in the comparison of competitive boiler technologies, the Steam Cost Code was developed which provides levelized steam costs reflecting the effects of a number of key steam cost parameters. The Steam Cost Code is a user interactive FORTRAN program designed to operate on a VAX computer system. The program requires the user to input a number of variables describing the design characteristics, capital costs, and operating conditions for a specific boiler system. Part of the input to the Steam Cost Code is the capital cost of the steam production system. The capital cost is obtained from a program called INDCEPT, developed by Oak Ridge National Laboratory under Department of Energy, Morgantown Energy Technology Center sponsorship.

  19. Evaluation on mechanical and corrosion properties of steam generator tubing materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Sup; Lee, Byong Whi; Lee, Sang Kyu; Lee, Young Ho; Kim, Jun Whan; Lee, Ju Seok; Kwon, Hyuk Sang; Kim, Su Jung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-06-15

    Steam generator is one of the major components of nuclear reactor pressure boundary. It's main function os transferring heat which generated in the reactor to turbine generator through steam generator tube. In these days, steam generator tubing materials of operating plant are used Inconel 600 alloys. But according to the operation time, there are many degradation phenomena which included mechanical damage due to flow induced vibration and corrosion damage due to PWSCC, IGA/SCC and pitting etc. Recently Inconel 690 alloys are selected as new and replacement steam generator tubes for domestic nuclear power plant. But there are few study about mechanical and corrosion properties of Inconel 600 and 690. The objectives of this study is to evaluate and compare mechanical and corrosion propertied of steam generator tube materials.

  20. Modeling local chemistry in PWR steam generator crevices

    Energy Technology Data Exchange (ETDEWEB)

    Millett, P.J. [EPRI, Palo Alto, CA (United States)

    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.

  1. Direct measurements of secondary water inventory of steam generator PGV-213 in operation

    Energy Technology Data Exchange (ETDEWEB)

    Tarankov, G.A.; Trunov, N.B.; Dranchenko, B.N.; Kamiagin, W.W. [OKB Gidropress (Russian Federation)

    1997-12-31

    Results of weight measurement of PGV-213 steam generator during filling in, heating-up and power increase are described. Special measurement system based on stress gauges has been developed. Method of derivation of secondary water inventory is described. Comparison of the data for two steam generators prove accuracy of the measurements. (orig.). 1 refs.

  2. Enviro-Friendly Hydrogen Generation from Steel Mill-Scale via Metal-Steam Reforming

    Science.gov (United States)

    Azad, Abdul-Majeed; Kesavan, Sathees

    2006-01-01

    An economically viable and environmental friendly method of generating hydrogen for fuel cells is by the reaction of certain metals with steam, called metal-steam reforming (MSR). This technique does not generate any toxic by-products nor contributes to the undesirable greenhouse effect. From the standpoint of favorable thermodynamics, total…

  3. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Woo, H.H.; Lu, S.C.

    1981-09-15

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  4. Ludwig: A Training Simulator of the Safety Operation of a CANDU Reactor

    Directory of Open Access Journals (Sweden)

    Gustavo Boroni

    2011-01-01

    Full Text Available This paper presents the application Ludwig designed to train operators of a CANDU Nuclear Power Plant (NPP by means of a computer control panel that simulates the response of the evolution of the physical variables of the plant under normal transients. The model includes a close set of equations representing the principal components of a CANDU NPP plant, a nodalized primary circuit, core, pressurizer, and steam generators. The design of the application was performed using the object-oriented programming paradigm, incorporating an event-driven process to reflect the action of the human operators and the automatic control system. A comprehensive set of online graphical displays are provided giving an in-depth understanding of transient neutronic and thermal hydraulic response of the power plant. The model was validated against data from a real transient occurring in the Argentine NPP Embalse Río Tercero, showing good agreement. However, it should be stressed that the aim of the simulator is in the training of operators and engineering students.

  5. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  6. Lessons learned from tubes pulled from French steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Berge, Ph.; Boursier, J.M.; Dallery, D.; De Keroulas, F.; Rouillon, Y. [Electricite de France, Generating and Transmission Div. (France)

    1998-07-01

    Since 1981, the Chinon Hot Laboratory has completed more than 380 metallurgical examinations of pulled French steam generator tubes. Electricite de France decided to perform such investigations from the very outset of the French nuclear program, in order to contribute to nuclear power plant safety. The main reasons for withdrawing tubes are to evaluate the degradation, to validate non destructive examination (NDE) techniques, to gain a better understanding of cracking phenomena, and to ensure that the criteria on which plugging operations are based remain conservative. Considerable experience has been accumulated in the field of primary water stress corrosion cracking (PWSCC), OD (secondary) side corrosion, leak and burst tests, and various tube plugging techniques. This paper focuses on the PWSCC phenomenon and on the secondary side corrosion process, and in particular, attempts to correlate French data from pulled tubes with the results of fundamental R and D studies. Finally, within the framework of the Nuclear Power Plant Safety and Maintenance Policy, all these results are discussed in terms of optimization of the field inspection of tube bundles and plugging criteria. (author)

  7. Analysis of the State of Steam Generator Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Bergunker, Olga [JSC OKB ' Gidropress' , 142103 Podolsk (Russian Federation)

    2008-07-01

    The problem of safe operation of SG heat exchanging tubes, of both economical and effective control of their state is still important these days. Issues connected with peculiarities of methods of SG tubes inspection, automated analysis of the inspection results, tubes state analysis and development of algorithms of forecasting their state are considered in this report. The need for effective use of extensive data arrays on SG operation has led to the necessity of creating software tools for collection, storage and analysis of these data. The data-analytical system 'NPP Steam Generators' meant for data systematization and visualization as well as various types of analyses of data on eddy current inspection of WWER-440 and WWER-1000 SG tubes is presented in this report. The main possibilities of the data-analytical system (DAS), the code current state and prospects of its development are shown. The main fields of DAS application are considered and some results of its practical use are mentioned, namely, in the field of forecasting SG tubes state. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  9. Analysis of Steam Generators Corrosion Products from Slovak NPP Bohunice

    Directory of Open Access Journals (Sweden)

    Jarmila Degmová

    2012-01-01

    Full Text Available One of the main goals of the nuclear industry is to increase the nuclear safety and reliability of nuclear power plants (NPPs. As the steam generator (SG is the most corrosion sensitive component of NPPs, it is important to analyze the corrosion process and optimize its construction materials to avoid damages like corrosion cracking. For this purpose two different kinds of SGs and its feed water distributing systems from the NPP Jaslovske Bohunice were studied by nondestructive Mössbauer spectroscopy. The samples were scraped from the surface and analyzed in transmission geometry. Magnetite and hematite were found to be the main components in the corrosion layers of both SGs. Dependant of the material the SG consisted of, and the location in the system where the samples were taken, the ratios between magnetite and hematite and the paramagnetic components were different. The obtained results can be used to improve corrosion safety of the VVER-440 secondary circuit as well as to optimize its water chemistry regime.

  10. Direct solar steam generation inside evacuated tube absorber

    Directory of Open Access Journals (Sweden)

    Khaled M. Bataineh

    2016-12-01

    Full Text Available Direct steam generation by solar radiation falling on absorber tube is studied in this paper. A system of single pipe covered by glass material in which the subcooled undergoes heating and evaporation process is analyzed. Mathematical equations are derived based on energy, momentum and mass balances for system components. A Matlab code is built to simulate the flow of water inside the absorber tube and determine properties of water along the pipe. Widely accepted empirical correlations and mathematical models of turbulent flow, pressure drop for single and multiphase flow, and heat transfer are used in the simulation. The influences of major parameters on the system performance are investigated. The pressure profiles obtained by present numerical solution for each operation condition (3 and 10 MPa matches very well experimental data from the DISS system of Plataforma Solar de Almería. Furthermore, results obtained by simulation model for pressure profiles are closer to the experimental data than those predicted by already existed other numerical model.

  11. Optimization Method of Chemical Cleaning of Horizontal Steam Generator in PWR

    Institute of Scientific and Technical Information of China (English)

    TIAN; Jue; WANG; Hui; CAO; Lin-yuan

    2015-01-01

    China has introduced two WWER-1000/428nuclear power units from Russia at present,each unit includes four PGV-1000M horizontal steam generators.According to the components data analysis of horizontal steam generator’s secondary side sediments,the single variable functions optimizing test was made based on the original

  12. CANDU in-reactor quantitative visual-based inspection techniques

    Science.gov (United States)

    Rochefort, P. A.

    2009-02-01

    This paper describes two separate visual-based inspection procedures used at CANDU nuclear power generating stations. The techniques are quantitative in nature and are delivered and operated in highly radioactive environments with access that is restrictive, and in one case is submerged. Visual-based inspections at stations are typically qualitative in nature. For example a video system will be used to search for a missing component, inspect for a broken fixture, or locate areas of excessive corrosion in a pipe. In contrast, the methods described here are used to measure characteristic component dimensions that in one case ensure ongoing safe operation of the reactor and in the other support reactor refurbishment. CANDU reactors are Pressurized Heavy Water Reactors (PHWR). The reactor vessel is a horizontal cylindrical low-pressure calandria tank approximately 6 m in diameter and length, containing heavy water as a neutron moderator. Inside the calandria, 380 horizontal fuel channels (FC) are supported at each end by integral end-shields. Each FC holds 12 fuel bundles. The heavy water primary heat transport water flows through the FC pressure tube, removing the heat from the fuel bundles and delivering it to the steam generator. The general design of the reactor governs both the type of measurements that are required and the methods to perform the measurements. The first inspection procedure is a method to remotely measure the gap between FC and other in-core horizontal components. The technique involves delivering vertically a module with a high-radiation-resistant camera and lighting into the core of a shutdown but fuelled reactor. The measurement is done using a line-of-sight technique between the components. Compensation for image perspective and viewing elevation to the measurement is required. The second inspection procedure measures flaws within the reactor's end shield FC calandria tube rolled joint area. The FC calandria tube (the outer shell of the FC) is

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Overview of steam generator tube-inspection technology

    Energy Technology Data Exchange (ETDEWEB)

    Obrutsky, L.; Renaud, J.; Lakhan, R., E-mail: obrutskl@aecl.ca, E-mail: renaudj@aecl.ca, E-mail: lakhanr@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-03-15

    Degradation of steam generator (SG) tubing due to both mechanical and corrosion modes has resulted in extensive repairs and replacement of SGs around the world. The variety of degradation modes challenges the integrity of SG tubing and, therefore, the stations' reliability. Inspection and monitoring aimed at timely detection and characterization of the degradation is a key element for ensuring tube integrity. Up to the early-70's, the in-service inspection of SG tubing was carried out using single-frequency eddy current testing (ET) bobbin coils, which were adequate for the detection of volumetric degradation. By the mid-80's, additional modes of degradation such as pitting, intergranular attack, and axial and circumferential inside or outside diameter stress corrosion cracking had to be addressed. The need for timely, fast detection and characterization of these diverse modes of degradation motivated the development in the 90's of inspection systems based on advanced probe technology coupled with versatile instruments operated by fast computers and remote communication systems. SG inspection systems have progressed in the new millennium to a much higher level of automation, efficiency and reliability. Also, the role of Non Destructive Evaluation (NDE) has evolved from simple detection tools to diagnostic tools that provide input into integrity assessment decisions, fitness-far-service and operational assessments. This new role was motivated by tighter regulatory requirements to assure the safety of the public and the environment, better SG life management strategies and often self-imposed regulations. It led to the development of advanced probe technologies, more reliable and versatile instruments and robotics, better training and qualification of personnel and better data management and analysis systems. This paper provides a brief historical perspective regarding the evolution of SG inspections and analyzes the motivations behind that

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Garcia, M.S.; Castano, M.L.; Lancha, A.M. [CIEMAT, Madrid (Spain)

    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.

  17. coustic Leak Detection Based on Wavelet Packet and Genetic Algorithm for LM FBR Steam Generators

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Steam generator is one kind of key equipments in liquid metal fast breeder reactors (LM FBR) whose reliability will influence the safety of nuclear power plant. We can see that SG is the highest risky equipment from the running experience

  18. Modeling of soluble impurities distribution in the steam generator secondary water

    Energy Technology Data Exchange (ETDEWEB)

    Matal, O.; Simo, T. [Energovyzkum s.r.o., Brno (Switzerland); Kucak, L.; Urban, F. [Slovak Technical Univ., Bratislava (Slovakia)

    1997-12-31

    A model was developed to compute concentration of impurities in the WWER 440 steam generator (SG) secondary water along the tube bundle. Calculated values were verified by concentration values obtained from secondary water sample chemical analysis. (orig.). 2 refs.

  19. Numerical discretization analysis of a HTR steam generator model for the thermal-hydraulics code trace

    Directory of Open Access Journals (Sweden)

    Esch Markus

    2014-01-01

    Full Text Available For future high temperature reactor projects, e. g., for electricity production or nuclear process heat applications, the steam generator is a crucial component. A typical design is a helical coil steam generator consisting of several tubes connected in parallel forming cylinders of different diameters. This type of steam generator was a significant component used at the thorium high temperature reactor. In the work presented the temperature profile is being analyzed by the nodal thermal hydraulics code TRACE for the thorium high temperature reactor steam generator. The influence of the nodalization is being investigated within the scope of this study and compared to experimental results from the past. The results of the standard TRACE code are compared to results using a modified Nusselt number for the primary side. The implemented heat transfer correlation was developed within the past German HTR program. This study shows that both TRACE versions are stable and provides a discussion of the nodalization requirements.

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

    Energy Technology Data Exchange (ETDEWEB)

    Verma, K.; MacNeil, C. [New Brunswick Power Corp., Lepreau (Canada); Odar, S.; Kuhnke, K. [Siemens AG, Erlangen (Germany)

    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.

  1. Design and Activation of a LOX/GH Chemical Steam Generator

    Science.gov (United States)

    Saunders, G. P.; Mulkey, C. A.; Taylor, S. A.

    2009-01-01

    The purpose of this paper is to give a detailed description of the design and activation of the LOX/GH fueled chemical steam generator installed in Cell 2 of the E3 test facility at Stennis Space Center, MS (SSC). The steam generator uses a liquid oxygen oxidizer with gaseous hydrogen fuel. The combustion products are then quenched with water to create steam at pressures from 150 to 450 psig at temperatures from 350 to 750 deg F (from saturation to piping temperature limits).

  2. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.; Snell, V.; Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); West, J. [Candesco Co., Toronto, Ontario (Canada)

    2006-09-15

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life, and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross electrical output of 1165 MWe. The ACR-1000 design has evolved from AECL's in-depth knowledge of CANDU systems, components, and materials, as well as the experience and feedback received from owners and operators of CANDU plants. The ACR design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. It also features major improvements in economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The CANDU system is ideally suited to this evolutionary approach since the modular fuel channel reactor design can be modified, through a series of incremental changes in the reactor core design, to increase the power output and improve the overall safety, economics, and performance. The safety enhancements made in ACR-1000 encompass improved safety margins, performance and reliability of safety related systems. In particular, the use of the CANFLEX-ACR fuel bundle, with lower linear rating and higher critical heat flux, provides increased operating and safety margins. Safety features draw from those of the existing CANDU plants (e.g., the two

  3. Modeling of an once through helical coil steam generator of a superheated cycle for sizing analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Sik; Sim, Yoon Sub; Kim, Eui Kwang [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    A thermal sizing code, named as HSGSA (Helical coil Steam Generator Sizing Analyzer), for a sodium heated helical coil steam generator is developed for KALIMER (Korea Advanced LIquid MEtal Reactor) design. The theoretical modeling of the shell and tube sides is described and relevant correlations are presented. For assessment of HSGSA, a reference plant design case is compared to the calculational outputs from HSGSA simulation. 9 refs., 6 figs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Chedeau, C.; Rassineux, B. [EDF/DER/MTC, Moret Sur Loing (France); Flesch, B. [EDF/EPN/DMAINT, Paris (France)] [and others

    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.

  5. Mathematical modelling of steam generator and design of temperature regulator

    Energy Technology Data Exchange (ETDEWEB)

    Bogdanovic, S.S. [EE Institute Nikola Tesla, Belgrade (Yugoslavia)

    1999-07-01

    The paper considers mathematical modelling of once-through power station boiler and numerical algorithm for simulation of the model. Fast and numerically stable algorithm based on the linearisation of model equations and on the simultaneous solving of differential and algebraic equations is proposed. The paper also presents the design of steam temperature regulator by using the method of projective controls. Dynamic behaviour of the system closed with optimal linear quadratic regulator is taken as the reference system. The desired proprieties of the reference system are retained and solutions for superheated steam temperature regulator are determined. (author)

  6. Simulation of the Thermal Hydraulic Processes in the Horizontal Steam Generator with the Use of the Different Interfacial Friction Correlations

    Directory of Open Access Journals (Sweden)

    Vladimir Melikhov

    2011-01-01

    Full Text Available The horizontal steam generator (SG is one of specific features of Russian-type pressurized water reactors (VVERs. The main advantages of horizontal steam generator are connected with low steam loads on evaporation surface, simple separation scheme and high circulation ratio. The complex three-dimensional steam-water flows in the steam generator vessel influence significantly the processes of the steam separation, distribution, and deposition of the soluble and nonsoluble impurities and determine the efficiency and reliability of the steam generator operation. The 3D code for simulation of the three-dimensional steam-water flows in the steam generator could be effective tool for design and optimization of the horizontal steam generator. The results of the code calculations are determined mainly by the set of the correlations describing interaction of the steam-water mixture with the inner constructions of the SG and interfacial friction. The results obtained by 3D code STEG with the usage of the different interfacial friction correlations are presented and discussed in the paper. These results are compared with the experimental ones obtained at the experimental test facility PGV-1500 constructed for investigation of the processes in the horizontal steam generator.

  7. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir, E-mail: luciano.ondir@gmail.com; Andrade, Delvonei Alves de, E-mail: delvonei@ipen.br

    2014-12-15

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs.

  8. Steam generation process control and automation; Automacao e controle no processo de geracao de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Souza Junior, Jose Cleodon de; Silva, Walmy Andre C.M. da [PETROBRAS S.A., Natal, RN (Brazil)

    2004-07-01

    This paper describes the implementation of the Supervisory Control and Data Acquisition System (SCADA) in the steam generation process for injection in heavy oil fields of the Alto do Rodrigues Production Asset, developed by PETROBRAS/E and P/UN-RNCE. This Asset is located in the northeastern region of Brazil, in Rio Grande do Norte State. It addresses to the steam generators for injection in oil wells and the upgrade project that installed remote terminal units and a new panel controlled by PLC, changed all the pneumatic transmitters by electronic and incorporated the steam quality and oxygen control, providing the remote supervision of the process. It also discusses the improvements obtained in the steam generation after the changes in the conception of the control and safety systems. (author)

  9. Ecotaxes and their impact in the cost of steam and electric energy generated by a steam turbine system

    Energy Technology Data Exchange (ETDEWEB)

    Montero, Gisela [Universidad Autonoma de Baja California, Mexicali, B.C. (Mexico). Instituto de Ingenieria; Pulido, Ricardo; Pineda, Carlos; Rivero, Ricardo [Instituto Mexicano del Petroleo, Eje Lazaro Cardenas, D.F. (Mexico)

    2006-12-15

    Ecotaxes allow the internalization of costs that are considered externalities associated with polluting industrial process emissions to the atmosphere. In this paper, ecotaxes internalize polluting emissions negative impacts that are added to electricity and steam generated costs of a steam turbine and heat recovery systems from a utilities refinery plant. Steam costs were calculated by means of an exergy analysis tool and Aspen Plus simulation models. Ecotaxes were calculated for specific substances emitted in the refinery flue gases, based on a toxicity and pollution scale. Ecotaxes were generated from a model that includes damages produced to biotic and abiotic resources and considers the relative position of those substances in a toxicity and pollution scale. These ecotaxes were internalized by an exergoeconomic analysis resulting in an increase in the cost per kWh produced. This kind of ecotax is not applied in Mexico. The values of ecotaxes used in the cost determination are referred to the values currently applied by some European countries to nitrogen oxides emissions. (author)

  10. French steam generator tubes: an overview of degradations

    Energy Technology Data Exchange (ETDEWEB)

    Buisine, D.; Bouvier, O. de; Rupa, N.; Thebault, Y.; Barbe, V. [EDF-CEIDRE Nuclear Engineering Division (France); Pitner, P. [EDF-UNIE Generation Nuclear Operation Division (France)

    2011-07-01

    The various damages (corrosion, fatigue cracks, wear, ...) observed on steam generator (SG) tubes are presented here as well as the techniques used to characterize these damages. The SG are equipped with tubes of 3 materials: 600 MA, 600 TT and 690 TT. Concerning PWSCC of 600 MA and 600 TT tubes, beyond the damages usually observed (corrosion in expansion transition zone and in 600 MA tubes small radius U-bend zone), a new event is to be noted: the phenomenon of denting (presumably induced by the deposit of sludge on the tubesheet) has induced circumferential cracking of the tube expansion transition zone. Concerning ODSCC of 600 MA tubes, beyond the classically observed damages (IGA and IGSCC in expansion transition zone and in TSP crevice), a new event is to be noted: the occurrence of circumferential cracks in tube- TSP crevice. Concerning fatigue cracking, two events have to be noted at upper TSP level in Cruas 1 and Cruas 4 units and in Fessenheim 2 unit. The first (Cruas) was due to the blockage in the broached hole tube support plate which can create critical velocity ratios for some tubes and the second (Fessenheim) to high-cycle fatigue. Concerning wear damage, beyond what is usually observed in the U-bend zone facing the anti-vibration bars (AVB), a new event is to be noted: a wear at TSP level is observed on SG equipped with an economizer, the wear indications being located at TSP 7 and 8 level, on outer tubes close to the central lane. The number of tubes plugged for ODSCC has declined due to the progressive replacement of SG with Alloy 600 MA tubing. Starting in 2004, the increasing plugging of 690 tubing is mainly due to AVB wear. Since 2006, extensive preventive plugging campaigns for tubes at risk of high-cycle fatigue at the upper support plate are performed. Risk of high-cycle fatigue has consequently become the dominant mechanism inducing plugging. PWSCC is the second dominant mechanism which affects 600 MA and 600 TT tube bundles: extensive

  11. Disposal of Steam Generators from Decommissioning of PWR Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Walberg, Mirko; Viermann, Joerg; Beverungen, Martin [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestrasse 7A, 45127 Essen (Germany); Kemp, Lutz [Kernkraftwerk Stade GmbH and Co.oHG, Bassenflether Chaussee, 21683 Stade (Germany); Lindstroem, Anders [Studsvik Nuclear AB, SE-611 82 Nykoeping (Sweden)

    2008-07-01

    Amongst other materials remarkable amounts of radioactively contaminated or activated scrap are generated from the dismantling of Nuclear Power Plants. These scrap materials include contaminated pipework, fittings, pumps, the reactor pressure vessel and other large components, most of them are heat exchangers. Taking into account all commercial and technical aspects an external processing and subsequent recycling of the material might be an advantageous option for many of these components. The disposal of steam generators makes up an especially challenging task because of their measures, their weight and compared to other heat exchangers high radioactive inventory. Based on its experiences from many years of disposal of smaller components of NPP still in operation or under decommissioning GNS and Studsvik Nuclear developed a concept for disposal of steam generators, also involving experiences made in Sweden. The concept comprises transport preparations and necessary supporting documents, the complete logistics chain, steam generator treatment and the processing of arising residues and materials not suitable for recycling. The first components to be prepared, shipped and treated according to this concept were four steam generators from the decommissioning of the German NPP Stade which were removed from the plant and shipped to the processing facility during the third quarter of 2007. Although the plant had undergone a full system decontamination, due to the remaining contamination in a number of plugged tubes the steam generators had to be qualified as industrial packages, type 2 (IP-2 packages), and according to a special requirement of the German Federal Office for Radiation Protection a license for a shipment under special arrangement had to be applied for. The presentation gives an overview of the calculations and evidences required within the course of the IP-2 qualification, additional requirements of the competent authorities during the licensing procedure as

  12. Correlation Dimension in Fault Diagnosis of 600 MW Steam Turbine Generator

    Institute of Scientific and Technical Information of China (English)

    YAO Bao-heng; YANG Xia-ju; TONG De-chun; CHEN Zhao-neng

    2005-01-01

    GP algorithm of correlation dimension computation is ameliorated which overcomes the shortage of traditional one. Improved process of GP algorithm takes the influence of temporal correlative pairs of points on correlation dimension into account and promotes the computational efficiency prominently. Iterative SVD method is applied to remove the influence of noise on the result of correlation dimension. The faults of steam flow turbulence and oil film disturbance which occur in 600MW Steam Turbine Generator are analyzed and whose correlation dimensions are computed. More distinct quantitative index than FFT is gained to distinguish two faults and it's of little importance to apply correlation dimension to study the influence of various factors on steam flow turbulence fault for nonexistence of convergent floor in correlation integral curve, which presents a new way to learn the operational function of large capacity steam turbine generator and carry out comprehensive condition monitoring.

  13. Energy balance for steam generation system with biomass dryer

    Energy Technology Data Exchange (ETDEWEB)

    Ramos, Pedro A.R. [Instituto Superior Politecnico Jose Antonio Echeverria (CUJAE), Ciudad de La Habana (Cuba). Facultad Ingenieria Mecanica]. E-mail: pedro@economia.cujae.edu.cu; Lombardi, Geraldo; Santos, Antonio Moreira dos [Universidade de Sao Paulo (USP), Sao Carlos, SP (Brazil). Escola de Engenharia]. E-mails: lombardi@sc.usp.br; asantos@sc.usp.br

    2008-07-01

    Water content is a major drainer of the energy available in the biomass, which justifies the proposal of a drying system with the potential to increase 80% of the biomass low heating value, also increasing the production of steam in the boiler and cogeneration of electricity. An example of biomass is the sugar cane bagasse of an alcohol mill producing 120,000 liters of alcohol per day, whose humidity from the extraction section is usually 50%. The present paper determines the increases in the mass flow rates of steam in the boiler, in the cogeneration of electricity and in the pay back time of the drying system and of the alcohol mill, as a consequence of the bagasse drying from 50 to 35%, considering 30% of air excess over the stoichiometric value admitted in the boiler for the bagasse burning. It also provides subsidies for the development and deployment of a drying system for the current boilers. (author)

  14. Effect of combined slow pyrolysis and steam gasification of sugarcane bagasse on hydrogen generation

    Energy Technology Data Exchange (ETDEWEB)

    Parthasarathy, Prakash; Narayanan, Sheeba [National Institute of Technology, Tamil Nadu (India)

    2015-11-15

    The present work aims at improving the generation of H2 from sugarcane bagasse in steam gasification process by incorporating slow pyrolysis technique. As a bench scale study, slow pyrolysis of sugarcane bagasse is performed at various pyrolysis temperature (350, 400, 450, 500 and 550 .deg. C) and feed particle size (90generation. In the combined process (slow pyrolysis of biomass followed by steam gasification of char), first slow pyrolysis is carried out at the effective conditions (pyrolysis temperature and particle size) of char generation (determined from bench scale study) and steam gasification is at varying gasification temperature (600, 650, 700, 750 and 800 .deg. C) and steam to biomass (S/B) ratio (1, 2, 3, 4, 5 and 6) to determine the effective conditions of H{sub 2} generation. The effect of temperature and S/B on gas product composition and overall product gas volume was also investigated. At effective conditions (gasification temperature and S/B) of H2 generation, individual slow pyrolysis and steam gasification were also experimented to evaluate the performance of combined process. The effective condition of H{sub 2} generation in combined process was found to be 800 .deg. C (gasification temperature) and 5 (S/B), respectively. The combined process produced 35.90% and 23.60% more gas volume (overall) than slow pyrolysis and steam gasification process, respectively. With respect to H{sub 2} composition, the combined process generated 72.37% more than slow pyrolysis and 17.91% more than steam gasification process.

  15. P controller with partial feed forward compensation and decoupling control for the steam generator water level

    Energy Technology Data Exchange (ETDEWEB)

    Liu Cheng, E-mail: liuch_2004@stu.xjtu.edu.c [School of Energy and Power Engineering, Xi' an Jiaotong University, Xi' an, Shaanxi 710049 (China); Zhao Fuyu; Hu Ping; Hou Suxia; Li Chong [School of Energy and Power Engineering, Xi' an Jiaotong University, Xi' an, Shaanxi 710049 (China)

    2010-01-15

    In this paper, a P controller with partial feed forward compensation and decoupling control for the steam generator water level is presented. While taking the steam flowrate as a disturbance to water level, the controller design can be completed in three stages. (1) Main circuit controller is designed without regard to disturbance. Since the transfer function of the steam generator model contains integrate element and differential element, the proportional (P) controller can selected as main circuit controller instead of PID controller for steam generator water level. (2) Partial feed forward compensation is introduced to remove the disturbance from the steam flowrate. If disregarding the differential element, the partial feed forward compensation's designing turns to be very simple. Partial feed forward compensation coefficient is set as reciprocal of P controller gain. (3) The coupling effects between the water level regulating and steam flowrate disturbance can be decreased by model reference decoupling control. The proposed methodology shows satisfactory transient responses, disturbance rejection and robustness.

  16. The small (or large) modular CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Meneley, D.; Harvel, G. [Univ. of Ontario Inst. of Tech., Oshawa, Ontario (Canada)

    2013-07-01

    This presentation outlines the design for small (or large) modular CANDU. The origins of this work go back many years to a comment by John Foster, then President of AECL CANDU. Foster noted that the CANDU reactor, with its many small fuel channels, was like a wood campfire. To make a bigger fire, just throw on some more logs (channels). If you want a smaller fire, just use fewer logs. The design process is greatly simplified.

  17. Starting up of new steam generator of N4 1450 MWE plants

    Energy Technology Data Exchange (ETDEWEB)

    Bussy, B. [EDF, Villeurbanne Cedex (France); Dague, G.; Slama, G. [Framatome, Paris-le-Defense (France)

    1998-07-01

    The first N4 plant, CHOOZ B1, was commissioned in 1997. This plant of 1450 MWe capacity is equipped with four steam generators of a new design fitted with an axial economiser. The axial economiser principle essentially consists in directing all the feedwater to the cold leg of the tube bundle and about 90% of the recirculated water to the hot leg. A double wrapper alone, the downcomer allows cold water to enter tube bundle as in a boiler-type steam generator andprevents excessive vibration due to cross flow. The main result of this design is to enhance the heat exchange efficiency between the primary and the secondary sides and to increase the steam pressure as compared to a boiler steam generator having the same heat exchange area. As it is the common practice in France for the first plant unit of a new model, one of the four steam generators has been specifically instrumented in order to assess its actual operating characteristics and verify their consistency with the predicted values resulting from the design studies and from the qualification tests. The test programme and the dedicated instrumentation allowed measurement of essential S.G. thermal hydraulics parameters (saturation pressure, flow velocities, pressure drops, circulation ratios, temperature distribution,...) and assessment of vibratory behaviour of tube bundle. The results are in agreement with the tests carried out in the course of the steam generator development and qualification programme, notably a few years ago on the large 25 MWth MEGEVE test facility. They confirm adequacy of the axial economiser design. The paper describes successively: the specificities of the new steam generator design; the instrumentation and the test programme; and the main results obtained from the numerous tests performed during plant start-up. (author)

  18. Optimum thermal sizing and operating conditions for once through steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Yi, Kunwoo; Ju, Kyongin; Im, Inyoung; Kim, Eunkee [KEPCO Engineering and Construction Company., Inc., Daejeon (Korea, Republic of)

    2014-10-15

    The steam generator is designed to be optimized so as to remove heat and to produce steam vapor. Because of its importance, theoretical and experimental researches have been performed on forced convection boiling heat transfer. The purpose of this study is to predict the thermal behavior and to perform optimum thermal sizing of once through steam generator. To estimate the tube thermal sizing and operating conditions of the steam generator, the analytical modeling is employed on the basis of the empirical correlation equations and theory. The optimized algorithm model, Non-dominated Sorting Genetic Algorithm (NSGA)-II, uses for this analysis. This research is focused on the design of in-vessel steam generator. An one dimensional analysis code is developed to evaluate previous researches and to optimize steam generator design parameters. The results of one-dimensional analysis need to be verified with experimental data. Goals of multi-objective optimization are to minimize tube length, pressure drop and tube number. Feedwater flow rate up to 115.425kg/s is selected so as to have margin of feedwater temperature 20 ..deg. C. For the design of 200MWth once through steam generator, it is evaluated that the tube length shall be over 12.0m for the number of tubes, 2500ea, and the length of the tube shall be over 8.0m for the number of tubes, 4500ea. The parallel coordinates chart can be provided to determine the optimal combination of number of tube, pressure drop, tube diameter and length.

  19. Proceedings of the third international steam generator and heat exchanger conference

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

    The Third International Steam Generator and Heat Exchanger conference had the objective to present the state of knowledge of steam generator performance and life management, and also heat exchanger technology. As this conference followed on from the previous conferences held in Toronto in 1990 and 1994, the emphasis was on recent developments, particularly those of the last 4 years. The conference provided an opportunity to operators, designers and researchers in the field of steam generation associated with electricity generation by nuclear energy to present their findings and exchange ideas. The conference endeavoured to do this over the widest possible range of subject areas,including: general operating experience, life management and fitness for service strategies, maintenance and inspection, thermalhydraulics, vibration, fretting and fatigue, materials, chemistry and corrosion and the regulatory issues.

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

    Energy Technology Data Exchange (ETDEWEB)

    Scheveneels, G. [AIB-Vincotte Nuclear, Brussels (Belgium)

    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.

  1. Modelling horizontal steam generator with ATHLET. Verification of different nodalization schemes and implementation of verified constitutive equations

    Energy Technology Data Exchange (ETDEWEB)

    Beliaev, J.; Trunov, N.; Tschekin, I. [OKB Gidropress (Russian Federation); Luther, W. [GRS Garching (Germany); Spolitak, S. [RNC-KI (Russian Federation)

    1995-12-31

    Currently the ATHLET code is widely applied for modelling of several Power Plants of WWER type with horizontal steam generators. A main drawback of all these applications is the insufficient verification of the models for the steam generator. This paper presents the nodalization schemes for the secondary side of the steam generator, the results of stationary calculations, and preliminary comparisons to experimental data. The consideration of circulation in the water inventory of the secondary side is proved to be necessary. (orig.). 3 refs.

  2. Next Generation Engineered Materials for Ultra Supercritical Steam Turbines

    Energy Technology Data Exchange (ETDEWEB)

    Douglas Arrell

    2006-05-31

    To reduce the effect of global warming on our climate, the levels of CO{sub 2} emissions should be reduced. One way to do this is to increase the efficiency of electricity production from fossil fuels. This will in turn reduce the amount of CO{sub 2} emissions for a given power output. Using US practice for efficiency calculations, then a move from a typical US plant running at 37% efficiency to a 760 C /38.5 MPa (1400 F/5580 psi) plant running at 48% efficiency would reduce CO2 emissions by 170kg/MW.hr or 25%. This report presents a literature review and roadmap for the materials development required to produce a 760 C (1400 F) / 38.5MPa (5580 psi) steam turbine without use of cooling steam to reduce the material temperature. The report reviews the materials solutions available for operation in components exposed to temperatures in the range of 600 to 760 C, i.e. above the current range of operating conditions for today's turbines. A roadmap of the timescale and approximate cost for carrying out the required development is also included. The nano-structured austenitic alloy CF8C+ was investigated during the program, and the mechanical behavior of this alloy is presented and discussed as an illustration of the potential benefits available from nano-control of the material structure.

  3. Experience in adjusting of the level regulation system of steam generators of the Rovno NPP

    Energy Technology Data Exchange (ETDEWEB)

    Patselyuk, S.N.; Sokolov, A.G.; Kazakov, V.I.; Dorosh, Yu.A.

    1984-07-01

    A system of feed water level control in steam generators at the Rovno NPP with WWER-440 reactors which comprises start-up as well as main regulators is described. The start-up regulator (single-pulsed with a signal by the level) keeps the level in the steam generator at loadings up to 30% of the nominal reactor power Nsub(nom.). The main regulator is connected in the three-pulsed circuit and it receives signals by steam and water flow rate and by the level in the steam generator. The main regulator has been started only at loadings above 40% Nsub(nom.). After reconstruction it was used in the 15-100% Nsub(nom.) range. Characteristics of the level control system in the steam generator at perturbations intoduced by the main circulating pump (MCP) and turbine disconnection as well as change in feed water flow rate have been studied. The studies have revealed that the system ensures necessary quality of control in stationary modes. The system operates stably at perturbations of feed water flow rate up to 50% Nsub(nom.). Perturbations by MCP connections and disconnections is most difficult for control system.

  4. Microfabricated rankine cycle steam turbine for power generation and methods of making the same

    Science.gov (United States)

    Frechette, Luc (Inventor); Muller, Norbert (Inventor); Lee, Changgu (Inventor)

    2009-01-01

    In accordance with the present invention, an integrated micro steam turbine power plant on-a-chip has been provided. The integrated micro steam turbine power plant on-a-chip of the present invention comprises a miniature electric power generation system fabricated using silicon microfabrication technology and lithographic patterning. The present invention converts heat to electricity by implementing a thermodynamic power cycle on a chip. The steam turbine power plant on-a-chip generally comprises a turbine, a pump, an electric generator, an evaporator, and a condenser. The turbine is formed by a rotatable, disk-shaped rotor having a plurality of rotor blades disposed thereon and a plurality of stator blades. The plurality of stator blades are interdigitated with the plurality of rotor blades to form the turbine. The generator is driven by the turbine and converts mechanical energy into electrical energy.

  5. The development of an inspection/maintence robot for steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Ho; Kim, Chang Hoi; Seo, Yong Chil [and others

    2003-05-01

    We developed the tele-robotic systems for inspection/maintenance of steam generator tubes. For easy handling and installation, it consists of three separable parts: the entering/leaving device, the base posture adjusting device, the manipulator. The inspection and repair tools, such as brushing, plugging, and sleeving tools, were developed. We also developed software programs for the eddy current test signal acquisition and evaluation. The semiconductor type dosimeter and the directional radiation mapping module were developed for measuring the accumulated radioactivity and for finding the radioactivity source location. The research for radiation shield and decontamination were carried out. The developed robotic system has been tested in the Ulchin NPP type steam generator mockup in our laboratory, and after evaluation and some modification the final functional test was carried out at the Kori NPP type steam generator mockup in the Kori training center.

  6. Distribution of liquid sodium in the inlet plenum of steam generator in a Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Patil, Laxman T. [Department of Chemical Engineering, Institute of Chemical Technology, N. M. Parikh Marg, Matunga, Mumbai 400019 (India); Patwardhan, A.W., E-mail: awp@udct.or [Department of Chemical Engineering, Institute of Chemical Technology, N. M. Parikh Marg, Matunga, Mumbai 400019 (India); Padmakumar, G.; Vaidyanathan, G. [Experimental Thermal Hydraulics Section, Separation Technology and Hydraulics Division, Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2010-04-15

    Experimental and Computational Fluid Dynamics (CFD) investigations have been carried out on a 1/5th scale model of the inlet plenum of steam generator (SG) used in the Fast Breeder Reactor (FBR) technology. The distribution of liquid sodium in the inlet plenum of the steam generator strongly affects the thermal as well as mechanical performance of the steam generator. In the present work, flow distribution in a scaled down model has been investigated. Various strategies adopted for obtaining uniform flow distribution have been evaluated. Experiments have been conducted to measure the axial and radial velocity distributions using Ultrasonic Velocity Profiler (UVP) under a variety of geometries. Computational Fluid Dynamics (CFD) studies have been carried out for various geometries. On the basis of these experiments and CFD simulations, various flow distribution devices have been compared.

  7. Review of EPRI's steam generator R and D program

    Energy Technology Data Exchange (ETDEWEB)

    Millett, P.J.; Welty, C.J. [EPRI, Palo Alto, CA (United States)

    1998-07-01

    EPRI has carried out an extensive R and D program on SG technology since the mid 1970's. Very early efforts under the auspices of the Steam Generator Owners Group (SGOG) focused on developing remedial actions for the critical SG corrosion issues of denting, wastage and pitting. Fundamental work was also carried out in the development of thermal hydraulic models for vibration and wear, chemical cleaning and tube repair techniques. In the late 1980's and continuing through today, the program has shifted emphasis towards management of steam generator degradation, primarily stress corrosion cracking of the SG tubes on both the primary and secondary sides. The current Steam Generator Management Program (SGMP) carries out R and D in four areas; materials, chemistry, thermal hydraulics and non-destructive testing. The strategic goals of this program and projects put in place to achieve these goals will be reviewed in detail in this paper. (author)

  8. Development of safety evaluation technique of steam generator tubes for the next generation

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyuk Sang; Kim, I. S.; Ann, Se Jin; Lee, S. J.; Seo, M. S.; Lee, Y. H.; Kim, J. H.; Hong, J. G. [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-02-15

    Subject 1 - a technique for predicting the SCC susceptibility of steam generator tube material based on the repassivation kinetics was developed and the effects of Pb in the repassivation rate and SCC susceptibility rate of tube material was investigated with this technique. An alloy with a higher slope value of log i(t) vs. q(t) plot based on the current transient curve obtained by scratch test and a lower slope value log i(t) vs. l/q(t) plot (cBV) is repassivated faster with a more protective passive film and it can be predicted that it will show higher resistance to SCC. With PbO addition in all solution studied (pH 4, pH 10, Cl- containing pH 4), alloy 690TT showed decreased repassivation rate. So it can be predict that PbO addition lower the resistance of SCC of steam generator tune material. Subject 2 - SG wear testing of tube and support materials has been conducted at various load and sliding amplitude in air environment. The results showed effect of normal load and sliding amplitude on SG tube wear damage. It was also shown that, for predominantly sliding motion, the SG wear coefficient of work-rate model is lower for Inconel 690TT compared with inconel 600MA. SG tube wear data show that, for work-rates ranging from 4 to 25mW, average tube wear coefficient of 43.76{approx}54.05 X 10{sup 15} Pa{sup -1} for Inconel 600MA and 26.88{approx}33.94 X 10{sup -15} Pa{sup 1} for Inconel 690TT against 405 and 409 stainless steels.

  9. Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

    Energy Technology Data Exchange (ETDEWEB)

    D`Auria, F.; Galassi, G.M. [Univ. of Pisa (Italy); Frogheri, M. [Univ. of Genova (Italy)

    1997-12-31

    The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Feasibility of leak-detection instrumentation for duplex-tube steam generator. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Berkey, E.; Witkowski, R.E.

    1974-01-01

    A literature search has been carried out to determine if current state-of-the-art for sodium vapor and water vapor detectors are feasible as leak detection instrumentation for the Westinghouse duplex-tube steam generator. A commercially available probe-type water vapor detector has been identified and a thermal ionization type sodium vapor detector, currently being developed by Westinghouse, has been selected as the reference sodium-vapor leak detector. Recommendations are made concerning the experimental studies required to adapt the selected instrumentation to steam-generator plant applications. Proposed future instrumentation development programs are also identified.

  12. Steam generators and waste heat boilers for process and plant engineers

    CERN Document Server

    Ganapathy, V

    2014-01-01

    Incorporates Worked-Out Real-World ProblemsSteam Generators and Waste Heat Boilers: For Process and Plant Engineers focuses on the thermal design and performance aspects of steam generators, HRSGs and fire tube, water tube waste heat boilers including air heaters, and condensing economizers. Over 120 real-life problems are fully worked out which will help plant engineers in evaluating new boilers or making modifications to existing boiler components without assistance from boiler suppliers. The book examines recent trends and developments in boiler design and technology and presents novel idea

  13. Qualification of inspection systems in the CANDU nuclear industry

    Energy Technology Data Exchange (ETDEWEB)

    Baron, J.A. [CANDU Owners Group, CANDU Inspection Qualification Bureau, Toronto, Ontario (Canada)

    2014-01-15

    Most jurisdictions that generate electricity through nuclear-electric plants have imposed requirements on inspection systems beyond the typical Level 1, 2 and 3 found in personnel qualification/certification schemes. The paper discusses the rationale for this obligation and describes how the requirement for inspection qualification has been implemented for CANDU plants. The paper discusses the qualification structure and process, including a brief overview of experience to-date in qualifying Inspection Procedures. (author)

  14. 以蒸汽发生器替代蒸汽锅炉的可行性分析%FEASIBILITY ANALYSIS OF STEAM GENERATOR SUBSTITUTE FOR STEAM BOILER

    Institute of Scientific and Technical Information of China (English)

    李文红; 施天裕

    2014-01-01

    The article introduces the working principle of steam generator and steam boiler;analyzes the existing problems of steam boiler; compare the actual expense of the two; explains the advantages of steam generator; and comes a conclusion.%文章简单介绍了蒸汽锅炉和蒸汽发生器的工作原理,分析了蒸汽锅炉存在的问题;结合上海交通大学附属第六人民医院的实际使用情况,对蒸汽锅炉和蒸汽发生器的费用支出进行了对比;详细介绍了与蒸汽锅炉相比蒸汽发生器的优势,最终得出了结论。

  15. Once-through steam generator (OTSG) materials and water chemistry. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Pocock, F.J.; Levstek, D.F.

    1974-01-01

    Materials and water chemistry research results associated with the development of the Oconee-1 Reactor steam generator are presented. A summary of water chemistry data acquired during preoperational testing and power operation to date is also included. These data confirm the operational practicality of the nuclear once-through concept using volatile water treatment and high purity condensate demineralized feedwater.

  16. Key Findings from the Artist Project on Aerosol Retention in a Dry Steam Generator

    Directory of Open Access Journals (Sweden)

    Abdelouahab Dehbi

    2016-08-01

    Full Text Available A steam generator tube rupture (SGTR event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI initiated the Aerosol Trapping In Steam GeneraTor (ARTIST Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

  17. Development and field validation of advanced array probes for steam generator inspection

    Energy Technology Data Exchange (ETDEWEB)

    Dodd, C.V.; Pate, J.R. [Oak Ridge National Lab., TN (United States)

    1995-04-01

    The aging of the steam generators at the nation`s nuclear power plants has led to the appearance of new forms of degradation in steam generator tubes and an increase in the frequency of forced outages due to major tube leak events. The eddy-current techniques currently being used for the inspection of steam generator tubing are no longer adequate to ensure that flaws will be detected before they lead to a shutdown of the plant. To meet the need for a fast and reliable method of inspection, ORNL has designed a 16-coil eddy-current array probe which combines an inspection speed similar to that of the bobbin coil with a sensitivity to cracks of any orientation similar to the rotating pancake coil. In addition, neural network and least square methods have been developed for the automatic analysis of the data acquired with the new probes. The probes and analysis software have been tested at two working steam generators where we have found an increase in the signal-to-noise ratio of a factor of five an increase in the inspection speed of a factor of 75 over the rotating pancake coil which maintaining similar detection and characterization capabilities.

  18. 76 FR 74834 - Interim Staff Guidance on Aging Management Program for Steam Generators

    Science.gov (United States)

    2011-12-01

    ... COMMISSION Interim Staff Guidance on Aging Management Program for Steam Generators AGENCY: Nuclear Regulatory Commission. ACTION: Interim staff guidance; issuance. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing License Renewal Interim Staff Guidance (LR-ISG), LR-ISG-2011-02, ``Aging...

  19. Analysis on Non-Uniform Flow in Steam Generator During Steady State Natural Circulation Cooling

    Directory of Open Access Journals (Sweden)

    Susyadi

    2007-07-01

    Full Text Available Investigation on non uniform flow behavior among U-tube in steam generator during natural circulation cooling has been conducted using RELAP5. The investigation is performed by modeling the steam generator into multi channel models, i.e. 9-tubes model. Two situations are implemented, high pressure and low pressure cases. Using partial model, the calculation simulates situation similar to the natural circulation test performed in LSTF. The imposed boundary conditions are flow rate, quality, pressure of the primary side, feed water temperature, steam generator liquid level, and pressure in the secondary side. Calculation result shows that simulation using model with nine tubes is capable to capture important non-uniform phenomena such as reverse flow, fill-and-dump, and stagnant vertical stratification. As a result of appropriate simulation of non uniform flow, the calculated steam generator outlet flow in the primary loop is stable as observed in the experiments. The results also clearly indicate the importance of simulation of non-uniform flow in predicting both the flow stability and heat transfer between the primary and secondary side. In addition, the history of transient plays important role on the selection of the flow distribution among tubes. © 2007 Atom Indonesia. All rights reserved

  20. Hard Sludge Formation in Modern Steam Generators of Nuclear Power Plants Formation, Risks and Mitigation

    Energy Technology Data Exchange (ETDEWEB)

    Strohmer, F.

    2013-07-01

    This article will discuss the physical and chemical reasons for the increased tendency to form hard sludge on the secondary side of modern nuclear steam generators (SG). The mechanism of hard sludge induced denting will be explained. Moreover, advice on operation and maintenance to mitigate hard sludge formation and denting damages will be presented.

  1. CFD evaluation on the thermohydraulic characteristics of tube support plates in steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, B.; Zhang, H.; Han, B.; Yang, B.W. [Xi' an Jiaotong Univ. (China). School of Nuclear Science and Technology; Mo, S.J.; Ren, H.B.; Qin, J.M.; Zuo, C.P. [China Nuclear Power Design Co. Ltd., ShenZhen (China)

    2016-07-15

    The integrity and thermal hydraulic characteristics of steam generator are of great concern in the nuclear industry. The tube support plates (TSP), one of the most important components of the steam generator, not only support the heat transfer tubes, but also affect the flow dynamic and thermal hydraulic characteristics of the secondary-side flow inside the steam generator. Different working conditions, ranging from single-phase adiabatic condition to two-phase high-void boiling condition, are simulated and analyzed. Calculated void fraction, under simple geometry, agrees well with the experiment data whilst the simulated heat transfer coefficient is tremendously close to the empirical correlation. Temperature, void fraction, and velocity distributions in different locations show reasonable distribution. The simulation results indicate that TSP can enhance the heat transfer in the secondary side of the steam generator. On the top of TSP, with the increase in cross-section flow area, the back-flow phenomenon occurs, which might lead to the contamination of precipitation.

  2. Current Status on the Development of a Double Wall Tube Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Ho Yun; Choi, Byoung Hae; Kim, Jong Man; Kim, Byung Ho

    2007-12-15

    A fast reactor, which uses sodium as a coolant, has a lot of merits as a next generation nuclear reactor. However, the possibility of a sodium-water reaction occurrence hinders the commercialization of this reactor. As one way to improve the reliability of a steam generator, a double-wall tube steam generator is being developed in GEN-4 program. In this report, the current state of the technical developments for a double-wall tube steam generator are reviewed and a future plan for the development of a double-wall tube steam generator is established. The current focuses of this research are an improvement of the heat transfer capability for a double-wall tube and the development of a proper leak detection method for the failure of a double-wall tube during a reactor operation. The ideal goal is an on-line leak detection of a double wall tube to prevent the sodium-water reaction. However, such a method is not developed as yet. An alternative method is being used to improve the reliability of a steam generator by performing a non-destructive test of a double wall tube during the refueling period of a reactor. In this method a straight double wall tube is employed to perform this test easily, but has a difficulty regarding an absorption of a thermal expansion of the used materials. If an on-line leak detection method is developed, the demerits of a straight double-wall tube are avoided by using a helical type double-wall tube, and the probability of a sodium-water reaction can be reduced to a level less than the design-based accident.

  3. Experiences with Direct Steam Generation at the Kanchanaburi Solar Thermal Power Plant

    OpenAIRE

    Krüger, Dirk; Krüger, Joachim; Pandian, Yuvaraj; O'Connell, Bryan; Feldhoff, Jan Fabian; Karthikeyan, Ramkumar; Hempel, Sören; Muniasamy, Karthik; Hirsch, Tobias; Eickhoff, Martin; Hennecke, Klaus

    2012-01-01

    In 2011 the parabolic trough power plant TSE1 has started operation in Thailand. As a novelty it uses the direct steam generation (DSG) process, evaporating and super heating water and steam directly in the solar field. During the commissioning phase and first months of operation the start-up procedure has been optimised for the solar field and turbine system resulting in a reduced start-up time. The DSG process can be controlled well in the evaporator and super heater section securing a safe...

  4. Adaptive H-infinity control of synchronous generators with steam valve via Hamiltonian function method

    Institute of Scientific and Technical Information of China (English)

    Shujuan LI; Yuzhen WANG

    2006-01-01

    Based on Hamiltonian formulation, this paper proposes a design approach to nonlinear feedback excitation control of synchronous generators with steam valve control, disturbances and unknown parameters. It is shown that the dynamics of the synchronous generators can be expressed as a dissipative Hamiltonian system, based on which an adaptive H-infinity controller is then designed for the systems by using the structure properties of dissipative Hamiltonian systems.Simulations show that the controller obtained in this paper is very effective.

  5. Signal processing system design for improved shutdown system of CANDU{sup ®} nuclear reactors in large break LOCA events

    Energy Technology Data Exchange (ETDEWEB)

    Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Faculty of Engineering and Applied Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Xia, Lingzhi; Isham, Manir U. [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Ponomarev, Vladimir [Megawatt Solutions, 1235 Radom St., unit 68, Pickering, ON, Canada L1W 1J3 (Canada)

    2016-03-15

    Highlights: • Neutronic signal processing system design to improve CANDU SDS1 performance. • Reactor modeling for CANDU LLOCA transient. • MATLAB/Simulink system implementation for the SDS1 trip logic. • Increasing the SDS1 trip response. - Abstract: For CANDU reactors, several options to improve CANDU nuclear power plant operation safety margin have been investigated in this paper. A particular attention is paid to the response time of CANDU shutdown system number 1 (SDS1) in case of large break loss of coolant accident (LLOCA). Based on point kinetic method, a systematic fundamental analysis is performed to CANDU LLOCA event, and the power transient signal is generated. In order to improve the SDS1 response time during LLOCA events, an innovative power measurement and signal processing system is particularly designed. The new signal processing system is implemented with the input of the LLOCA power transient, and the simulation results of the reactor trip time and signal are compared to those of the existing system in CANDU power plants. It is demonstrated that the new signal processing system can not only achieve a shorter reactor trip time than the existing system, but also accommodate the spurious trip immunity. This will significantly enhance the safety margin for the power plant operation, or bring extra economical benefits to the power plant units.

  6. Fatigue damage of steam turbine shaft at asynchronous connections of turbine generator to electrical network

    Science.gov (United States)

    Bovsunovsky, A. P.

    2015-07-01

    The investigations of cracks growth in the fractured turbine rotors point out at theirs fatigue nature. The main reason of turbine shafts fatigue damage is theirs periodical startups which are typical for steam turbines. Each startup of a turbine is accompanied by the connection of turbine generator to electrical network. During the connection because of the phase shift between the vector of electromotive force of turbine generator and the vector of supply-line voltage the short-term but powerful reactive shaft torque arises. This torque causes torsional vibrations and fatigue damage of turbine shafts of different intensity. Based on the 3D finite element model of turbine shaft of the steam turbine K-200-130 and the mechanical properties of rotor steel there was estimated the fatigue damage of the shaft at its torsional vibrations arising as a result of connection of turbine generator to electric network.

  7. Direct steam generation (DSG) solar thermal power plant in Thailand

    Energy Technology Data Exchange (ETDEWEB)

    Sukchai, Sukruedee; Chramsa-ard, Wisut; Sonsaree, Sorawit; Boonsu, Rungrudee [Naresuan Univ., Phitsanulok (Thailand). School of Renewable Energy Technology; Krueger, Joachim; Pandian, Yuvaraj [Solarlite GmbH, Duckwitz (Germany)

    2012-07-01

    In 2010, the total electricity consumption in Thailand was 149,301 GWh, increased by 10.5% compared with that in the previous year. The economic sector accounting for the highest share of national electricity consumption was the industrial sector, holding a share of 46%; while the household and commercial sectors accounted for a share of 22% and 15% respectively. The electricity is generated from natural gas, coal, oil, hydro, import and other of 72%, 18%, 0.4%, 3%, 4%, and 2% respectively. In the past, the Electricity Generating Authority of Thailand (EGAT) was the sole power producer. Later, the government had formulated a policy promoting the private sector role in the power generation sector in order to encourage competition in the generation business. Currently, it is resulting in a growing number of Very Small Power Producers (VSPP), using renewable energy as main fuel, supplying power to the grid. In this presentation, general background and situation of solar thermal power plant (DSG) in Thailand will be presented. The resource potential which presented by solar map for the central, north and northeast parts of the country is quite clear sky that receive the highest direct normal irradiation of 1,350 - 1,400 kWh/m{sup 2}-year stand for 43% of the total areas of the country. Together with the high direct normal irradiation is received during summer from January to April about 14-17 MJ/m{sup 2}-day. The first of solar thermal power plant in Thailand is presented. Solar energy development that is one of renewable energy promotion program in the nation master plan has been reviewed and discussed to indicate the recommendation. Barriers as educational, technical and financial to promote solar thermal power plant is also presented. From the investigation, this presentation proposes some idea to be the guideline for policy setting, overcome the solar thermal power plant barrier in Thailand. (orig.)

  8. Aerosol trapping in steam generator (artist): an investigation of aerosol and iodine behaviour in the secondary side of a steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Guentay, S.; Birchley, J.; Suckow, D.; Dehbi, A

    2000-07-01

    Incidents such as a steam generator tube rupture (SGTR) with stuck-open relief valve are important accident sequences for analysis by virtue of the open path for release of radioactivity which ensues. The release may be mitigated by deposition of fission products on the steam generator (SG) tubes and other structures, or by scrubbing in the secondary coolant. The absence of empirical data, the complexity of the geometry and controllingprocesses, however, make the retention difficult to quantify and its full import is typically not taken into account in risk assessment studies. The ARTIST experimental programme at PSI will simulate the flow and retention of aerosol-borne fission products in the SG secondary, and thus provide a unique database to support safety assessments and analytical models. Scaling of the break flow represents a particular challenge since the aerosol retention processes operate at contrasting length scales. Preliminary calculations have identified a baseline set of conditions, and confirmed the feasibility of the rig design and scaling principles. Flexibility of the rig layout enables simulations to be performed for a range of SG designs, accident situations and accident management philosophies. (authors)

  9. LBB in Candu plants

    Energy Technology Data Exchange (ETDEWEB)

    Kozluk, M.J.; Vijay, D.K. [Ontario Hydro Nuclear, Toronto, Ontario (Canada)

    1997-04-01

    Postulated catastrophic rupture of high-energy piping systems is the fundamental criterion used for the safety design basis of both light and heavy water nuclear generating stations. Historically, the criterion has been applied by assuming a nonmechanistic instantaneous double-ended guillotine rupture of the largest diameter pipes inside of containment. Nonmechanistic, meaning that the assumption of an instantaneous guillotine rupture has not been based on stresses in the pipe, failure mechanisms, toughness of the piping material, nor the dynamics of the ruptured pipe ends as they separate. This postulated instantaneous double-ended guillotine rupture of a pipe was a convenient simplifying assumption that resulted in a conservative accident scenario. This conservative accident scenario has now become entrenched as the design basis accident for: containment design, shutdown system design, emergency fuel cooling systems design, and to establish environmental qualification temperature and pressure conditions. The requirement to address dynamic effects associated with the postulated pipe rupture subsequently evolved. The dynamic effects include: potential missiles, pipe whipping, blowdown jets, and thermal-hydraulic transients. Recent advances in fracture mechanics research have demonstrated that certain pipes under specific conditions cannot crack in ways that result in an instantaneous guillotine rupture. Canadian utilities are now using mechanistic fracture mechanics and leak-before-break assessments on a case-by-case basis, in limited applications, to support licensing cases which seek exemption from the need to consider the various dynamic effects associated with postulated instantaneous catastrophic rupture of high-energy piping systems inside and outside of containment.

  10. Verification tests for CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1997-07-01

    For the development of a CANDU advanced fuel, the CANFLEX-NU fuel bundles were tested under reactor operating conditions at the CANDU-Hot test loop. This report describes test results and test methods in the performance verification tests for the CANFLEX-NU bundle design. The main items described in the report are as follows. - Fuel bundle cross-flow test - Endurance fretting/vibration test - Freon CHF test - Production of technical document. (author). 25 refs., 45 tabs., 46 figs.

  11. The development of a control system for a small high speed steam microturbine generator system

    Science.gov (United States)

    Alford, A.; Nichol, P.; Saunders, M.; Frisby, B.

    2015-08-01

    Steam is a widely used energy source. In many situations steam is generated at high pressures and then reduced in pressure through control valves before reaching point of use. An opportunity was identified to convert some of the energy at the point of pressure reduction into electricity. To take advantage of a market identified for small scale systems, a microturbine generator was designed based on a small high speed turbo machine. This machine was packaged with the necessary control valves and systems to allow connection of the machine to the grid. Traditional machines vary the speed of the generator to match the grid frequency. This was not possible due to the high speed of this machine. The characteristics of the rotating unit had to be understood to allow a control that allowed export of energy at the right frequency to the grid under the widest possible range of steam conditions. A further goal of the control system was to maximise the efficiency of generation under all conditions. A further complication was to provide adequate protection for the rotating unit in the event of the loss of connection to the grid. The system to meet these challenges is outlined with the solutions employed and tested for this application.

  12. Control scheme for direct steam generation in parabolic troughs under recirculation operation mode

    Energy Technology Data Exchange (ETDEWEB)

    Valenzuela, L.; Zarza, E. [CIEMAT, Plataforma Solar de Almeria, Ctra. Senes s/n, P.O. Box 22, E-04200 Tabernas, Almeria (Spain); Berenguel, M. [Universidad de Almeria, Dpto. Lenguajes y Computacion, Ctra. Sacramento s/n, E-04120 Almeria (Spain); Camacho, E.F. [Universidad de Sevilla, Dpto. de Ingenieria de Sistemas y Automatica, Camino de los Descubrimientos s/n, E-41092 Sevilla (Spain)

    2006-01-15

    Electricity production using solar thermal energy is one of the main research areas at present in the field of renewable energies, these systems being characterised by the need of reliable control systems aimed at maintaining desired operating conditions in the face of changes in solar radiation, which is the main source of energy. A new prototype of solar system with parabolic trough collectors was implemented at the Plataforma Solar de Almeria (PSA, South-East Spain) to investigate the direct steam generation process under real solar conditions in the parabolic solar collector field of a thermal power plant prototype. This paper presents details and some results of the application of a control scheme designed and tested for the recirculation operation mode, for which the main objective is to obtain steam at constant temperature and pressure at the outlet of the solar field, so that changes produced in the inlet water conditions and/or solar radiation will only affect the amount of steam produced by the solar field. The steam quality and consequently the nominal efficiency of the plant are thus maintained. (author)

  13. Prediction of structural integrity of steam generator tubes under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S. [Argonne National Lab., IL (United States)

    1999-11-01

    Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating and design-basis accident conditions are reviewed. These rate-independent flow stress models are inadequate for predicting failure of steam generator tubes under severe accident conditions because the temperature of the tubes during such accidents can reach as high as 800 C where creep effects become important. Therefore, a creep rupture model for predicting failure was developed and validated by tests on unflawed and flawed specimens containing axial and circumferential flaws and loaded by constant as well as ramped temperature and pressure loadings. Finally, tests were conducted using pressure and temperature histories that are calculated to occur during postulated severe accidents. In all cases, the creep rupture model predicted the failure temperature and time more accurately than the flow stress models. (orig.)

  14. Model-based process management and process optimisation of industrial steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Kaestner, Wolfgang; Hampel, Rainer; Foerster, Tom; Freund, Matthias [Univ. of Applied Sciences Zittau/Goerlitz, Zittau (Germany). IPM; Haake, Dietmar; Kanisch, Heiko [Vattenfall Europe Generation AG, Cottbus (Germany); Altmann, Ulrich-Steffen; Mueller, Frank [CombTec GmbH, Zittau (Germany)

    2010-07-01

    Within the framework of a joint project sponsored by BMBF the development of algorithms for the optimal process management of industrial steam generators was carried out. The research work was focussed on the reduction of high-temperature corrosion at the membrane walls of lignite-fired steam generators. In the result a system for demonstration of technology was installed successfully in a Vattenfall power plant in 2009. During the operation of the power plant the capability of the system to guarantee a stable and symmetric fire position was demonstrated. Current work is focussed on the further development and retrofitting of the system to apply it under real condition in a Vattenfall power plant. (orig.)

  15. Thermo-economic study on the implementation of steam turbine concepts for flexible operation on a direct steam generation solar tower power plant

    Science.gov (United States)

    Topel, Monika; Ellakany, Farid; Guédez, Rafael; Genrup, Magnus; Laumert, Björn

    2016-05-01

    Among concentrating solar power technologies, direct steam generation solar tower power plants represent a promising option. These systems eliminate the usage of heat transfer fluids allowing for the power block to be run at greater operating temperatures and therefore further increasing the thermal efficiency of the power cycle. On the other hand, the current state of the art of these systems does not comprise thermal energy storage as there are no currently available and techno-economically feasible storage integration options. This situation makes direct steam generation configurations even more susceptible to the already existing variability of operating conditions due to the fluctuation of the solar supply. In the interest of improving the annual performance and competitiveness of direct steam generation solar tower systems, the present study examines the influence of implementing two flexibility enhancing concepts which control the steam flow to the turbine as a function of the incoming solar irradiation. The proposed concepts were implemented in a reference plant model previously developed by the authors. Then, a multi-objective optimization was carried out in order to understand which configurations of the steam turbine concepts yield reductions of the levelized cost of electricity at a lower investment costs when compared to the reference model. Results show that the implementation of the proposed strategies can enhance the thermo-economic performance of direct steam generation systems by yielding a reduction of up to 9.2% on the levelized cost of electricity, mainly due to allowing 20% increase in the capacity factor, while increasing the investment costs by 7.8%.

  16. Development of laser repair technique for cutting impurities in secondary side of a steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Cheol Jung; Chung, Chin Man; Kim, Min Suk; Jeong, Tae Moon

    1999-12-01

    In this research, the laser repair technique was investigated for the purpose of cutting impurities in secondary side of a steam generator. For this research, a high quality Nd:YAG laser was manufactured and the beam delivery experiments was performed with multimode optical fibers. Also, the small size focusing system was designed and fabricated for remote cutting and the laser cutting experiment was performed. (author)

  17. Minimising hydrogen sulphide generation during steam assisted production of heavy oil

    OpenAIRE

    Wren Montgomery; Sephton, Mark A.; Watson, Jonathan S.; Huang Zeng; Andrew C. Rees

    2015-01-01

    The majority of global petroleum is in the form of highly viscous heavy oil. Traditionally heavy oil in sands at shallow depths is accessed by large scale mining activities. Recently steam has been used to allow heavy oil extraction with greatly reduced surface disturbance. However, in situ thermal recovery processes can generate hydrogen sulphide, high levels of which are toxic to humans and corrosive to equipment. Avoiding hydrogen sulphide production is the best possible mitigation strateg...

  18. Reducing And Analysizing of Flow Accelerated Corrosion at Thermal Power Plant, Heat Recovery Steam Generators

    OpenAIRE

    Akın Avşaroğlu; Suphi URAL

    2017-01-01

    The purpose of this study is to Reducing and Analysing of Flow Accelerated Corrosion in Thermal Plant Heat Recovery Steam Generators. All these studies have been performed in a new and 16 year-old established Combined Cycle Power Plants in Turkey. Corrosion cases have been investigated due to Mechanical Outage Reports at Power Plant in 2011-2015. Flow Accelerated Corrosion study has been based on specific zone related with Economizer Low Pressure connection pipings. It was issued a performanc...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Coolant stratification and its thermohydrodynamic specificity under natural circulation in horizontal steam generator collectors

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A.; Leontieva, V.; Mitriukhin, A. [Saint-Petersburg Technical Univ. (Russian Federation)

    1997-12-31

    The experiments and the test facilities for the study of the stratification phenomenon in the hot plenum of reactor and the upper parts of the steam generator collectors in a nuclear power plant are described. The aim of the experiments was to define the conditions of the stratification initiation, to study the temperature field in the upper part, the definition of the characteristics in the stratification layer, and also to study the factors which cause the intensity of the stagnant volume cooling.

  1. Self-assembly of highly efficient, broadband plasmonic absorbers for solar steam generation

    OpenAIRE

    Lin ZHOU; Tan, Yingling; Ji, Dengxin; Zhu, Bin; Zhang, Pei; Xu, Jun; Gan, Qiaoqiang; Yu, Zongfu; Zhu, Jia

    2016-01-01

    The study of ideal absorbers, which can efficiently absorb light over a broad range of wavelengths, is of fundamental importance, as well as critical for many applications from solar steam generation and thermophotovoltaics to light/thermal detectors. As a result of recent advances in plasmonics, plasmonic absorbers have attracted a lot of attention. However, the performance and scalability of these absorbers, predominantly fabricated by the top-down approach, need to be further improved to e...

  2. Mathematical simulation of processes in horizontal steam generator and the program of calculation of its characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Titov, V.F.; Zorin, V.M.; Gorburov, V.I. [OKB Gidropress, Moscow Energy Inst. (Russian Federation)

    1995-12-31

    On the basis of mathematical models describing the processes in horizontal steam generator (SG) the code giving the possibility to calculate the hydrodynamical characteristics in any point of water volume, has been developed. The code simulates the processes in SG in the stationary (or quasi-stationary) mode or operation only. The code may be used as a next step to calculations of the SG characteristics in the non-stationary modes of operation.

  3. Multi-element eddy current probe. For inspecting steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Savin, E.; Sartre, B. [FRAMATOME, 92 - Paris-La-Defense (France); Placko, D.; Premel, D. [Ecole Nationale Superieure de Cachan, 94 (France)

    2000-10-01

    Framatome and the Ecole Normale Superieure de Cachan are developing a multi-element eddy current probe for inspecting steam generator tubes of 900 MWe PWR reactors. The device is intended to replace much slower rotating probes. Using its measurements, the conductivity image of any point in the tube can be reconstructed, thanks to a numerical, thanks to a numerical model, thus allowing diagnosis. The first trial results on mockups seem already competitive with those obtained using a rotary probe. (authors)

  4. Validation of the THIRST steam generator thermalhydraulic code against the CLOTAIRE phase II experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Pietralik, J.M.; Campagna, A.O.; Frisina, V.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1998-07-01

    Steam generator thermalhydraulic codes are used frequently to calculate both global and local parameters inside the steam generator. The former include heat transfer output, recirculation ratio, outlet temperatures, and pressure drops for operating and abnormal conditions. The latter are used in further analyses of flow-induced vibration, fretting wear, sludge deposition, and flow accelerated corrosion. For these purposes, detailed, three-dimensional two-phase flow and heat transfer parameters are needed. To make the predictions more accurate and reliable, the codes need to be validated in geometries representative of real conditions. One such study is an international cooperative experimental program called CLOTAIRE based in France. COG participated in the first two phases of the program; the results of the validation of Phase 1 were presented at the 1994 Steam Generator and Heat Exchanger Conference, and the results of the validation of Phase II are the subject of this paper. THIRST is a thermalhydraulic, finite volume code to predict the flow and heat transfer in steam generators. The local results of CLOTAIRE Phase II have been used to validate the code. These consist of the measurements of void fraction and axial gas-phase velocity in the U-bend region. The measurements were done using bi-optical probes. A comparison of global results indicates that the THIRST predictions, with the Chisholm void fraction model, are within 2 to 3% of the experimental results. Using THIRST with the homogeneous void fraction model, the global results were less accurate but still well predicted with the greatest error of 10% for the separator pressure drop. Comparisons of the local predictions for void fraction and axial gas-phase show good agreement. The Chisholm void fraction model generally gives better agreement with the experimental data while the homogeneous model tends to overpredict the void fraction and underpredict the gas velocity. (author)

  5. The relative impact of sizing errors on steam generator tube failure probability

    Energy Technology Data Exchange (ETDEWEB)

    Cizelj, L.; Dvorsek, T. [Jozef Stefan Inst., Ljubljana (Slovenia)

    1998-07-01

    The Outside Diameter Stress Corrosion Cracking (ODSCC) at tube support plates is currently the major degradation mechanism affecting the steam generator tubes made of Inconel 600. This caused development and licensing of degradation specific maintenance approaches, which addressed two main failure modes of the degraded piping: tube rupture; and excessive leakage through degraded tubes. A methodology aiming at assessing the efficiency of a given set of possible maintenance approaches has already been proposed by the authors. It pointed out better performance of the degradation specific over generic approaches in (1) lower probability of single and multiple steam generator tube rupture (SGTR), (2) lower estimated accidental leak rates and (3) less tubes plugged. A sensitivity analysis was also performed pointing out the relative contributions of uncertain input parameters to the tube rupture probabilities. The dominant contribution was assigned to the uncertainties inherent to the regression models used to correlate the defect size and tube burst pressure. The uncertainties, which can be estimated from the in-service inspections, are further analysed in this paper. The defect growth was found to have significant and to some extent unrealistic impact on the probability of single tube rupture. Since the defect growth estimates were based on the past inspection records they strongly depend on the sizing errors. Therefore, an attempt was made to filter out the sizing errors and to arrive at more realistic estimates of the defect growth. The impact of different assumptions regarding sizing errors on the tube rupture probability was studied using a realistic numerical example. The data used is obtained from a series of inspection results from Krsko NPP with 2 Westinghouse D-4 steam generators. The results obtained are considered useful in safety assessment and maintenance of affected steam generators. (author)

  6. Steam generators, industrial power plants, and cogeneration plants. Lectures; Dampferzeuger, Industrie- und Heizkraftwerke 2010. Vortraege

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-01

    The proceedings of the meeting on steam generators, industrial power plants, and cogeneration plants include the following lectures: Assignments and scopes of the VGB working group ''industry and thermal power plants, BHKW''. Combustion power and process control - application of feed grate firing in industrial technology - experiences and perspectives. A new possibility of biomass co-combustion. Biomass co-combustion in Vattenfall Waerme AG. Biomass plant with optimized control. The new energy supply concept for the paper plant in Plattling. Efficient steam boiler facilities for industry and thermal power plants - case studies. Adaptation of flue gas purification for co-combustion with experiences of prototype plants. Modern risk and insurance management for power plants. Reliability oriented maintenance. Surrogate fuel IHKW Gersthofen - planning, construction and preliminary operational experiences. Assignments and scopes of the VGB working group ''steam generators''. New developments in process safety management of E.ON UK coal-fired power plants. Station-supply reduction by power drive reconstruction to frequency control - modern injection technology at high plant parameters. Self-optimizing control of fuel/air regulation. CO{sub 2} reduction by automatic power plant modeling. Virtual reality as QA tool in 3D planning. Thermodynamic studies in power plants using VDI 2048. Heating surface cleaning with explosion generators - an alternative to soot blowers. results of laboratory study on urgent material questions.

  7. Guidelines for random excitation forces due to cross flow in steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, C.E.; Pettigrew, M.J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1998-07-01

    Random excitation forces can cause low-amplitude tube motion that will result in long-term fretting-wear or fatigue. To prevent these tube failures in steam generators and other heat exchangers, designers and trouble-shooters must have guidelines that incorporate random or turbulent fluid forces. Experiments designed to measure fluid forces have been carried out at Chalk River Laboratories and at other labs around the world. The data from these experiments have been studied and collated to determine suitable guidelines for random excitation forces. In this paper, a guideline for random excitation forces in single-phase cross flow is presented in the form of normalised spectra that are applicable to a wide range of flow conditions and tube frequencies. In particular, the experimental results used in this study were carried out over the full range of flow conditions found in a nuclear steam generator. The proposed guidelines are applicable to steam generators, condensers, reheaters and other shell-and-tube heat exchangers. They may be used for flow-induced vibration analysis of new or existing components, as input to vibration analysis computer codes and as specifications in procurement documents. (author)

  8. Thermophoretic and turbulent deposition of graphite dust in HTGR steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei, E-mail: pengwei@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Zhang, Tianqi; Sun, Xiaokai [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2016-04-15

    Graphical abstract: Most of small graphite dust particles are deposited in steam generators at low reactor power levels, mostly small and large particles are deposited with few medium-size particles at high reactor power levels. - Highlights: • Thermophoretic deposition is the main mechanism for small particles. • Turbulent deposition is dominant for large particles. • Mostly small particles were deposited at low reactor power. • Both small and large particles are deposited at high reactor power. - Abstract: The present study calculated the graphite dust deposition in the steam generator of HTGR by using a thermophoretic deposition model and a turbulent deposition model based on the temperature and flow field distributions calculated by a computational fluid dynamics (CFD) model. The results showed that the heat flux along the heat transfer surface was not evenly distributed which affect the particle deposition. Thermophoretic deposition is the main factor causing small graphite dust particles to deposit on the surface while turbulent deposition plays the dominant role for large particle deposition. The results also indicate that the deposited particles are mainly small graphite dust particles in steam generator when the reactor is running at low power, with both small and large graphite dust particles at high reactor power levels with fewer medium size particles.

  9. Two-Phase Instability Characteristics of Printed Circuit Steam Generator for the Low Pressure Condition

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Han-Ok; Han, Hun Sik; Kim, Young-In; Kim, Keung Koo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Reduction of installation space for steam generators can lead to much smaller reactor vessel with resultant decrease of overall manufacturing cost for the components. A PCHE(Printed Circuit Heat Exchanger) is one of the compact types of heat exchangers available as an alternative to conventional shell and tube heat exchangers. Its name is derived from the procedure used to manufacture the flat metal plates that form the core of the heat exchanger, which is done by chemical milling. These plates are then stacked and diffusion bonded, converting the plates into a solid metal block containing precisely engineered fluid flow passages. PCSG(Printed Circuit Steam Generator) is a potential candidate to be applied to the integral reactor with its compactness and mechanical robustness. For the introduction of new steam generator, design requirement for the two-phase flow instability should be considered. This paper describes two-phase flow instability characteristics of PCSG for the low pressure condition. PCSG is a potential candidate to be applied to the integral reactor with its compactness and mechanical robustness. Interconnecting flow path was developed to mitigate the two-phase flow instability in the cold side. The flow characteristics of two-phase flow instability at the PCSG is examined experimentally in this study.

  10. Correlation of Steam Generator Mixing Parameters for Severe Accident Hot-Leg Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Liao, Yehong; Guentay, Salih [Paul Scherrer Institut, Villigen PSI, CH-5232 (Switzerland)

    2008-07-01

    Steam generator inlet plenum mixing phenomenon with hot-leg counter-current natural circulation during a PWR station blackout severe accident is one of the important processes governing which component will fail first as a result of thermal challenge from the circulating gas with high temperature and pressure. Since steam generator tube failure represents bypass release of fission product from the reactor to environment, study of inlet plenum mixing parameters is important to risk analysis. Probability distribution functions of individual mixing parameter should be obtained from experiments or calculated by analysis. In order to perform sensitivity studies of the synergetic effects of all mixing parameters on the severe accident-induced steam generator tube failure, the distribution and correlation of these mixing parameters must be known to remove undue conservatism in thermal-hydraulic calculations. This paper discusses physical laws governing three mixing parameters in a steady state and setups the correlation among these mixing parameters. The correlation is then applied to obtain the distribution of one of the mixing parameters that has not been given in the previous CFD analysis. Using the distributions and considering the inter-dependence of the three mixing parameters, three sensitivity cases enveloping the mixing parameter uncertainties are recommended for the plant analysis. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, J. [Associacion Nuclear Asco AIE, Barcelona (Spain); 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.

  12. Influence of nonuniformity of the submerged perforated sheet on steam demand leveling on the evaporation surface of a VVER steam generator

    Science.gov (United States)

    Blinkov, V. N.; Elkin, I. V.; Emelianov, D. A.; Melikhov, V. I.; Melikhov, O. I.; Nerovnov, A. A.; Nikonov, S. M.; Parfenov, Y. V.

    2016-01-01

    The results of a calculation and experimental research of the influence of nonuniformity of the submerged perforated sheet on steam demand leveling on the evaporation surface are published in the current article. A short description of the PGV test facility and a measuring system whose test section is a transverse "cut" of an actual PGV-1000 steam generator with the internals is presented. The methods of experimental starts are explained and instrumentations are described. A uniformly perforated sheet with the flow section of 5.7% and a nonuniformly perforated sheet with the flow section of 4.3% on the cold half and 8.1% on the hot half were used in the experiments. The system pressure was approximately 7 MPa, the inlet steam flow rate was varied between 4.23 and 7.94 t/h, i.e., the steam velocity on the evaporation surface was 0.15-0.29 m/s. The experimental results were analyzed with (1) the engineering method based on estimating the flow rates of steam on hot and cold half by the experimental values of the pressure drop on submerged list and (2) the STEG code, which was developed for three-dimensional mathematical modeling of the two-phase thermohydraulics in the heat exchanger volume and upgraded. It was established that changing the perforation from a uniform to a nonuniform one increases the residual nonuniformity coefficient, which characterizes the flow of steam from the hot side to the cold side under the sheet. However, the steam separation becomes worse because of a high local residual nonuniformity coefficients near the border of two plates with different perforation levels.

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

    Institute of Scientific and Technical Information of China (English)

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

    2008-01-01

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

  14. Simulation analysis of static and dynamic characteristics of once-through steam generator in concentric annuli tube

    Institute of Scientific and Technical Information of China (English)

    ZHANG Wei; BIAN Xin-qian; XIA Guo-qing

    2006-01-01

    The once-through steam generator (OTSG) in concentric annuli tube is a new type of steam generator which applies double side to transfer heat. The heat flux between the water of centric tube, outside annuli tube and that of annulus channel is assumed to be equal, and then the steam generator's model is built by lumped parameters with moving boundary. In the basis of the built model, static and dynamic characteristics are analyzed.The static characteristics are proved by experiment results in a 19-tube once-through steam generator of Babcock & Wilcox. The characteristics that the lengths of three regions (subcooled region, nucleate boiling region, superheat region) change with power can be explained by theory analysis. The dynamic characteristics accord with the heat and hydraulics and the results of analysis according to the mechanism.

  15. Numerical Study of Thermal Hydraulics for Secondary side of Steam Generator by CUPID/MARS Coupled Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Ryong; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As a thermal-hydraulic behavior in the secondary side of steam generator such as two-phase boiling flow, flow-induce vibration of U-tubes is quite complicated, the importance to numerically investigate the flow behavior has been arisen. Recently, multi-scale analyses have been developed to take into account the primary side as well. In this study, the coupled CUPID and MARS code was used for the simulation of boiler side of the PWR steam generator. Calculation results are compared with the existing code quantitatively. Coupled CUPID/MARS code was applied for the simulation of the steam generator. The primary side of the steam generator and other RCS was simulated by MARS and the secondary side was calculated by CUPID with porous media approach.

  16. System Identification of a Nonlinear Multivariable Steam Generator Power Plant Using Time Delay and Wavelet Neural Networks

    Directory of Open Access Journals (Sweden)

    Laila Khalilzadeh Ganjali-khani

    2013-01-01

    Full Text Available One of the most effective strategies for steam generator efficiency enhancement is to improve the control system. For such an improvement, it is essential to have an accurate model for the steam generator of power plant. In this paper, an industrial steam generator is considered as a nonlinear multivariable system for identification. An important step in nonlinear system identification is the development of a nonlinear model. In recent years, artificial neural networks have been successfully used for identification of nonlinear systems in many researches. Wavelet neural networks (WNNs also are used as a powerful tool for nonlinear system identification. In this paper we present a time delay neural network model and a WNN model in order to identify an industrial steam generator. Simulation results show the effectiveness of the proposed models in the system identification and demonstrate that the WNN model is more precise to estimate the plant outputs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Cuvelliez, Ch.; Roussel, G. [AIB-Vincotte Nuclear, Brussels (Belgium)

    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.

  18. Novel metallic alloys as phase change materials for heat storage in direct steam generation applications

    Science.gov (United States)

    Nieto-Maestre, J.; Iparraguirre-Torres, I.; Velasco, Z. Amondarain; Kaltzakorta, I.; Zubieta, M. Merchan

    2016-05-01

    Concentrating Solar Power (CSP) is one of the key electricity production renewable energy technologies with a clear distinguishing advantage: the possibility to store the heat generated during the sunny periods, turning it into a dispatchable technology. Current CSP Plants use an intermediate Heat Transfer Fluid (HTF), thermal oil or inorganic salt, to transfer heat from the Solar Field (SF) either to the heat exchanger (HX) unit to produce high pressure steam that can be leaded to a turbine for electricity production, or to the Thermal Energy Storage (TES) system. In recent years, a novel CSP technology is attracting great interest: Direct Steam Generation (DSG). The direct use of water/steam as HTF would lead to lower investment costs for CSP Plants by the suppression of the HX unit. Moreover, water is more environmentally friendly than thermal oils or salts, not flammable and compatible with container materials (pipes, tanks). However, this technology also has some important challenges, being one of the major the need for optimized TES systems. In DSG, from the exergy point of view, optimized TES systems based on two sensible heat TES systems (for preheating of water and superheating vapour) and a latent heat TES system for the evaporation of water (around the 70% of energy) is the preferred solution. This concept has been extensively tested [1, 2, 3] using mainly NaNO3 as latent heat storage medium. Its interesting melting temperature (Tm) of 306°C, considering a driving temperature difference of 10°C, means TES charging steam conditions of 107 bar at 316°C and discharging conditions of 81bar at 296°C. The average value for the heat of fusion (ΔHf) of NaNO3 from literature data is 178 J/g [4]. The main disadvantage of inorganic salts is their very low thermal conductivity (0.5 W/m.K) requiring sophisticated heat exchanging designs. The use of high thermal conductivity eutectic metal alloys has been recently proposed [5, 6, 7] as a feasible alternative. Tms

  19. Solar tower power plant using a particle-heated steam generator: Modeling and parametric study

    Science.gov (United States)

    Krüger, Michael; Bartsch, Philipp; Pointner, Harald; Zunft, Stefan

    2016-05-01

    Within the framework of the project HiTExStor II, a system model for the entire power plant consisting of volumetric air receiver, air-sand heat exchanger, sand storage system, steam generator and water-steam cycle was implemented in software "Ebsilon Professional". As a steam generator, the two technologies fluidized bed cooler and moving bed heat exchangers were considered. Physical models for the non-conventional power plant components as air- sand heat exchanger, fluidized bed coolers and moving bed heat exchanger had to be created and implemented in the simulation environment. Using the simulation model for the power plant, the individual components and subassemblies have been designed and the operating parameters were optimized in extensive parametric studies in terms of the essential degrees of freedom. The annual net electricity output for different systems was determined in annual performance calculations at a selected location (Huelva, Spain) using the optimized values for the studied parameters. The solution with moderate regenerative feed water heating has been found the most advantageous. Furthermore, the system with moving bed heat exchanger prevails over the system with fluidized bed cooler due to a 6 % higher net electricity yield.

  20. Design and construction of a steam generator with feedback; Projeto e construcao de um gerador de vapor com realimentacao

    Energy Technology Data Exchange (ETDEWEB)

    Camargo, Camila C., E-mail: camilacamargo@outlook.com [Universidade Federal de Sao Paulo (UNIFESP), Sao Jose dos Campos, SP (Brazil); Placco, Guilherme M., E-mail: placco@ieav.cta.br [Instituto de Tecnologia Aeronautica (ITA/CTA), Sao Jose dos Campos, SP (Brazil); Guimaraes, Lamartine N.F., E-mail: guimarae@ieav.cta.br [Instituto de Estudos Avancado (IEAv/DCTA), Sao Jose dos Campos, SP (Brazil). Departamento ENU

    2013-07-01

    The EARTH project aims to develop technologies to design and build systems that generate electricity in space, using microreactors. One of the activities within the TERRA project aims to build a closed thermal cycle Rankine type in order to test a Tesla turbine type. The objective of this work is to design and build a steam generator with feedback, which should ensure a satisfactory range of steam supply, security system, feedback system and heating system.

  1. About technical possibility to use VEERA facility for investigation of coolant stratification phenomenon in horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Mitioukov, V.; Mitrioukhine, A. [St. Petersburg State Technical Univ. (Russian Federation); Korteniemi, V. [Lappeenranta Univ. of Technology (Finland)

    1997-12-31

    The presentation gives a brief insight on possibility of using the VEERA facility in studying the stratification phenomenon. The idea for such experiments is to use the facility upper plenum part to simulate the conditions in upper part of horizontal steam generator hot collector. The upper part of steam generator hot collector is one of the locations where the stratification can take part during natural circulation mode. 4 refs.

  2. Simple evaluations of fluid-induced vibrations for steam generator tube arrays in advanced marine reactors (MRX, DRX)

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Kazuo [Ishikawajima-Harima Heavy Industries Co., Ltd., Tokyo (Japan); Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-06-01

    Advanced Marine Reactor (MRX) and Deep Sea Research Reactor (DRX) are the integral-type PWR, and the steam generators are installed in the reactor vessels. Steam generators are of the once-through, helical-coil tube types. Heat transfer tubes surround inner shroud in annular space of the reactor vessel. Flow-induced vibrations are calculated by simple methods, and the arrangement of tube support structures are evaluated. (author)

  3. Post-scram Liquid Metal cooled Fast Breeder Reactor (LMFBR) neat transport system dynamics and steam generator control

    Science.gov (United States)

    Brukx, J. F. L. M.

    1982-06-01

    Loop type LMFBR heat transport system dynamics after reactor shutdown and during subsequent decay heat removal are considered with emphasis on steam generator dynamics including the development and evaluation of various post-scram steam generator control systems, and natural circulation of the sodium coolant, including the influence of superimposed free convection on forced convection heat transfer and pressure drop. The normal operating and decay heat removal functions of the overall heat transport system are described.

  4. Structural Evaluation of a PGSFR Steam Generator for a Steady State Condition

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang-Gyu; Kim, Jong-Bum; Kim, Hoe-Woong; Koo, Gyeong-Hoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In this study, design loads for design condition and normal operating steady state condition were classified and the structural analyses for each design loads were carried out. And, structural integrities under each service level were evaluated according to ASME design code. The structural analyses of a steam generator are carried out and its structural integrity under the given service levels is evaluated per ASME Code rule. The design loads according to design condition and normal operating steady condition are classified and stresses calculated from stress analyses are linearized and summarized in their stress components. As a result, the SG structure satisfies with design criteria for both service levels. Though the steam header is designed as a thick hemisphere, its design margin is not so high in spite of just steady state condition. Thus, additional evaluation by considering various operating events will be followed.

  5. DESIGN OF COMBINED CYCLE GENERATION SYSTEM WITH HIGH TEMPERATURE FUEL CELL AND STEAM TURBINE

    Institute of Scientific and Technical Information of China (English)

    Yu Lijun; Yuan Junqi; Cao Guangyi

    2003-01-01

    For environment protection and high efficiency, development of new concept power plant has been required in China. The fuel cell is expected to be used in a power plant as a centralized power station or distributed power plant. It is a chemical power generation device that converts the energy of a chemical reaction directly into electrical energy and not limited by Carnot cycle efficiency. The molten carbonate fuel cell (MCFC) power plant has several attractive features I.e. High efficiency and lower emission of Nox and Sox. A combined cycle generation system with MCFC and steam turbine is designed. Its net electrical efficiency LHV is about 55%.

  6. Environment Canada's codes of practice and emission guidelines for steam electric power generation

    Energy Technology Data Exchange (ETDEWEB)

    Doiron, C.; Ross, G.

    1992-01-01

    Environment Canada is nearing completion on three significant initiatives undertaken in consultation with, among others, the Canadian Electrical Association. These are the Operations Phase and Decommissioning Phase Environmental Codes of Practice for Steam Electric Power Generation, and the revision of the national emission guidelines for thermal power generation. An overview of each initiative is presented with emphasis on the consultative process used in its development. The nature and content of the codes and the major revisions to the guidelines are outlined, and the success of the consultation process is commented on. A summary is provided of recommendations concerning the Operations Phase Code of Practice.

  7. Boiler and steam generator corrosion. (Latest citations from the NTIS bibliographic database). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-11-01

    The bibliography contains citations concerning corrosion effects, mechanisms, detection, and inhibition in fossil fuel fired boilers and nuclear powered steam generators. Corrosion studies performed on the water side and hot gas side of heat exchanger tubes and support structures are presented. Water treatment, chemical cleaning, and descaling methods are considered. Although emphasis is placed on large-scale power generation systems, residential and commercial heating systems are also discussed. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  8. An Expert System Using A Neural Network For Steam Generator Tube Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kilyoo; Huh, Younghwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Woo, Heegon; Choi, Sungsoo [Korea Electric Power Corporation, Daejeon (Korea, Republic of)

    1991-04-15

    An expert system using neural network is built to automatically evaluate eddy current (EC) signals generated during steam generator (S/G) tubes inspection. The system consists of three subsystem, i.e., syntactic pattern recognition subsystem, neural network subsystem and rule based production subsystem. The syntactic pattern recognition subsystem makes it easy to process the vast EC signal data, screens EC signals and detects event signals such as defect signals and structural signals. The neural network subsystem is useful to classify the event signals which often contain noise signals. The expert system implemented on HP 9000/370 workstation also supplies a good EC test data management function.

  9. Design of nonlinear adaptive steam valve controllers for a turbo-generator system

    Energy Technology Data Exchange (ETDEWEB)

    Bekiaris-Liberis, N.K.; Paraskevopoulos, P.N. [National Technical Univ. of Athens Zographou, Athens (Greece); Boglou, A.K. [Technology Education Inst. of Kavala Agios Loukas, Kavala (Greece); Arvanitis, K.G.; Pasgianos, G.D. [Agricultural Univ. of Athens, Athens (Greece)

    2008-07-01

    This paper reported on a study that investigated the control of power systems consisting of interconnected networks of transmission lines linking generators and loads. Improving both small and large perturbation stability and dynamic performance is important because power systems have become less stable in the past 15 years due to the use of controllers that have been designed on the basis of linearized synchronous generators and turbine models. The high nonlinear nature of power system models and the resulting disturbances render conventional linear controller design techniques obsolete for use in power systems control. Power system engineers are becoming aware of the role of turbine steam valves in improving the dynamic stability of power systems and damping low frequency oscillations. Advanced nonlinear control strategies are needed since the conventional steam valve control theory cannot guarantee transient stability in cases where operational conditions and parameters vary considerably. A design approach to a nonlinear adaptive control system with unknown parameters was developed and applied to the turbine main steam valve control of a power system. A fourth order machine model was used along with an adaptive backstepping method to construct the Lyapunov function in order to obtain a nonlinear adaptive controller to solve the turbine fast valving nonlinear control problem. The newly designed nonlinear adaptive controller can make the resulting adaptive system asymptotically stable. The proposed controller is accompanied by a dynamic estimator of parameters and includes nonlinear damping terms, which guarantee input-output stability even without the use of the adaptive law. Simulation results showed that the proposed nonlinear adaptive controller performs better than other turbine main steam valve control techniques. It can face large parametric uncertainty and results in a closed-loop system that is able to face large and smaller disturbances, providing a

  10. Multi-region fuzzy logic controller with local PID controllers for U-tube steam generator in nuclear power plant

    Directory of Open Access Journals (Sweden)

    Puchalski Bartosz

    2015-12-01

    Full Text Available In the paper, analysis of multi-region fuzzy logic controller with local PID controllers for steam generator of pressurized water reactor (PWR working in wide range of thermal power changes is presented. The U-tube steam generator has a nonlinear dynamics depending on thermal power transferred from coolant of the primary loop of the PWR plant. Control of water level in the steam generator conducted by a traditional PID controller which is designed for nominal power level of the nuclear reactor operates insufficiently well in wide range of operational conditions, especially at the low thermal power level. Thus the steam generator is often controlled manually by operators. Incorrect water level in the steam generator may lead to accidental shutdown of the nuclear reactor and consequently financial losses. In the paper a comparison of proposed multi region fuzzy logic controller and traditional PID controllers designed only for nominal condition is presented. The gains of the local PID controllers have been derived by solving appropriate optimization tasks with the cost function in a form of integrated squared error (ISE criterion. In both cases, a model of steam generator which is readily available in literature was used for control algorithms synthesis purposes. The proposed multi-region fuzzy logic controller and traditional PID controller were subjected to broad-based simulation tests in rapid prototyping software - Matlab/Simulink. These tests proved the advantage of multi-region fuzzy logic controller with local PID controllers over its traditional counterpart.

  11. Application of principal component analysis for the diagnosis of neutron overpower system oscillations in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara; Gabbar, Hossam A., E-mail: Hossam.gabbar@uoit.ca

    2014-04-01

    Highlights: • Diagnosis of neutron overpower protection (NOP) in CANDU reactors. • Accurate reactor detector modeling. • NOP detectors response analysis. • Statistical methods for quantitative analysis of NOP detector behavior. - Abstract: An accurate fault modeling and troubleshooting methodology is required to aid in making risk-informed decisions related to design and operational activities of current and future generation of CANDU{sup ®} designs. This paper attempts to develop an explanation for the unanticipated detector response and overall behavior phenomena using statistical methods to compliment traditional engineering analysis techniques. Principal component analysis (PCA) methodology is used for pattern recognition using a case study of Bruce B zone-control level oscillations.

  12. Sub-atmospheric disk generators for coal-fired MHD/steam combined cycle power plant

    Energy Technology Data Exchange (ETDEWEB)

    Messerle, H.K.; Fang, Y.; Simpson, S.W.; Marty, S.M. (Sydney Univ. (Australia). School of Electrical Engineering)

    1989-01-01

    A coal fired MHD disk generator in a combined cycle MHD/steam power generation system with a diffuser operating at sub-atmospheric pressure is proposed. The effects of pressure on the performance of a radial outflow MHD disk generator and other system components are analysed. Using a previous study as a reference case, preliminary calculations show that, in such a sub-atmospheric system, improved power station efficiency can be achieved. In addition, operation at reduced values of magnetic field strength would be feasible. Calculations have also been carried out for a 30 MW{sub th} experimental disk generator operating at reduced pressure with a magnetic field strength of 2 T. Flow conditions at sub-atmospheric pressure would provide an improved simulation of a full-scale generator operating at normal pressures. (author).

  13. Design Evolution and Verification of the A-3 Chemical Steam Generator

    Science.gov (United States)

    Kirchner, Casey K.

    2009-01-01

    Following is an overview of the Chemical Steam Generator system selected to provide vacuum conditions for a new altitude test facility, the A-3 Test Stand at Stennis Space Center (SSC) in Bay St. Louis, MS. A-3 will serve as NASA s primary facility for altitude testing of the J-2X rocket engine, to be used as the primary propulsion device for the upper stages of the Ares launch vehicles. The Chemical Steam Generators (CSGs) will produce vacuum conditions in the test cell through the production and subsequent supersonic ejection of steam into a diffuser downstream of the J-2X engine nozzle exit. The Chemical Steam Generators chosen have a rich heritage of operation at rocket engine altitude test facilities since the days of the Apollo program and are still in use at NASA White Sands Test Facility (WSTF) in New Mexico. The generators at WSTF have been modified to a degree, but are still very close to the heritage design. The intent for the A-3 implementation is to maintain this heritage design as much as possible, making minimal updates only where necessary to substitute for obsolete parts and to increase reliability. Reliability improvements are especially desired because the proposed system will require 27 generators, which is nine times the largest system installed in the 1960s. Improvements were suggested by the original design firm, Reaction Motors, by NASA SSC and NASA WSTF engineers, and by the A-3 test stand design contractor, Jacobs Technology, Inc. (JTI). This paper describes the range of improvements made to the design to date, starting with the heritage generator and the minor modifications made over time at WSTF, to the modernized configuration which will be used at A-3. The paper will discuss NASA s investment in modifications to SSC s E-2 test facility fire a full-scale Chemical Steam Generator in advance of the larger steam system installation at A-3. Risk mitigation testing will be performed in early 2009 at this test facility to verify that the CSGs

  14. Pre-service baseline inspection using x-probe of Oconee replacement steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Addario, M.; Shipp, P. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada)]. E-mail: pwshipp@babcock.com; Davis, K. [Duke Power, Charlotte, North Carolina (United States); Fogal, C. [R and D Tech, Port Hope, Ontario (Canada)

    2003-07-01

    The eddy current method has been the industry standard for inspecting steam generator tubing for many years and the level of sophistication of coil technology has continued to evolve during that time. State of the art array probe systems now employ multiple sensitivity zones in the probe to better detect and characterize defects in an efficient manner. Owners and regulators of nuclear power plants are interested in the most effective and efficient inspection possible. The ultimate goal has been to meet or exceed new and existing regulatory and design requirements by maximizing the quantity and quality of eddy current data collected while minimizing both the time needed to perform the inspection and the radiation exposure. The X-Probe is an example of this new eddy current array technology. Qualified to detect all types of known defects in steam generator tubing, the technology is comprised of a system of probe, data acquisition instrumentation, computer and human interface software. Recently, Duke Power, along with Babcock and Wilcox Canada and the system developer R/D Tech, collaborated to implement this technology in a first of a kind full scale pre-service inspection of replacement steam generators for Duke Power's Oconee nuclear generating station at Babcock and Wilcox Canada's Cambridge plant. The discussion in this paper will briefly describe the X-Probe technology, describe the system required to perform the inspection, present the general results of the inspection and finally draw some comparative benefit conclusions for both pre-service and in-service applications. (author)

  15. Minimising hydrogen sulphide generation during steam assisted production of heavy oil

    Science.gov (United States)

    Montgomery, Wren; Sephton, Mark A.; Watson, Jonathan S.; Zeng, Huang; Rees, Andrew C.

    2015-02-01

    The majority of global petroleum is in the form of highly viscous heavy oil. Traditionally heavy oil in sands at shallow depths is accessed by large scale mining activities. Recently steam has been used to allow heavy oil extraction with greatly reduced surface disturbance. However, in situ thermal recovery processes can generate hydrogen sulphide, high levels of which are toxic to humans and corrosive to equipment. Avoiding hydrogen sulphide production is the best possible mitigation strategy. Here we use laboratory aquathermolysis to reproduce conditions that may be experienced during thermal extraction. The results indicate that hydrogen sulphide generation occurs within a specific temperature and pressure window and corresponds to chemical and physical changes in the oil. Asphaltenes are identified as the major source of sulphur. Our findings reveal that for high sulphur heavy oils, the generation of hydrogen sulphide during steam assisted thermal recovery is minimal if temperature and pressure are maintained within specific criteria. This strict pressure and temperature dependence of hydrogen sulphide release can allow access to the world's most voluminous oil deposits without generating excessive amounts of this unwanted gas product.

  16. Steam generation under one sun enabled by a floating structure with thermal concentration

    Science.gov (United States)

    Ni, George; Li, Gabriel; Boriskina, Svetlana V.; Li, Hongxia; Yang, Weilin; Zhang, Tiejun; Chen, Gang

    2016-09-01

    Harvesting solar energy as heat has many applications, such as power generation, residential water heating, desalination, distillation and wastewater treatment. However, the solar flux is diffuse, and often requires optical concentration, a costly component, to generate the high temperatures needed for some of these applications. Here we demonstrate a floating solar receiver capable of generating 100 ∘C steam under ambient air conditions without optical concentration. The high temperatures are achieved by using thermal concentration and heat localization, which reduce the convective, conductive and radiative heat losses. This demonstration of a low-cost and scalable solar vapour generator holds the promise of significantly expanding the application domain and reducing the cost of solar thermal systems.

  17. Performance analysis of a Kalina cycle for a central receiver solar thermal power plant with direct steam generation

    DEFF Research Database (Denmark)

    Modi, Anish; Haglind, Fredrik

    2014-01-01

    Solar thermal power plants have attracted increasing interest in the past few years - with respect to both the design of the various plant components, and extending the operation hours by employing different types of storage systems. One approach to improve the overall plant efficiency is to use...... without corroding the equipment by using suitable additives with the mixture. The purpose of the study reported here was to investigate if there is any benefit of using a Kalina cycle for a direct steam generation, central receiver solar thermal power plant with high live steam temperature (450 C...... direct steam generation with water/steam as both the heat transfer fluid in the solar receivers and the cycle working fluid. This enables operating the plant with higher turbine inlet temperatures. Available literature suggests that it is feasible to use ammonia-water mixtures at high temperatures...

  18. Hydrogen generation from 2,2,4-trimethyl pentane reforming over molybdenum carbide at low steam-to-carbon ratios

    Science.gov (United States)

    Cheekatamarla, Praveen K.; Thomson, William J.

    Because of the need for an efficient and inexpensive reforming catalyst, the objective of this work is to determine the feasibility of employing Mo 2C catalyst for the steam reforming and oxy-steam reforming of the higher hydrocarbons typical of transportation fuels such as gasoline. It is shown that bulk Mo 2C catalysts can successfully reform 2,2,4-trimethyl pentane (isooctane) to generate H 2, CO and CO 2 at very low steam/carbon ratios, without coke formation, eliminating the need for pre-reforming. Maximum hydrogen generation was observed at a S/C ratio of 1.3 and 1000 °C during SR reactions and S/C of 0.71, O 2/C of 0.12 at 900 °C during oxidative steam reforming reactions.

  19. Reducing And Analysizing of Flow Accelerated Corrosion at Thermal Power Plant, Heat Recovery Steam Generators

    Directory of Open Access Journals (Sweden)

    Akın Avşaroğlu

    2017-01-01

    Full Text Available The purpose of this study is to Reducing and Analysing of Flow Accelerated Corrosion in Thermal Plant Heat Recovery Steam Generators. All these studies have been performed in a new and 16 year-old established Combined Cycle Power Plants in Turkey. Corrosion cases have been investigated due to Mechanical Outage Reports at Power Plant in 2011-2015. Flow Accelerated Corrosion study has been based on specific zone related with Economizer Low Pressure connection pipings. It was issued a performance report. Results and lessons learnt from these studies will be used as a preventive action manner in all similar Plants.

  20. Residue management and Canada's environmental codes of practice for steam electric power generation

    Energy Technology Data Exchange (ETDEWEB)

    Finlay, P.G.; Stobbs, R.A.; Ross, G.C.; Pinault, P.H.; Doiron, C.C. (Environment Canada, Ottawa, ON (Canada))

    1993-01-01

    Canada's 'Environmental Codes of Practice for Steam Electric Power Generation' are comprehensive environmental protection standards for the various phases of the life cycle of power plants. Codes published include the Siting, Design, Construction, Operation and Decommissioning Phase Codes. Various practices are recommended for the management of air emissions, water, fuels, chemicals, wastewater, and liquid and solid residues. Topics discussed include the production, utilization and disposal of residues, including combustion ashes and desulphurization by-products. An example of application of the Codes at the 'zero discharge' Shand Power Station is discussed. 6 refs., 1 fig., 1 tab.

  1. Compact solar autoclave based on steam generation using broadband light-harvesting nanoparticles.

    Science.gov (United States)

    Neumann, Oara; Feronti, Curtis; Neumann, Albert D; Dong, Anjie; Schell, Kevin; Lu, Benjamin; Kim, Eric; Quinn, Mary; Thompson, Shea; Grady, Nathaniel; Nordlander, Peter; Oden, Maria; Halas, Naomi J

    2013-07-16

    The lack of readily available sterilization processes for medicine and dentistry practices in the developing world is a major risk factor for the propagation of disease. Modern medical facilities in the developed world often use autoclave systems to sterilize medical instruments and equipment and process waste that could contain harmful contagions. Here, we show the use of broadband light-absorbing nanoparticles as solar photothermal heaters, which generate high-temperature steam for a standalone, efficient solar autoclave useful for sanitation of instruments or materials in resource-limited, remote locations. Sterilization was verified using a standard Geobacillus stearothermophilus-based biological indicator.

  2. Active acoustic leak detection for LMFBR steam generators. Pt. 7. Potential for small leak detection

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Hiromichi; Yoshida, Kazuo [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

    1998-05-01

    In order to prevent the expansion of tube damage and to maintain structural integrity in the steam generators (SGs) of fast breeder reactors (FBR), it is necessary to detect precisely and immediately the leakage of water from heat transfer tubes. Therefore, an active acoustic method, which detects the sound attenuation due to bubbles generated in the sodium-water reactions, is being developed. Previous studies have revealed that the active acoustic method can detect bubbles of 10 l/s (equivalence water leak rate about 10 g/s) within 10 seconds in practical steam generators. In order to prevent the expansion of damage to neighboring tubes, however, it is necessary to detect smaller leakage of water from heat transfer tubes. In this study, in order to evaluate the detection sensitivity of the active method, the signal processing methods for emitter and receiver sound and the detection method for leakage within 1 g/s are investigated experimentally, using an SG full-sector model that simulates the actual SGs. A typical result shows that detection of 0.4 l/s air bubbles (equivalent water leak rate about 0.4 g/s) takes about 80 seconds, which is shorter than the propagation time of damage to neighboring tubes. (author)

  3. Numerical simulation of a parabolic trough solar collector for hot water and steam generation

    Science.gov (United States)

    Hachicha, Ahmed Amine

    2016-05-01

    Parabolic trough solar collectors (PTCs) are currently one of the most mature and prominent solar technology for the production of electricity. In order to reduce the electricity cost and improve the overall efficiency, Direct Steam generation (DSG) technology can be used for industrial heat process as well as in the solar fields for electricity production. In the last decades, this technology is experiencing an important development last decades and it is considered as one of the most feasible process for the next generation of power plants using PTCs. A numerical model based on Finite Volume Method (FVM) balance is presented to predict the thermal behavior of a parabolic trough solar collector used for hot water and steam generation. The realistic non-uniform solar flux is calculated in a pre-processing task and inserted to the general model. A numerical-geometrical method based on ray trace and FVM techniques is used to determine the solar flux distribution around the absorber tube with high accuracy.

  4. Detailed partial load investigation of a thermal energy storage concept for solar thermal power plants with direct steam generation

    Science.gov (United States)

    Seitz, M.; Hübner, S.; Johnson, M.

    2016-05-01

    Direct steam generation enables the implementation of a higher steam temperature for parabolic trough concentrated solar power plants. This leads to much better cycle efficiencies and lower electricity generating costs. For a flexible and more economic operation of such a power plant, it is necessary to develop thermal energy storage systems for the extension of the production time of the power plant. In the case of steam as the heat transfer fluid, it is important to use a storage material that uses latent heat for the storage process. This leads to a minimum of exergy losses during the storage process. In the case of a concentrating solar power plant, superheated steam is needed during the discharging process. This steam cannot be superheated by the latent heat storage system. Therefore, a sensible molten salt storage system is used for this task. In contrast to the state-of-the-art thermal energy storages within the concentrating solar power area of application, a storage system for a direct steam generation plant consists of a latent and a sensible storage part. Thus far, no partial load behaviors of sensible and latent heat storage systems have been analyzed in detail. In this work, an optimized fin structure was developed in order to minimize the costs of the latent heat storage. A complete system simulation of the power plant process, including the solar field, power block and sensible and latent heat energy storage calculates the interaction between the solar field, the power block and the thermal energy storage system.

  5. Demand management in steam generation?; Administracion de la demanda en la generacion de vapor?

    Energy Technology Data Exchange (ETDEWEB)

    Plauchu L., A.; Plauchu A., J. A. [Ingenieros consultores (Mexico)

    1997-12-31

    Energy management has acquired capital importance in all the branches of the industrial activity and services either private o public and it has been, from years behind, more familiar the technology for the demand control in the electric power distribution systems, with excellent results in the energy conservation and economy. The rate of consumption for the different forms of energy shows the prevalence of the thermal energy in a variety of end uses. Not much or almost nothing has been said of the term Demand Management or Control in Steam Generation although 12% of the total energy supply is used for this purpose. Steam demand management and control can avoid unnecessary investments that bring along the increment of operational and maintenance problems and it is for sure that it will always give raise to a more rational and economical utilization of the real capacity of steam generation. A real case is commented. [Espanol] La administracion de la energia ha cobrado importancia capital en todos los giros de actividad industrial y de servicios, privados y publicos y es de varios anos atras cada vez mas familiar la tecnologia en el control de demanda en los sistemas de distribucion de energia electrica, con magnificos resultados en el ahorro, energetico y economico. La relacion de consumos para las diferentes formas de energia muestra el predominio de la energia termica en una variedad de usos finales. Poco o nada oimos del termino Administracion o Control de Demanda en la Generacion de Vapor aun cuando el 12% de la oferta total de energia tiene como destino esta aplicacion. La administracion y control de la demanda de vapor puede evitar inversiones innecesarias e incremento de problemas de operacion y mantenimiento y con seguridad reportaran siempre una utilizacion mas racional y economica de la capacidad real de generacion de vapor, se comenta un caso real.

  6. Development of modern CANDU PHWR cross-section libraries for SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Shoman, Nathan T., E-mail: nshoman@vols.utk.edu; Skutnik, Steven E., E-mail: sskutnik@utk.edu

    2016-06-15

    Highlights: • New ORIGEN libraries for CANDU 28 and 37-element fuel assemblies have been created. • These new reactor data libraries are based on modern ENDF/B-VII.0 cross-section data. • The updated CANDU data libraries show good agreement with radiochemical assay data. • Eu-154 overestimated when using ENDF-VII.0 due to a lower thermal capture cross-section. - Abstract: A new set of SCALE fuel lattice models have been developed for the 28-element and 37-element CANDU fuel assembly designs using modern cross-section data from ENDF-B/VII.0 in order to produce new reactor data libraries for SCALE/ORIGEN depletion analyses. These new libraries are intended to provide users with a convenient means of evaluating depletion of CANDU fuel assemblies using ORIGEN through pre-generated cross sections based on SCALE lattice physics calculations. The performance of the new CANDU ORIGEN libraries in depletion analysis benchmarks to radiochemical assay data were compared to the previous version of the CANDU libraries provided with SCALE (based on WIMS-AECL models). Benchmark comparisons with available radiochemical assay data indicate that the new cross-section libraries perform well at matching major actinide species (U/Pu), which are generally within 1–4% of experimental values. The library also showed similar or better results over the WIMS-AECL library regarding fission product species and minor actinoids (Np, Am, and Cm). However, a notable exception was in calculated inventories of {sup 154}Eu and {sup 155}Eu, where the new library employing modern nuclear data (ENDF/B-VII.0) performed substantially poorer than the previous WIMS-AECL library (which used ENDF-B/VI.8 cross-sections for these species). The cause for this discrepancy appears to be due to differences in the {sup 154}Eu thermal capture cross-section between ENDF/B-VI.8 and ENDF/B-VII.0, an effect which is exacerbated by the highly thermalized flux of a CANDU heavy water reactor compared to that of a

  7. Physics and Control Assessment of AN 850 Mw(e) - Leu-Candu Reactor.

    Science.gov (United States)

    Barbone, Michelangelo

    The physics and control assessment of an 850 MW(e) Low Enriched Uranium CANDU Pressurized Heavy Water (LEU -CANDU-PHW) reactor constitute the major objective of this thesis. The use of Low Enriched Uranium fuel in the present CANDU nuclear power generating stations is recognized as economically beneficial due to reduced fuelling costs. The LEU fuel cycle is also recognized as a stepping stone to transit from the present CANDU-PHW once-through natural Uranium cycle to advanced cycles such as those based on Plutonium recycle, once-through Th + U-235 cycle, Thorium with Uranium recycle and net U-235 feed, Thorium with Uranium recycle and Plutonium feed. However, although the use of Low Enriched Uranium in the present CANDU-PHW reactor has economic advantages, and it would act as a technical bridge between the present cycle and advanced cycles, technical problems in different areas of reactor physics and fuel management were anticipated. The present thesis research work adresses the areas of reactor physics, fuel management, and control (in particular, the spatial control of large CANDU-PHW reactors). The main conclusions that have been drawn following these studies are as follows: (1) The Low Enriched Uranium Cycle is feasible in a CANDU-PHW reactor of present design and provided that: (a) The enrichment is kept relatively low (that is, about 1% instead of 0.711%); (b) the number of bundles to be replaced at every refuelling operation is about one-half that of the natural Uranium fuel case; (c) The channels are refuelled in the same direction as the coolant. (2) The response of an LEU-CANDU-PHW reactor to reactivity perturbation such as single- and two-channel refuelling operation, shim transient, shutdown-start-up transient with enrichment levels of 0.9% and 1.2% is essentially very similar {provided that certain conditions in (1) are respected} to that of the natural uranium reactor core case without any reactor reoptimization. The general behaviour of the reactor

  8. Maintenance and repair of LMFBR steam generators: specialists` meeting, O-Arai Engineering Center, Japan, 4-8 June 1984. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1984-07-01

    The Specialists` Meeting on "Maintenance and Repair of LMFBR Steam Generators" was held in Oarai, Japan, from 4-8 June 1984. The meeting was sponsored by the International Atomic Energy Agency on the recommendation of the IAEA International Working Group on Fast Reactors and was hosted by the Power Reactor and Nuclear Fuel Development Corporation of Japan. The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topical areas were discussed by participants: national review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; research and development work on maintenance and repair; and experience on steam generator maintenance and repair.

  9. 饱和蒸汽发电技术应用现状分析%Saturated Steam Generation Technology Overview

    Institute of Scientific and Technical Information of China (English)

    王政伟; 张争光; 韩三飞

    2013-01-01

    Saturated steam power generation technology is an important trend of low temperature waste heat power generation.The situation of saturated steam power generation of modem enterprise is induced and summarized,the key technology and the optimum parameters of the saturated steam power generation are put forward combined with the characteristics of saturated steam and power generation efficiency.The results have an important reference value for each enterprise on adopting the the low temperature of saturated steam.%饱和蒸汽发电技术是低温余热发电的重要趋势.本文通过对现代各企业饱和蒸汽发电的情况进行总结和归纳,提出饱和蒸汽发电的关键技术和趋势,并且结合饱和水蒸气的特点和发电效率提出最佳饱和蒸汽发电参数,对各企业开展低温饱和蒸汽发电具有重要的参考价值.

  10. Study on the regulatory approach of KNGR multiple steam generator tube rupture events

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Sun; Kweon, Y. C.; Lee, S. J.; Lee, Y. S.; Cheong, D. Y.; Park, T. J.; Lee, M. G.; Cheon, Y. H. [Sunmoon Univ., Asan (Korea, Republic of); Cheong, J. H. [Baekseok College of Cultural Studies, Cheonan (Korea, Republic of)

    2001-10-15

    The scope and contents performed in this project are as follows : firstly, reviews of the structure and contents of local and foreign regulatory requirements as well as analysis of design features related to safety improvement and containment bypass during multiple steam generator tube failure of advanced reactors of domestic and foreign countries. Secondly, analyses of the state-of-the-art of the development of local and foreign regulatory requirements, research trends, design features and safety goals of advanced reactors, especially for technical issues related to the containment bypass during MSGTR event. Thirdly, analyses of the event of MSGTR for the KNGR using MAS 1.4 which is the best-estimate system code developed by Korea Atomic Energy Research Institute. Errors in input-decks established last year have been corrected during this analysis. Fourthly, assessment of the effects of several parameters on the consequences following a MSGTR event. Tube rupture location, selection of affected steam generator, tube modeling method, discharge coefficient (C{sub D}) are examined. Fifthly, establishment of regulatory direction of technical issues related to the containment bypass during MSGTR event.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  12. Development of a Robust Model-Based Water Level Controller for U-Tube Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Basher, A.M.H.

    2001-09-04

    Poor control of steam generator water level of a nuclear power plant may lead to frequent nuclear reactor shutdowns. These shutdowns are more common at low power where the plant exhibits strong non-minimum phase characteristics and flow measurements at low power are unreliable in many instances. There is need to investigate this problem and systematically design a controller for water level regulation. This work is concerned with the study and the design of a suitable controller for a U-Tube Steam Generator (UTSG) of a Pressurized Water Reactor (PWR) which has time varying dynamics. The controller should be suitable for the water level control of UTSG without manual operation from start-up to full load transient condition. Some preliminary simulation results are presented that demonstrate the effectiveness of the proposed controller. The development of the complete control algorithm includes components such as robust output tracking, and adaptively estimating both the system parameters and state variables simultaneously. At the present time all these components are not completed due to time constraints. A robust tracking component of the controller for water level control is developed and its effectiveness on the parameter variations is demonstrated in this study. The results appear encouraging and they are only preliminary. Additional work is warranted to resolve other issues such as robust adaptive estimation.

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

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S. [Argonne National Lab., IL (United States)

    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.

  14. Lab assessment of Bruce Unit 4 steam generator top-of-tubesheet cracking

    Energy Technology Data Exchange (ETDEWEB)

    Jevec, J.; Sarver, J. [Babcock and Wilcox Research Center (United States); King, P.; Yu, J. [Babcock and Wilcox Canada Ltd., Cambridge, Ontario (Canada); Sedman, K.; Durance, D. [Bruce Power, Tiverton, Ontario (Canada)

    2009-07-01

    An increasing number of significant circumferential indications were detected at the roll transition zone (RTZ) of Bruce Power Unit 4 steam generator (SG) tubing (sensitized Alloy 600) during the 2006 and 2007 Spring outages. Metallurgical examination of removed tubes found significant IGA/SCC associated with these indications. However, no circumferential indications were detected on Unit 4 SG tubing during the subsequent Fall 2007 and Spring 2008 outages. Based on a review of outage layup conditions it was theorized that the observed degradation occurs during an outage when the steam generator is drained for maintenance in combination with the presence of detrimental contaminants such as sulfur and copper. This theory was tested in the laboratory using a series of electrochemical and simulated crevice exposure tests. The oxygen/hydrazine reaction at room temperature and the resultant effect on the electrochemical potential of the sensitized Alloy 600 tubing were also studied in this program. Results from this test program are presented in this paper. The results indicate that exposure of the solutions to air tends to keep the sample in the sludge at a more reducing condition as compared to the free span tubing above the sludge resulting in a larger driving force for corrosion of the sample in the sludge. The theory that the defects in the RTZ were caused during drain-down outage conditions was shown to be plausible. (author)

  15. Leak rate and burst test data for McGuire Unit 1 steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Sherburne, P.A. [B& W Nuclear Service Co., Lynchburg, VA (United States); Frye, C.R. [Babcock & Wilcox Co., Lynchburg, VA (United States); Mayes, D.B. [Duke Power Co., Charlotte, NC (United States)

    1992-12-31

    To support the development of tube plugging criteria that would allow tubes with through-wall cracks to remain in service, sections of 12 tubes were removed from the McGuire Unit-1 steam generators. These tubes were sent to B&W Nuclear Service Company for metallographic examination and for determination of burst pressure and leak rate at both operating and faulted conditions. Primary water stress corrosion cracking (PWSCC) had degraded these tubes in the tube-to-tubesheet roll transitions. To measure primary-to-secondary leakage at pressures and temperatures equivalent to those in the McGuire Unit-1 steam generators, an autoclave-based test loop was designed and installed at the Babcock & Wilcox Lynchburg Research Center. Sections of the tube containing the roll transitions were then installed in the autoclave and actual primary- to-secondary leakage was measured at 288{degrees}C (550{degrees}F) and at 9 and 18.3 MPa (1300 and 2650 psi) pressure differentials. Following the leak test, the tubes were pressurized internally until the tube wall ruptured. Leak rate, burst pressure, and eddy-current information were then correlated with the through-wall crack lengths as determined by metallographic examination. Results confirm the ability to measure the crack length with eddy-current techniques. Results also support analytical and empirical models developed by the nuclear industry in calculating critical crack lengths in roll transitions.

  16. Experimental verification of the horizontal steam generator boil-off transfer degradation at natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Hyvaerinen, J. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1997-12-31

    The presentation summarises the highlights of experimental results obtained for VVER type horizontal steam generator heat transfer, primary side flow pattern, and mixing in the hot collector during secondary side boil-off with primary at single-phase natural circulation. The experiments were performed using the PACTEL facility with Large Diameter (LD) steam generator models, with collector instrumentation designed specifically for these tests. The key findings are as follows: (1) the primary to secondary heat transfer degrades as the secondary water inventory is depleted, following closely the wetted tube area; (2) a circulatory flow pattern exists in the tube bundle, resulting in reversed flow (from cold to the hot collector) in the lower part of the tube bundle, and continuous flow through the upper part, including the tubes that have already dried out; and (3) mixing of the hot leg flow entering the hot collector and reversed, cold, tube flow remains confined within the collector itself, extending only a row or two above the elevation at which tube flow reversal has taken place. 6 refs.

  17. Wood-Graphene Oxide Composite for Highly Efficient Solar Steam Generation and Desalination.

    Science.gov (United States)

    Liu, Keng-Ku; Jiang, Qisheng; Tadepalli, Sirimuvva; Raliya, Ramesh; Biswas, Pratim; Naik, Rajesh R; Singamaneni, Srikanth

    2017-03-01

    Solar steam generation is a highly promising technology for harvesting solar energy, desalination and water purification. We introduce a novel bilayered structure composed of wood and graphene oxide (GO) for highly efficient solar steam generation. The GO layer deposited on the microporous wood provides broad optical absorption and high photothermal conversion resulting in rapid increase in the temperature at the liquid surface. On the other hand, wood serves as a thermal insulator to confine the photothermal heat to the evaporative surface and to facilitate the efficient transport of water from the bulk to the photothermally active space. Owing to the tailored bilayer structure and the optimal thermo-optical properties of the individual components, the wood-GO composite structure exhibited a solar thermal efficiency of ∼83% under simulated solar excitation at a power density of 12 kW/m(2). The novel composite structure demonstrated here is highly scalable and cost-efficient, making it an attractive material for various applications involving large light absorption, photothermal conversion and heat localization.

  18. Self-assembly of highly efficient, broadband plasmonic absorbers for solar steam generation.

    Science.gov (United States)

    Zhou, Lin; Tan, Yingling; Ji, Dengxin; Zhu, Bin; Zhang, Pei; Xu, Jun; Gan, Qiaoqiang; Yu, Zongfu; Zhu, Jia

    2016-04-01

    The study of ideal absorbers, which can efficiently absorb light over a broad range of wavelengths, is of fundamental importance, as well as critical for many applications from solar steam generation and thermophotovoltaics to light/thermal detectors. As a result of recent advances in plasmonics, plasmonic absorbers have attracted a lot of attention. However, the performance and scalability of these absorbers, predominantly fabricated by the top-down approach, need to be further improved to enable widespread applications. We report a plasmonic absorber which can enable an average measured absorbance of ~99% across the wavelengths from 400 nm to 10 μm, the most efficient and broadband plasmonic absorber reported to date. The absorber is fabricated through self-assembly of metallic nanoparticles onto a nanoporous template by a one-step deposition process. Because of its efficient light absorption, strong field enhancement, and porous structures, which together enable not only efficient solar absorption but also significant local heating and continuous stream flow, plasmonic absorber-based solar steam generation has over 90% efficiency under solar irradiation of only 4-sun intensity (4 kW m(-2)). The pronounced light absorption effect coupled with the high-throughput self-assembly process could lead toward large-scale manufacturing of other nanophotonic structures and devices.

  19. Dryout occurrence in a helically coiled steam generator for nuclear power application

    Directory of Open Access Journals (Sweden)

    Santini L.

    2014-03-01

    Full Text Available Dryout phenomena have been experimentally investigated in a helically coiled steam generator tube. The experiences carried out in the present work are part of a wide experimental program devoted to the study of a GEN III+ innovative nuclear power plant [1].The experimental facility consists in an electrically heated AISI 316L stainless steel coiled tube. The tube is 32 meters long, 12.53 mm of inner diameter, with a coil diameter of 1m and a pitch of 0.79 m, resulting in a total height of the steam generator of 8 meters. The thermo-hydraulics conditions for dryout investigations covered a spectrum of mass fluxes between 199 and 810 kg/m2s, the pressures ranges from 10.7 to 60.7 bar, heat fluxes between 43.6 to 209.3 kW/m2.Very high first qualities dryout, between 0.72 and 0.92, were found in the range of explored conditions, comparison of our results with literature available correlations shows the difficulty in predicting high qualities dryout in helical coils., immediately following the heading. The text should be set to 1.15 line spacing. The abstract should be centred across the page, indented 15 mm from the left and right page margins and justified. It should not normally exceed 200 words.

  20. Fluid-Structure Interaction Effects Modeling for the Modal Analysis of a Steam Generator Tube Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Sigrist, J.F. [DCNS Prop, Serv Tech et Sci, F-44620 La Montagne, (France); Broc, D. [CEA Saclay, Serv Etud Mecan et Sism, F-91191 Gif Sur Yvette, (France)

    2009-07-01

    Seismic analysis of steam generator is of paramount importance in the safety assessment of nuclear installations. These analyses require, in particular, the calculation of frequency, mode shape, and effective modal mass of the system Eigenmodes. As fluid-structure interaction effects can significantly affect the dynamic behavior of immersed structures, the numerical modeling of the steam generator has to take into account FSI. A complete modeling of heat exchangers (including pressure vessel, tubes, and fluid) is not accessible to the engineer for industrial design studies. In the past decades, homogenization methods have been studied and developed in order to model tubes and fluid through an equivalent continuous media, thus avoiding the tedious task to mesh all structure and fluid sub-domains within the tube bundle. Few of these methods have nonetheless been implemented in industrial finite element codes. In a previous paper (Sigrist, 2007, 'Fluid-Structure Interaction Effects Modeling for the Modal Analysis of a Nuclear Pressure Vessel', J. Pressure Vessel Technol., 123, p. 1-6), a homogenization method has been applied to an industrial case for the modal analysis of a nuclear rector with internal structures and coupling effects modeling. The present paper aims at investigating the extension of the proposed method for the dynamic analysis of tube bundles with fluid-structure interaction modeling. The homogenization method is compared with the classical coupled method in terms of eigenfrequencies, Eigenmodes, and effective modal masses. (authors)

  1. Technical and economic assessment of power generation from municipal solid waste incineration on steam cycle

    Energy Technology Data Exchange (ETDEWEB)

    Romero Luna, Carlos Manuel; Carrocci, Luiz Roberto; Ferrufino, Gretta Larisa Aurora Arce; Balestieri, Jose Antonio Perrella [Dept. of Energy. UNESP, Sao Paulo State University, Guaratingueta, SP (Brazil)], e-mails: carrocci@feg.unesp.br, perrella@feg.unesp.br

    2010-07-01

    Nowadays, there is a concern in development of environmentally friendly methods for a municipal solid waste (MSW) management and demand for renewable energy sources. The source of waste is increasing, and the capacity and availability Landfill treatment and disposal are coming to be insufficient. In Sao Paulo City, the 10 million inhabitants produce 10,000 t of residential solid waste daily, being that 76% this quantity goes to landfill sites. In order to adopt a new treatment technology for MSW that will promote a solution minimizing this problem, within the order of priorities regarding waste management, the MSW incineration with energy recovery shown as the leading choice on the point of view of efficiency in converting energy. MSW incineration with energy recovery received wide acceptance from various countries including European Union members and the rest of the world in the past 15 years. Incineration has the ability decrease 90 % the volume of waste to be used in landfills, increasing the useful life of existing as well as a reduction in the emission of greenhouse gases. MSW incineration systems have a low global warming potential (GWP). now has become a less important source of dioxins and furans due to the current available technology. MSW incineration with energy recovery could contribute considerably in the energy matrix, thus promote the conservation of non-renewable resources. This paper proposes the assessment the technical and economic feasibility of a steam cycle with conventional steam generator for MSW incineration with energy recovery for power generation in Sao Paulo City. Will be developed a thermoeconomic analysis aiming at the total power generation product of MSW incineration, and the assessment investment cost regarding the total sale of power generated. The study shows that Sao Paulo City has potential for power generation from the MSW incineration, although it has a high cost investment this technology shown as a suitable alternative for

  2. Dimensional Measurements of Fresh CANDU Fuel Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Jo, Chang Keun; Jung, Jong Yeob; Koo, Dae Seo; Cho, Moon Sung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    This paper intends to provide the dimensional measurements of fresh CANDU fuel (37-element) bundle for the estimation of deformation of post-irradiated (PI) bundle. It is expensive and difficult to measure the fretting wear of bearing pad, the element bowing and the waviness of endplate at the two-phase high flow condition (above 24 kg/s) of out-of-reactor test. So, it is recommended to compare the geometry of fresh bundle with that of PI bundle to estimate the integrity of fuel bundle in the CANDU-6 fuel channel with two-phase flow condition. The measurement system has been developed to provide the visual inspection and the dimensional measurements within the accuracy of 10 {mu}m. It is applicable in-air and underwater to the CANDU bundle as well as the CANFLEX bundle. The in-air measurements of the 36 fresh CANDU bundles (S/N: B400892 {approx} B400927) are done by this system from February 2004 to March 2004 in the PHWR fresh fuel storage building of KNFC. These bundles are produced by KNFC manufacturing procedure and are waiting for the delivery to the Wolsong-3 plant, and are planned to load into the proposed test channels. The detail measurements contain the outer rod profile (including the bearing pad), the diameter of bundle, the bowing of bundle, the rod length and the surface profile of end plate (waviness)

  3. Determination of Fuel Consumption Indexes of Co-generation Combined Cycle Steam and Gas Units with unfired waste heat boilers

    Directory of Open Access Journals (Sweden)

    S. A. Kachan

    2010-01-01

    Full Text Available The paper presents the developed methodology and the results of determination of fuel consumption indexes of co-generation combined cycle steam and gas units (PGU with unfired waste heat boilers apply to PGU-230 of 3-d co-generation power plant ofMinsk. 

  4. Optimisation of a Kalina cycle for a central receiver solar thermal power plant with direct steam generation

    DEFF Research Database (Denmark)

    Modi, Anish; Haglind, Fredrik

    2014-01-01

    Central receiver solar thermal power plants are regarded as one of the promising ways to generate electricity in near future. They offer the possibility of using high temperatures and pressures to achieve high efficiencies with standard power cycles. A direct steam generation approach can be used...... for a central receiver solar thermal power plant with direct steam generation. The variation in the cycle performance with respect to the turbine inlet ammonia mass fraction and pressure and a comparison of the initial investment with that of the basic Rankine cycle are also presented. Only high live steam...... for such plants for improved performance. This approach can also be combined with using advanced power cycles like the Kalina cycle, which uses a zeotropic mixture of ammonia and water instead of pure water as the working fluid. This paper presents the optimisation of a particular Kalina cycle layout...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  6. Evaluation of Hybrid Power Plants using Biomass, Photovoltaics and Steam Electrolysis for Hydrogen and Power Generation

    Science.gov (United States)

    Petrakopoulou, F.; Sanz, J.

    2014-12-01

    Steam electrolysis is a promising process of large-scale centralized hydrogen production, while it is also considered an excellent option for the efficient use of renewable solar and geothermal energy resources. This work studies the operation of an intermediate temperature steam electrolyzer (ITSE) and its incorporation into hybrid power plants that include biomass combustion and photovoltaic panels (PV). The plants generate both electricity and hydrogen. The reference -biomass- power plant and four variations of a hybrid biomass-PV incorporating the reference biomass plant and the ITSE are simulated and evaluated using exergetic analysis. The variations of the hybrid power plants are associated with (1) the air recirculation from the electrolyzer to the biomass power plant, (2) the elimination of the sweep gas of the electrolyzer, (3) the replacement of two electric heaters with gas/gas heat exchangers, and (4) the replacement two heat exchangers of the reference electrolyzer unit with one heat exchanger that uses steam from the biomass power plant. In all cases, 60% of the electricity required in the electrolyzer is covered by the biomass plant and 40% by the photovoltaic panels. When comparing the hybrid plants with the reference biomass power plant that has identical operation and structure as that incorporated in the hybrid plants, we observe an efficiency decrease that varies depending on the scenario. The efficiency decrease stems mainly from the low effectiveness of the photovoltaic panels (14.4%). When comparing the hybrid scenarios, we see that the elimination of the sweep gas decreases the power consumption due to the elimination of the compressor used to cover the pressure losses of the filter, the heat exchangers and the electrolyzer. Nevertheless, if the sweep gas is used to preheat the air entering the boiler of the biomass power plant, the efficiency of the plant increases. When replacing the electric heaters with gas-gas heat exchangers, the

  7. Instantaneous determination of heat transfer coefficients in a steam generator for an alternative energy upgrade system

    Energy Technology Data Exchange (ETDEWEB)

    Sotelo, S.S.; Romero, R.J. [Univ. Autonoma del Estado de Morelos, Cuernavaca Morelos (Mexico). Centro di Investigacion en Ingeneria y Ciencias Aplicadas; Best, R. [Univ. Autonoma de Mexico, Temixco, Morelos (Mexico). Centro de Investigacion en Energie

    2009-07-01

    A mathematical model was used to characterize the thermal behaviour of a steam generator in an alternative energy upgrade system. A thermodynamic cycle was used to increase the temperatures produced by solar, geothermal, and waste heat from industrial processes. The absorption heat transformer (AHT) process can be used in industrial processes where low temperature heat flows occur. Alternative energy was supplied to the generator where the working fluid was condensed and then transported to the evaporator through an expansion valve. Vapor was then transported to the absorber in order to deliver heat at a higher temperature. The solution was then returned to the generator in order to start the cycle again. A heat exchanger was placed between the absorber and the generator in order to preheat incoming solutions from the generator. The mathematical model was used to simulate heat transfer in the generator in order to determine optimal operating conditions. Heat transfer coefficients were calculated using equations reported for single phase flow. It was concluded that the highest heat transfer coefficients were obtained for a Reynolds number of 2300 with an alternative energy source of 90 degrees C at mass flows of 4 L/m. 33 refs., 14 figs.

  8. Effect of Ovality on Maximum External Pressure of Helically Coiled Steam Generator Tubes with a Rectangular Wear

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Dong In; Lim, Eun Mo; Huh, Nam Su [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of); Choi, Shin Beom; Yu, Je Yong; Kim, Ji Ho; Choi, Suhn [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    A structural integrity of steam generator tubes of nuclear power plants is one of crucial parameters for safe operation of nuclear power plants. Thus, many studies have been made to provide engineering methods to assess integrity of defective tubes of commercial nuclear power plants considering its operating environments and defect characteristics. As described above, the geometric and operating conditions of steam generator tubes in integral reactor are significantly different from those of commercial reactor. Therefore, the structural integrity assessment of defective tubes of integral reactor taking into account its own operating conditions and geometric characteristics, i. e., external pressure and helically coiled shape, should be made to demonstrate compliance with the current design criteria. Also, ovality is very specific characteristics of the helically coiled tube because it is occurred during the coiling processes. The wear, occurring from FIV (Flow Induced Vibration) and so on, is main degradation of steam generator tube. In the present study, maximum external pressure of helically coiled steam generator tube with wear is predicted based on the detailed 3-dimensional finite element analysis. As for shape of wear defect, the rectangular shape is considered. In particular, the effect of ovality on the maximum external pressure of helically coiled tubes with rectangular shaped wear is investigated. In the present work, the maximum external pressure of helically coiled steam generator tube with rectangular shaped wear is investigated via detailed 3-D FE analyses. In order to cover a practical range of geometries for defective tube, the variables affecting the maximum external pressure were systematically varied. In particular, the effect of tube ovality on the maximum external pressure is evaluated. It is expected that the present results can be used as a technical backgrounds for establishing a practical structural integrity assessment guideline of

  9. Evaluation of CFD Methods for Simulation of Two-Phase Boiling Flow Phenomena in a Helical Coil Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Shaver, Dillon [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, Yang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Vegendla, Prasad [Argonne National Lab. (ANL), Argonne, IL (United States); Tentner, Adrian [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-30

    The U.S. Department of Energy, Office of Nuclear Energy charges participants in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program with the development of advanced modeling and simulation capabilities that can be used to address design, performance and safety challenges in the development and deployment of advanced reactor technology. The NEAMS has established a high impact problem (HIP) team to demonstrate the applicability of these tools to identification and mitigation of sources of steam generator flow induced vibration (SGFIV). The SGFIV HIP team is working to evaluate vibration sources in an advanced helical coil steam generator using computational fluid dynamics (CFD) simulations of the turbulent primary coolant flow over the outside of the tubes and CFD simulations of the turbulent multiphase boiling secondary coolant flow inside the tubes integrated with high resolution finite element method assessments of the tubes and their associated structural supports. This report summarizes the demonstration of a methodology for the multiphase boiling flow analysis inside the helical coil steam generator tube. A helical coil steam generator configuration has been defined based on the experiments completed by Polytecnico di Milano in the SIET helical coil steam generator tube facility. Simulations of the defined problem have been completed using the Eulerian-Eulerian multi-fluid modeling capabilities of the commercial CFD code STAR-CCM+. Simulations suggest that the two phases will quickly stratify in the slightly inclined pipe of the helical coil steam generator. These results have been successfully benchmarked against both empirical correlations for pressure drop and simulations using an alternate CFD methodology, the dispersed phase mixture modeling capabilities of the open source CFD code Nek5000.

  10. Ultrasonic inspection of steam-generator tube axial cracking using Lamb wave

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Seok

    2007-02-15

    In this study, the interaction of Lamb wave propagating thin tube structure with finite vertical discontinuity was studied using both modal decomposition method (MDM) and experimental method. For MDM, a global matrix formulation and orthogonality of Lamb mode was employed to describe the boundary condition of finite vertical discontinuity of the tube and the mode conversion phenomenon respectively. The final form of governing equation by MDM was a linear matrix equation which could be solved using a simple matrix identity. The calculation result showed that, below the cut-off frequency, reflection amplitudes of both A0 and S0 Lamb mode increase as the depth of discontinuity increased beyond the threshold value. An experimental investigation was performed using a Hertzian-contact transducer and steam-generator tubes to verify the calculation results by MDM. A0 Lamb mode was selected as a test signal considering the characteristics of the transducer and previous studies. The experiment for mode identification using half-sectioned tube verified that the Hertzian-contact transducer effectively generated A0 Lamb mode. Tests performed using steam-generator tubes with EDM (electric discharge machined) axial notches showed that the deeper notches produced the higher reflection echo. A0 Lamb mode interacted with the notch having a depth larger than 1/40 of wave length, or corresponding to 30% of the wall thickness. This finding was in good agreement with previous studies and the prediction by MDM. The experiment using real crack specimens to estimate the deviation of reflection amplitude showed that the reflection cross-section of real crack was very similar with that of EDM notch. Therefore, specimens with EDM notches can be used as reference blocks for Lamb wave UT calibration.

  11. An innovative approach for Steam Generator Pressure Control of a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Gaikwad, Avinash J., E-mail: avinashg@barc.gov.in [Reactor Safety Division, BARC, Trombay, Mumbai 400094 (India); Vijayan, P.K. [Reactor Engineering Divisions, BARC, Trombay, Mumbai 400094 (India); Bhartiya, Sharad [Chemical Engineering Departments, IIT, Powai, Mumbai (India); Kumar, Rajesh; Lele, H.G.; Vaze, K.K. [Reactor Safety Division, BARC, Trombay, Mumbai 400094 (India)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer Most of the transients/accidents have their origin in the mismatch among the heat generated in the reactor core and the heat removal in the SGs. Black-Right-Pointing-Pointer The main objective of balancing the heat generation, transfer and removal gets lost due to simplification of SGPC leading to reduced availability. Black-Right-Pointing-Pointer A new Advanced Process Control (APC) is proposed to ride over the existing SGPC to achieve the goal of prompt removal of the heat transfer mismatch. Black-Right-Pointing-Pointer The APC logic will lead to overall performance improvements and plant availability for all other transients also. - Abstract: The main function of the Steam Generator Pressure Control (SGPC) Program is to match the power (heat) generation in the reactor core with the heat removal in the steam generators (SGs). For most of the designs these programs have been over simplified to cater to the limitation of the instrumentation and control, hardware and software. The main objective of balancing the heat generation, transfer and removal gets lost in the process, which leads to reduction in the availability of the nuclear power plant. This is reflected in under utilization of the process and control system provisions to avoid reactor trips on low/high pressure. Most of the transients/accidents have their origin in the mismatch among the heat generated in the reactor core and the heat removal in the SGs. A new Advanced Process Control (APC) based supervisory controller is proposed to ride over the existing SGPC to achieve the goal. This APC makes use of the estimated/measured heat generation-removal error to alter the SGPC set point to tide over the transients after detection. The transients are detected based on the magnitude of this error to activate the APC. After tiding over the transient successfully the control switches back to the existing SGPC. For evaluation of this error additional instrumentation is

  12. EVALUATION OF THE APPLICABLE REACTIVITY RANGE OF A REACTIVITY COMPUTER FOR A CANDU-6 REACTOR

    Directory of Open Access Journals (Sweden)

    EUN KI LEE

    2014-04-01

    Full Text Available Recently, a CANDU digital reactivity computer system (CDRCS to measure the worth of the liquid zone controller in a CANDU-6 was developed and successfully applied to a physics test of refurbished Wolsong Unit 1. In advance of using the CDRCS, its measureable reactivity range should be investigated and confirmed. There are two reasons for this investigation. First, the CANDU-6 has a larger reactor and smaller excore detectors than a general PWR and consequently the measured reactivity is likely to reflect the peripheral power variation only, not the whole core. The second reason is photo neutrons generated from the interaction of the moderator and gamma-rays, which are never considered in a PWR. To evaluate the limitations of the CDRCS, several tens of three-dimensional steady and transient simulations were performed. The simulated detector signals were used to obtain the dynamic reactivity. The difference between the dynamic reactivity and the static worth increases in line with the water level changes. The maximum allowable reactivity was determined to be 1.4 mk in the case of CANDU-6 by confining the difference to less than 1%.

  13. Research on steam-supply performance of ship micro-superheated steam generating system%船舶微过热蒸汽发生系统供汽性能研究

    Institute of Scientific and Technical Information of China (English)

    杨元龙

    2016-01-01

    In order to improve ship micro-superheated steam generating system stability and optimize its performance parameters, the micro-superheated steam generating system steam-supply response characteristic under the steady and transient state will be cleared. The ship micro-superheated steam generating system was taken as the mechanism model herein. The steady characteristics of velocity pressure, and temperature field for micro-superheated steam generating system were calculated by method of CFD simulation. The boundary conditions were introduced to treat as actual operating parameters of this system. The dynamic simulation study on micro-superheated steam generating system steam-supply response characteristic was carried out. The key parameters distributions of saturated and superheated steam mixing massflow, micro-superheated steam pressure and temperature were obtained. Meanwhile, the micro-superheated steam mixing factor was proposed, which could express quantificationally micro-superheated steam generating system mixing characteristics and effects of micro-superheated steam temperature. The calculated results showed that the pressure drop of saturated steam was higher than one of superheated steam. The micro-superheated steam pressure reduced gradually, leading to larger saturated and superheated steam massflow. These caused micro-superheated steam mixing factor to reduce, which resulted in that the micro-superheated steam temperature reduced slightly. Based on the analysis of the micro-superheated steam generating system steam-supply performance parameters, it could satisfy demand for equipment performance. These could be used to design ship steam power system.%为提高船舶微过热蒸汽发生系统的稳定性和优化微过热系统性能参数,探析稳态、动态工况下微过热蒸汽发生系统供汽响应特性。本文以船舶微过热蒸汽发生系统为机理模型,采用 CFD模拟方法计算了微过热蒸汽发生系统速度场、压

  14. Development of a Program for Predicting Flow Instability in a Once-through Sodium- Heated Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Kim, Dehee; Kim, Jong Bum; Lee, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A SG selected for PGSFR is of a once-through integrated type. It is a vertical counter flow shell and tube heat exchanger with sodium on the shell side and water-steam in the tubes. The phenomenon of two-phase flow instability has been observed in many industrial domains such as boiling systems and steam generators. In this paper, a computer program developed for predicting two-phase flow instability in a steam generator under axial non-uniform heat flux is presented, and analysis results for verification are presented. A computer code was developed for investigating the two-phase flow stability under sodium-heated conditions in the shell-side of a SG. A solution algorithm for the sodium flow field and tube conduction has been developed for application to sodium-heated SG.

  15. CCFL in hot legs and steam generators and its prediction with the CATHARE code

    Energy Technology Data Exchange (ETDEWEB)

    Geffraye, G.; Bazin, P.; Pichon, P. [CEA/DRN/STR, Grenoble (France)

    1995-09-01

    This paper presents a study about the Counter-Current Flow Limitation (CCFL) prediction in hot legs and steam generators (SG) in both system test facilities and pressurized water reactors. Experimental data are analyzed, particularly the recent MHYRESA test data. Geometrical and scale effects on the flooding behavior are shown. The CATHARE code modelling problems concerning the CCFL prediction are discussed. A method which gives the user the possibility of controlling the flooding limit at a given location is developed. In order to minimize the user effect, a methodology is proposed to the user in case of a calculation with a counter-current flow between the upper plenum and the SF U-tubes. The following questions have to be made clear for the user: when to use the CATHARE CCFL option, which correlation to use, and where to locate the flooding limit.

  16. Laser cleaning of steam generator tubing based on acoustic emission technology

    Energy Technology Data Exchange (ETDEWEB)

    Hou, Su-xia; Luo, Ji-jun; Shen, Tao; Li, Ru-song [Xi' an Hi-Tech Institute, Xi' an (China)

    2015-12-15

    As a physical method, laser cleaning technology in equipment maintenance will be a good prospect. The experimental apparatus for laser cleaning of heat tubes in the steam generator was designed according to the results of theoretical analysis. There are two conclusions; one is that laser cleaning technology is attached importance to traditional methods. Which has advantages in saving on much manpower and material resource and it is a good cleaning method for heat tubes. The other is that the acoustic emission signal includes lots of information on the laser cleaning process, which can be used as real-time monitoring in laser cleaning processes. When the laser acts for 350 s, 100 % contaminants of heat tubes is cleaned off, and the sensor only receives weak AE signal at that time.

  17. Simulation modeling of nuclear steam generator water level process--a case study

    Science.gov (United States)

    Zhao; Ou; Du

    2000-01-01

    Simulation modeling of the nuclear steam generator (SG) water level process in Qinshan Nuclear Power Plant (QNPP) is described in this paper. A practical methodology was adopted so that the model is both simple and accurate for control engineering implementation. The structure of the model is in the form of a transfer function, which was determined based on first-principles analysis and expert experience. The parameters of the model were obtained by taking advantage of the recorded historical response curves under the existing closed-loop control system. The results of process dimensional data verification and experimental tests demonstrate that the simulation model depicts the main dynamic characteristics of the SG water level process and is in accordance with the field recorded response curves. The model has been successfully applied to the design and test of an advanced digital feedwater control system in QNPP.

  18. Evaluation of the Dynamic Velocity Effect for Steam Generator Wide Range Water Level

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, In Soo; Nam, Ki Haeng; Kim, Jeong Hoon; Yun, Jae Hee [Korea Power Engineering Company, Daejeon (Korea, Republic of)

    2010-05-15

    The measurement of Steam Generator (SG) water level is based upon pressure differential of the level transmitter. As shown in Fig. 1, if the location of a lower tap is in the downcomer region, a deviation between the indicated level and the actual level occurs. This phenomenon is called 'velocity effect' or 'dynamic effect.' This effect needs to be addressed to obtain a more accurate SG water level. Korean Utility Requirements Document (KURD) requires Downcomer Velocity Effect (DVE) to be quantified and to be considered in the instrument requirements. In this paper, DVE occurred through downcomer will be evaluated for SG wide range (WR) level for OPR1000

  19. Generation and characterization of OH and O radicals by atmospheric pressure steam/oxygen plasma

    CERN Document Server

    Roy, N C; Alam, M K; Talukder, M R

    2016-01-01

    Atmospheric pressure steam/oxygen plasma is generated by a 88 Hz, 6kV AC power supply. The properties of the produced plasma are investigated by optical emission spectroscopy (OES). The relative intensity, rotational, vibrational, excitation temperatures and electron density are studied as function of applied voltage, electrode spacing and oxygen flow rate. The rotational and vibrational temperatures are determined simulating the bands with the aid of LIFBASE simulation software. The excitation temperature is obtained from the CuI transition taking non-thermal equilibrium condition into account employing intensity ratio method. The electron density is approximated from the H_{\\alpha} Stark broadening using the Voigt profile fitting method. It is observed that the rotational and vibrational temperatures are decreased with increasing electrode spacing and O2 flow rate, but increased with the applied voltage. The excitation temperature is found to increase with increasing applied voltage and O2 flow rate, but de...

  20. Tire gasification, fuels produced and their use to generate steam and/or electricity

    Energy Technology Data Exchange (ETDEWEB)

    Sefchovich, E.; Goodman, J.; Miliaras, E.S.

    1995-12-31

    This paper focuses on gasification, a technology which addresses the environmental and health problems which the storage and disposal of spent tires represents. Gasification can help bring under control the growing concern over the 1/4 billion tires this country alone discards each year, and the 3 billion tires already contained, in existing piles in an environmentally benign and economically desirable manner. Gasification reverses, in essence, the process by which rubber is obtained from petroleum. Gasification converts the rubber into gaseous and liquid fuels, leaving behind a relatively small amount of solid residue. Such fuels can be used to generate steam and/or electricity, or used as a raw material to manufacture other products.

  1. Computational simulation of cold work effect on PWSCC growth in Alloy 600TT steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Jun-Young; Kim, Yun-Jae [Korea University, Seoul (Korea, Republic of); Kim, Jong-Sung [Sunchon National University, Suncheon (Korea, Republic of)

    2016-02-15

    The paper presents verification results for the validity of a numerical method considering the effect of cold work on Primary water stress corrosion cracking (PWSCC) growth rate in the Alloy 600TT steam generator tubes with a part-through single axial PWSCC. PWSCC growth simulations using Finite element (FE) analysis were performed with considering various cold work levels of the material. From the FE analysis results, the cold work effect was investigated from the variations of the PWSCC growth rate vs. Stress intensity factor (SIF) for the various cold work degrees and initial SIF values. Investigated results were compared with experimental test data available. It was identified that the numerical method could adequately assess the cold work effect on PWSCC growth in the Alloy 600TT tubes. In the simulation, it was found that the cold work could strongly influence the PWSCC growth rate even in a low degree of cold work, less than 2%.

  2. Description of a program for steam generators; Descripcion de un programa de generadores de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Campana, F. J.

    2014-10-01

    Steam Generators (SGs) are a key component of PWR nuclear power plants, maintaining their structural integrity throughout their life time is necessary to allow for long term operation (LTD) of PWR plants. NEI 97-06 provides the fundamental elements to be included in a SG Program. In addiction it describes performance criteria that SG tubes have to meet in order to provide reasonable assurance that the tubes are still able to maintain specific safety function. Hence, it is mandatory for plants with SGs to have defined a SG program consistent with NEI 97-06 and contains the elements which are described by it. This Program must contain some elements such as, Degradation Assessment, inspection and Integrity Assessment, among other. (Author)

  3. Active acoustic leak detection for LMFBR steam generator. Sound attenuation due to bubbles

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Hiromichi; Sakuma, Toshio [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

    1995-06-01

    In the steam generators (SG) of LMFBR, it is necessary to detect the leakage of water from tubes of heat exchangers as soon as it occurs. The active acoustic detection method has drawn general interest owing to its short response time and reduction of the influence of background noise. In this paper, the application of the active acoustic detection method for SG is proposed, and sound attenuation by bubbles is investigated experimentally. Furthermore, using the SG sector model, sound field characteristics and sound attenuation characteristics due to injection of bubbles are studied. It is clarified that the sound attenuation depends upon bubble size as well as void fraction, that the distance attenuation of sound in the SG model containing heat transfer tubes is 6dB for each two-fold increase of distance, and that emitted sound attenuates immediately upon injection of bubbles. (author).

  4. Development of the experimental evaluation method for crevice chemistry in steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, In Hyoung [Soonchunghyang Univ., Cheonan (Korea); Hwang, Il Soon; Lee, Na Young; Kim, Ji Hyun; Lim, Jung Yeon; Bahn, Chi Bum; Oh, Young Jin; Han, Byung Chan; Oh, Si Hyoung [Seoul National Univ., Seoul (Korea)

    2001-04-01

    Steam generator tube degradation problems is very sensitive to water chemistry. But even if the secondary water chemistry is well controlled, it is needed. Tubesheet crevice has three boiling regimes with depth: liquid penetration and discharge(or wet) region, liquid drop scattering(or dry and wet) region, and dryout region. This results showed a good agreement with earlier works. High temperature, high pressure tubesheet crevice simulation system was constructed. As {delta}T increased, the temperature gradient in crevice and time constant for concentration increased. When the experimental results were compared with MULTEQ calculation results, a similar behavior was shown, packed crevice have longer time constant for Na concentration and showed heavier concentration that open crevice. The verification experiment for Molar Ratio Control and advanced Molar Ration Control test were conducted. To check the applicability of boric acid as pH neutralizer another experiment was conducted. 40 refs., 102 figs., 3 tabs. (Author)

  5. Simulation of the fluid-structure-interaction of steam generator tubes and bluff bodies

    Energy Technology Data Exchange (ETDEWEB)

    Kuehlert, Karl [ANSYS, Inc. (United States)], E-mail: kue@fluent.com; Webb, Stephen [Sandia National Laboratories (United States); Schowalter, David; Holmes, William; Chilka, Amarvir; Reuss, Steve [ANSYS, Inc. (United States)

    2008-08-15

    The accuracy of computational fluid dynamics in simulating the cross-flow around a steam generator and the feasibility of a full scale coupled CFD/FEA fluid-structure-interaction (FSI) analysis is examined through successive validations. The study begins with a comparison between experiment and computation of flow within a stationary tube bank. Results from the simulation of an individual tube experiencing two-degree-of-freedom flow-induced vibration (at a Reynolds number of 3800) are then shown to compare favorably to experimental results. Finally, free vibration of a single cantilevered hydrofoil is simulated with comparison of mean square acceleration at resonant and non-resonant velocities, respectively. The magnitudes and frequencies of vibration are shown to be accurately captured.

  6. Formation and nonvolatile memory characteristics of W nanocrystals by in-situ steam generation oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Shih-Cheng [Department of Electrical Engineering and Institute of Electronic Engineering, National Tsing Hua University, Taiwan (China); Chang, Ting-Chang, E-mail: tcchang@mail.phys.nsysu.edu.t [Department of Physics and Center for Nanoscience and Nanotechnology, National Sun Yat-Sen University Taiwan (China); Hsieh, Chieh-Ming [Institute of Electronics, National Chiao Tung University, Taiwan, HsinChu, 300 Taiwan (China); Li, Hung-Wei [Department of Photonics and Institute of Electro-Optical Engineering, National Chiao Tung University, Hsinchu, Taiwan (China); Sze, S.M. [Institute of Electronics, National Chiao Tung University, Taiwan, HsinChu, 300 Taiwan (China); Nien, Wen-Ping; Chan, Chia-Wei [ProMOS Technologies, No. 19 Li Hsin Rd., Science-Based Industrial Park, Hsinchu, 300 Taiwan (China); Yeh, Fon-Shan [Department of Electrical Engineering and Institute of Electronic Engineering, National Tsing Hua University, Taiwan (China); Tai, Ya-Hsiang [Department of Photonics and Display Institute, National Chiao Tung University, Hsinchu, Taiwan (China)

    2010-12-30

    The authors provide the formation and memory effects of W nanocrystals nonvolatile memory in this study. The charge trapping layer of stacked a-Si and WSi{sub 2} was deposited by low pressure chemical vapor deposition (LPCVD) and was oxidized by in-situ steam generation system to form uniform W nanocrystals embedded in SiO{sub 2}. Transmission electron microscopy analyses revealed the microstructure in the thin film and X-ray photon-emission spectra indicated the variation of chemical composition under different oxidizing conditions. Electrical measurement analyses showed the different charge storage effects because the different oxidizing conditions influence composition of trapping layer and surrounding oxide quality. Moreover, the data retention and endurance characteristics of the formed W nanocrystal memory devices were compared and studied. The results show that the reliability of the structure with 2% hydrogen and 98% oxygen at 950 {sup o}C oxidizing condition has the best performance among the samples.

  7. Selection of statistical distributions for prediction of steam generator tube degradation

    Energy Technology Data Exchange (ETDEWEB)

    Stavropoulos, K.D.; Gorman, J.A. [Dominion Engr., Inc., McLean, VA (United States); Staehle, R.W. [Univ. of Minnesota, Minneapolis, MN (United States); Welty, C.S. Jr. [Electric Power Research Institute, Palo Alto, CA (United States)

    1992-12-31

    This paper presents the first part of a project directed at developing methods for characterizing and predicting the progression of degradation of PWR steam generator tubes. This first part covers the evaluation of statistical distributions for use in such analyses. The data used in the evaluation of statistical distributions included data for primary water stress corrosion cracking (PWSCC) at roll transitions and U-bends, and intergranular attack/stress corrosion cracking (IGA/SCC) at tube sheet and tube support plate crevices. Laboratory data for PWSCC of reverse U-bends were also used. The review of statistical distributions indicated that the Weibull distribution provides an easy to use and effective method. Another statistical function, the log-normal, was found to provide essentially equivalent results. Two parameter fits, without an initiation time, were found to provide the most reliable predictions.

  8. Steam generator tube support plate degradation in French plants: maintenance strategy

    Energy Technology Data Exchange (ETDEWEB)

    Gauchet, J.-P. [EDF, NPP Operations/Maintenance Dept. (France); Gillet, N. [FRAMATOME, Steam Generator Dept. (France); Stindel, M. [EDF, Central Labs. (France)

    1998-07-01

    This paper reports on the degradations of Steam Generator (SG) Tube Support Plates (TSPs) observed in French plants and the maintenance strategy adopted to continue operating the plant without any decrease of the required safety level. Only drilled carbon steel TSPs of early SGs are affected. Except the particular damage of the TSP8 of FESSENHEIM 2 caused by chemical cleaning procedures implemented in 1992, two main problems were observed almost exclusively on the upper TSP: Ligaments ruptured near the aseismic block located at 215 degrees. This degradation is perfectly detectable by bobbin coil inspection. It occurs very early in the life of the SG as can be seen from the records of previous inspections and no evolution of the signals was observed. This damage can be detected for 51M model SGs on several sites; Wastage of the ligaments resulting in enlargement of flow holes with in some cases complete consumption of a ligament. This damage was only observed for SGs of at GRAVELINES. This damage evolved cycle after cycle. Detailed studies were performed to analyze tubing behavior when a tube is not supported by the upper TSP because of missing ligaments. These studies evaluated the risk of vibratory instability, the behavior of both the TSP and the tubing in case of a seismic event or a LOCA and finally the behavior of the TSP in case of a Steam Line Break. Concerning vibratory instability it was possible to define zones where stability could not be demonstrated. Dampine, cables and sentinel plugs were then used when necessary to eliminate the risk of Steam Generator Tube Rupture (SGTR). For accidental conditions, it could be shown that no unacceptable damage occurs and that the core cooling function of the SG is always maintained if some tubes are plugged. From this analysis, It was possible to define the inspection programs for the different plants taking into account the specific situation of each plant regarding the damages detected. These programs include

  9. Fuel condition in Canadian CANDU 6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, R.H.; Macici, N [Hydro-Quebec, Montreal, Quebec (Canada); Gibb, R. [New Brunswick Power, Lepreau, NB (Canada); Purdy, P.L.; Manzer, A.M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Hydro, Toronto, Ontario (Canada)

    1997-07-01

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO{sub 2} fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  11. Performance tests and efficiency analysis of Solar Invictus 53S - A parabolic dish solar collector for direct steam generation

    Science.gov (United States)

    Jamil, Umer; Ali, Wajahat

    2016-05-01

    This paper presents the results of performance tests conducted on Solar Invictus 53S `system'; an economically effective solar steam generation solution designed and developed by ZED Solar Ltd. The system consists of a dual axis tracking parabolic solar dish and bespoke cavity type receiver, which works as a Once Through Solar Steam Generator `OTSSG' mounted at the focal point of the dish. The overall performance and efficiency of the system depends primarily on the optical efficiency of the solar dish and thermal efficiency of the OTSSG. Optical testing performed include `on sun' tests using CCD camera images and `burn plate' testing to evaluate the sunspot for size and quality. The intercept factor was calculated using a colour look-back method to determine the percentage of solar rays focused into the receiver. Solar dish tracking stability tests were carried out at different times of day to account for varying dish elevation angles and positions, movement of the sunspot centroid was recorded and logged using a CCD camera. Finally the overall performance and net solar to steam efficiency of the system was calculated by experimentally measuring the output steam parameters at varying Direct Normal Insolation (DNI) levels at ZED Solar's test facility in Lahore, Pakistan. Thermal losses from OTSSG were calculated using the known optical efficiency and measured changes in output steam enthalpy.

  12. Validation of WIMS-CANDU using Pin-Cell Lattices

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Young; Min, Byung Joo; Park, Joo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The WIMS-CANDU is a lattice code which has a depletion capability for the analysis of reactor physics problems related to a design and safety. The WIMS-CANDU code has been developed from the WIMSD5B, a version of the WIMS code released from the OECD/NEA data bank in 1998. The lattice code POWDERPUFS-V (PPV) has been used for the physics design and analysis of a natural uranium fuel for the CANDU reactor. However since the application of PPV is limited to a fresh fuel due to its empirical correlations, the WIMS-AECL code has been developed by AECL to substitute the PPV. Also, the WIMS-CANDU code is being developed to perform the physics analysis of the present operating CANDU reactors as a replacement of PPV. As one of the developing work of WIMS-CANDU, the U{sup 238} absorption cross-section in the nuclear data library of WIMS-CANDU was updated and WIMS-CANDU was validated using the benchmark problems for pin-cell lattices such as TRX-1, TRX-2, Bapl-1, Bapl-2 and Bapl-3. The results by the WIMS-CANDU and the WIMS-AECL were compared with the experimental data.

  13. Steam Turbines

    Science.gov (United States)

    1981-01-01

    Turbonetics Energy, Inc.'s steam turbines are used as power generating systems in the oil and gas, chemical, pharmaceuticals, metals and mining, and pulp and paper industries. The Turbonetics line benefited from use of NASA research data on radial inflow steam turbines and from company contact with personnel of Lewis Research Center, also use of Lewis-developed computer programs to determine performance characteristics of turbines.

  14. Automated refueling simulations of a CANDU for the exploitation of thorium fuels

    Science.gov (United States)

    Holmes, Bradford

    CANDU nuclear reactors are in a unique circumstance where they are able to utilize and exploit a number of different fuel options to provide power as a utility. Thorium, a fertile isotope found naturally, is one option that should be explored. Thorium is more abundant than uranium, which is the typical fuel in the reactor and the availability of thorium makes nuclear energy desirable to more countries. This document contains the culmination of a project that explores, tests, and analyzes the feasibility of using thorium in a CANDU reactor. The project first develops a set of twodimensional lattice and three dimensional control rod simulations using the DRAGON Version 4 nuclear physics codes. This step is repeated for many concentrations of thorium. The data generated in these steps is then used to determine a functional enrichment of thorium. This is done via a procedural elimination and optimization of certain key parameters including but not limited to average exit burnup and reactivity evolution. For the purposes of this project, an enrichment of 1 % thorium was found viable. Full core calculations were done using the DONJON 4 code. CANFUEL, a program which simulates the refueling operations of a CANDU reactor for this fuel type was developed and ran for a simulation period of one hundred days. The program and the fuel selection met all selected requirements for the entirety of the simulation period. CANFUEL requires optimization for fuel selection before it can be used extensively. The fuel selection was further scrutinized when a reactivity insertion event was simulated. The adjuster rod 11 withdrawal from the core was analyzed and compared to classical CANDU results in order to ensure no significant deviations or unwanted evolutions were encountered. For this case, the simulation results were deemed acceptable with no significant deviations from the classical CANDU case.

  15. A comprehensive flow-induced vibration model to predict crack growth and leakage potential in steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    El Bouzidi, Salim [School of Engineering, University of Guelph, Guelph, Ontario N1G 2W1 (Canada); Hassan, Marwan, E-mail: mahassan@uoguelph.ca [School of Engineering, University of Guelph, Guelph, Ontario N1G 2W1 (Canada); Riznic, Jovica [Operational Engineering Assessment Division, Canadian Nuclear Safety Commission, Ottawa, Ontario K1P 5S9 (Canada)

    2015-10-15

    Highlights: • Comprehensive flow induced vibrations time domain model was developed. • Simulations of fluidelastic instability and turbulence were conducted. • Nonlinear effect due to the clearances at the supports was studied. • Prediction of stresses due to fluid excitation was obtained. • Deterministic and stochastic analyses for crack and leakage rate were conducted. - Abstract: Flow-induced vibrations (FIVs) are a major threat to the operation of nuclear steam generators. Turbulence and fluidelastic instability are the two main excitation mechanisms leading to tube vibrations. The consequences to the operation of steam generators are premature wear of the tubes, as well as development of cracks that may leak hazardous fluids. This paper investigates the effect of tube support clearance on the integrity of tube bundles within steam generators. Special emphasis will be placed on crack propagation and leakage rates. A crack growth model is used to simulate the growth of surface flaws and through-wall cracks of various initial sizes due to a wide range of support clearances. Leakage rates are predicted using a two-phase flow leakage model. Nonlinear finite element analysis is used to simulate a full U-bend subjected to fluidelastic and turbulence forces. Monte Carlo simulations are then used to conduct a probabilistic assessment of steam generator life due to crack development.

  16. Data bank on hydrodynamics, thermal tests and tube temperature regimes of PGV-4 and PVG-1000 natural steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Ageev, A.G.; Vasileva, R.V.; Nigmatulin, B.I.; Titov, V.F.; Tarankov, G.N. [EREC Electrogorsk Research and Engineering Centre of Nuclear Plants Safety, Moscow (Russian Federation)

    1995-12-31

    The data bank was prepared by EREC, OKB `Hydropress` using results of static and dynamic tests of PGV-4 and PGV- 1000 natural steam generators cared out at Kolskaya, Novo-Voronezhskaya, Ugno-Ukrainskaya, Balakov-skaya and Hmelnitskaya NPP within period of 1974-1993. It is destined for making calculation codes verification. (authors).

  17. Cleaning of OPR1000 Steam Generator by Ultrasonic Cavitation in Water

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Wootae [Korea Hydro and Nuclear Power Co., Ltd, Daejeon (Korea, Republic of); Kim, Sangtae; Yoon, Sangjung; Choi, Yongseok [Saean Engineering Corporation, Seoul (Korea, Republic of)

    2013-05-15

    Magnetic wheels are attached to the transducers to prevent tube damage which may be caused by wear between the transducers and SG tubes. To remove heat generated by transducers, we used water to water heat exchanger. Sludge removed from tube sheet area of the steam generator was pumped to filtering station for removing impurities in it. We designed an ultrasonic cleaning system for application to OPR1000 S/G. The technology was developed for removing sludge in OPR1000 S/G. However, the technology could easily be applied to other types of S/Gs. For cleaning OPR1000 SG, we designed an ultrasonic cleaning system with 12 transducers, 15 generators, a WRS, and a water treatment system. An experiment with a single transducer and the full scale OPR1000 S/G mock-up did not show very satisfactory result in ultrasound energy level. However, we expect sufficient effects if we apply 12 or more transducers in this case considering our previous experimental results as shown in the references. The ultrasonic cleaning system will be ready in August this year for performance test. After several experiments and the experiments followed, we are planning to apply this cleaning system for removing sludge in Korean OPR1000 S/Gs.

  18. Design and modelling of an innovative three-stage thermal storage system for direct steam generation CSP plants

    Science.gov (United States)

    Garcia, Pierre; Vuillerme, Valéry; Olcese, Marco; El Mourchid, Nadim

    2016-05-01

    Thermal Energy Storage systems (TES) for a Direct Steam Generation (DSG) solar plant feature preferably three stages in series including a latent heat storage module so that steam can be recovered with a limited temperature loss. The storage system designed within the Alsolen Sup project is characterized by an innovative combination of sensible and latent modules. A dynamic model of this three-stage storage has been developed and applied to size the storage system of the Alsolen Sup® plant demonstrator at CEA Cadarache. Results of this simulation show that this promising concept is an efficient way to store heat in DSG solar plants.

  19. Computer code analysis of steam generator in thermal-hydraulic test facility simulating nuclear power plant; Ydinvoimalaitosta kuvaavan koelaitteiston hoeyrystimien analysointi tietokoneohjelmilla

    Energy Technology Data Exchange (ETDEWEB)

    Virtanen, E.

    1995-12-31

    In the study three loss-of-feedwater type experiments which were preformed with the PACTEL facility has been calculated with two computer codes. The purpose of the experiments was to gain information about the behaviour of horizontal steam generator in a situation where the water level on the secondary side of the steam generator is decreasing. At the same time data that can be used in the assessment of thermal-hydraulic computer codes was assembled. The purpose of the work was to study the capabilities of two computer codes, APROS version 2.11 and RELAP5/MOD3.1, to calculate the phenomena in horizontal steam generator. In order to make the comparison of the calculation results easier the same kind of model of the steam generator was made for both codes. Only the steam generator was modelled, the rest of the facility was given for the codes as a boundary condition. (23 refs.).

  20. Characteristics Evaluation of a CO2-Caputuring Power Generation System with Reheat Cycle Utilizing Regenerative Oxygen-Combustion Steam-Superheater

    Science.gov (United States)

    Pak, Pyong Sik

    A new CO2-capturing power generation system is proposed that can be easily realized by applying conventional technologies. In the proposed system, the temperature of middle-pressure steam in a thermal power plant is raised by utilizing oxygen-combusting regenerative steam-superheater. The generated CO2 by combusting fuel in the superheater can be easily separated and captured from the exhaust gas at condenser outlet, and is liquefied. The superheated steam is used to drive a steam turbine power generation system. By adopting a high efficient combined cycle power generation system as an example, it has been shown that the proposed system can increase power output by 10.8%, decrease the CO2 emission amount of the total integrated system by 18.6% with power generation efficiency drop of 2.36% compared with the original power plant without CO2-capture, when superheated steam temperature is 750°C

  1. Field test of two high-pressure direct-contact downhole steam generators. Volume II. Oxygen/diesel system

    Energy Technology Data Exchange (ETDEWEB)

    Moreno, J.B.

    1983-07-01

    A field test of an oxygen/diesel fuel, direct contact steam generator has been completed. The field test, which was a part of Project DEEP STEAM and was sponsored by the US Department of Energy, involved the thermal stimulation of a well pattern in the Tar Zone of the Wilmington Oil Field. The activity was carried out in cooperation with the City of Long Beach and the Long Beach Oil Development Company. The steam generator was operated at ground level, with the steam and combustion products delivered to the reservoir through 2022 feet of calcium-silicate insulated tubing. The objectives of the test included demonstrations of safety, operational ease, reliability and lifetime; investigations of reservoir response, environmental impact, and economics; and comparison of those points with a second generator that used air rather than oxygen. The test was extensively instrumented to provide the required data. Excluding interruptions not attributable to the oxygen/diesel system, steam was injected 78% of the time. System lifetime was limited by the combustor, which required some parts replacement every 2 to 3 weeks. For the conditions of this particular test, the use of trucked-in LOX resulted in liess expense than did the production of the equivalent amount of high pressure air using on site compressors. No statistically significant production change in the eight-acre oxygen system well pattern occurred during the test, nor were any adverse effects on the reservoir character detected. Gas analyses during the field test showed very low levels of SOX (less than or equal to 1 ppM) in the generator gaseous effluent. The SOX and NOX data did not permit any conclusion to be drawn regarding reservoir scrubbing. Appreciable levels of CO (less than or equal to 5%) were measured at the generator, and in this case produced-gas analyses showed evidence of significant gas scrubbing. 64 figures, 10 tables.

  2. Characterization of steam generated anti-corrosive oxide films on Aluminium alloys

    DEFF Research Database (Denmark)

    Din, Rameez Ud; Jellesen, Morten Stendahl; Ambat, Rajan

    2014-01-01

    alloy surfaces were exposed to high pressure steam produced by an autoclave at a temperature of 107 – 121 °C and pressure of 15 -17 psi for 10 minutes to produce a thin coating of aluminium oxide. The aim of this study is to understand the effect of high pressure steam with and without different...

  3. On the influence of manufacturing practices on the SCC behavior of Alloy 690 steam generator tubing

    Energy Technology Data Exchange (ETDEWEB)

    Doherty, P.E.; Doyle, D.M. [Babcock and Wilcox International Div., Cambridge, Ontario (Canada); Sarver, J.M.; Miglin, B.P. [Babcock and Wilcox Research Div., Alliance, OH (United States)

    1996-12-31

    Thermally treated (TT) Alloy 690 is the tubing materials of choice for replacement steam generators (RSGs) throughout the world. It is manufactured using a variety of processing methods with regards to melt practice and thermomechanical forming. Studies assessing the IGSCC resistance of Alloy 690 TT SG tubing have identified a variability in the corrosion performance of nominally identical alloys. While tubing of comparable bulk chemistry may exhibit variations in microchemistry as a result of different melt practice, the correlation between melt practice and SCC resistance is difficult to assess due to other contributing factors. The other contributing factors are identified in this investigation as microstructural features whose generation is dependent on features of particular strain-anneal forming methods by which SG tubes are fabricated. In this study the microstructural characteristics which appear to affect inservice corrosion performance of Alloy 690 TT SG tubes were evaluated. The studies included extensive microstructural examinations in addition to CERT tests performed on actual Alloy 690 TT nuclear SG tubing. The CERT test results indicate that Alloy 690 TT tubing processed at higher mill anneal temperatures display the highest degree of stress corrosion cracking (SCC) resistance. This observation is discussed with reference to carbide distributions, textural aspects and grain boundary orientation character.

  4. Development of active acoustic method for water leak detection of LMFBR steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Hiromichi; Yoshida, Kazuo; Kinoshita, Izumi [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    2001-06-01

    In order to prevent the expansion of tube damage and to maintain structural integrity in the steam generators (SGs) of fast breeder reactors (FBRs), it is necessary to detect precisely and immediately the leakage of water from heat transfer tubes. Therefore, an active acoustic method, which detects the sound attenuation due to bubbles generated in the sodium-water reactions, is being developed. In this study, in order to evaluate the detection sensitivity of the active method, the signal processing methods for emitter and receiver and the detection method for leakage are investigated experimentally. In-water experiments performed by using an SG full-sector model that simulates the actual SGs. As an experimental result, the received sound attenuation for 10s was more than 10dB from air bubble injection when injected bubble of 10 l/s (equivalence water leak rate about 10 g/s.) The attenuation of sound are least affected by bubble injection position of heat transfer tubes bunch department. It is clarified that the background noise hardly influenced water leak detection performance as a result of having examined influence of background noise. (author)

  5. Doppler method leak detection for LMFBR steam generators. Pt. 3. Investigation of detection sensitivity and method

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Hiromichi; Kinoshita, Izumi [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab

    2001-04-01

    To prevent the expansion of tube damage and to maintain structural integrity in the steam generators (SGs) of a fast breeder reactor (FBR), it is necessary to detect precisely and immediately any leakage of water from heat transfer tubes. Therefore, the Doppler method was developed. Previous studies have revealed that, in the SG full-sector model that simulates actual SGs, the Doppler method can detect bubbles of 0.4 l/s within a few seconds. However in consideration of the dissolution rate of hydrogen generated by a sodium-water reaction even from a small water leak, it is necessary to detect smaller leakages of water from the heat transfer tubes. The detection sensitivity of the Doppler method and the influence of background noise were experimentally investigated. In-water experiments were performed using the SG model. The results show that the Doppler method can detect bubbles of 0.01 l/s (equivalent to a water leak rate of about 0.01 g/s) within a few seconds and that the background noise has little effect on water leak detection performance. The Doppler method thus has great potential for the detection of water leakage in SGs. (author)

  6. Post-scram Liquid Metal cooled Fast Breeder Reactor (LMFBR) heat transport system dynamics and steam generator control: Figures

    Science.gov (United States)

    Brukx, J. F. L. M.

    1982-06-01

    Dynamic modeling of LMFBR heat transport system is discussed. Uncontrolled transient behavior of individual components and of the integrated heat transport system are considered. For each component, results showing specific dynamic features of the component and/or model capability were generated. Controlled dynamic behavior for alternative steam generator control systems during forced and natural sodium coolant circulation was analyzed. Combined free and forced convection of laminar and turbulent vertical pipe flow of liquid metals was investigated.

  7. Potential use of wood and agriculture wastes as steam generator fuel for thermal enhanced oil recovery. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Kosstrin, H.M.; McDonald, R.K.

    1979-01-01

    Enhanced oil recovery by steam injection methods produces over 200,000 barrels per day of crude oil in California. A sizeable portion of the produced crude, up to 40% for some projects, may be burned to generate steam for injection into the reservoir. The purpose of this study is to evaluate the potential to use wood and agriculture wastes to replace crude oil as steam generator fuel. The Bakersfield area of California's San Joaquin Valley is the focus for this paper. Production from thermal EOR methods centers around Bakersfield and agriculture and wood wastes are available from the San Joaquin Valley and the nearby Sierra Nevada mountains. This paper documents the production of waste materials by county, estimated energy value of each material, and estimated transportation cost for each material. Both agriculture and wood wastes were found to be available in sizeable quantities and could become attractive steam generation fuels. However, some qualifications need to be made on the use of these materials. Transportation costs will probably limit the range of shipping these materials to perhaps 50 to 100 miles. Availability is subject to competition from existing and developing uses of these materials, such as energy sources in their immediate production area. Existing steam generators probably cannot be retrofitted to burn these materials. Fluidized bed combustion, or low Btu gasification, may be a good technology for utilization. FBC or FBG could accept a variety of waste materials. This will be important because the amount of any single waste may not be large enough to support the energy requirements of a good size thermal f a good size thermal EOR operation.

  8. Analysis on the Current Status of Chemical Decontamination Technology of Steam Generators in the Oversea Nuclear Power Plants (NPPs)

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Taebin; Kim, Sukhoon; Kim, Juyoul; Kim, Juyub; Lee, Seunghee [FNC Technology Co. Ltd., Yongin (Korea, Republic of)

    2015-10-15

    The steam generators in Hanbit Unit 3 and 4 are scheduled to be replaced in 2018 and 2019, respectively. Nevertheless, the wastes from the dismantled steam generators are currently just on-site stored in the NPP because there are no disposal measures for the waste and lack of the decontamination techniques for large-sized metallic equipment. In contrast, in the oversea NPPs, there are many practical cases of chemical decontamination not only for oversized components in the NPPs such as reactor pressure vessel and steam generator, but also for major pipes. Chemical decontamination technique is more effective in decontaminating the components with complicated shape compared with mechanical one. Moreover, a high decontamination factor can be obtained by using strong solvent, and thereby most of radionuclides can be removed. Due to these advantages, the chemical decontamination has been used most frequently for operation of decontaminating the large-sized equipment. In this study, an analysis on the current status of chemical decontamination technique used for the steam generators of the foreign commercial NPPs was performed. In this study, the three major chemical decontamination processes were reviewed, which are applied to the decommissioning process of the steam generators in the commercial NPPs of the United States, Germany, and Belgium. The three processes have the different features in aspect of solvent, while those are based in common on the oxidation and reduction between the target metal surface and solvents. In addition, they have the same goals for improving the decontamination efficiency and decreasing the amount of the secondary waste generation. Based on the analysis results on component sub-processes and major advantages and disadvantages of each process, Table 2 shows the key fundamental technologies for decontamination of the steam generator in Korea and the major considerations in the development process of each technology. It is necessary to prepare

  9. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.K.; Snell, V. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca; West, J. [Candesco, Toronto, Ontario (Canada); Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2006-07-01

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life, and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. AECL initiated pre-licensing reviews of the ACR reactor design in Canada, US and China, with an objective to take into account regulatory feedback early in the design process. The Canadian Nuclear Safety Commission (CNSC) is performing a pre-project pre-licensing assessment of the ACR design. The objective of the assessment is to issue a formal statement as to whether there are any fundamental barriers that would prevent the licensing of the new CANDU reactor design in Canada under the Nuclear Safety and Control Act. The CNSC review is being conducted in four phases. In Phase 1 (September 2003 to September 2004) CNSC performed a pre-licensing review of the ACR-700, and focused on the design process, methodology, design concepts and R and D. CNSC staff reviewed about 100 reports, and submitted to AECL questions and comments. In Phase 2 (September 2004 to August 2005) AECL provided responses and additional information to CNSC on their comments and questions in Phase 1. Phase 3 is the Transition Phase (September 2005 to May 2006), bridging the transition from the ACR-700 to the ACR-1000 design. Phase 3 focused on review of generic aspects of the ACR design, on the Safety

  10. Field test of two high-pressure, direct-contact downhole steam generators. Volume I. Air/diesel system

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, B.W.

    1983-05-01

    As a part of the Project DEEP STEAM to develop technology to more efficiently utilize steam for the recovery of heavy oil from deep reservoirs, a field test of a downhole steam generator (DSG) was performed. The DSG burned No. 2 diesel fuel in air and was a direct-contact, high pressure device which mixed the steam with the combustion products and injected the resulting mixture directly into the oil reservoir. The objectives of the test program included demonstration of long-term operation of a DSG, development of operational methods, assessment of the effects of the steam/combustion gases on the reservoir and comparison of this air/diesel DSG with an adjacent oxygen/diesel direct contact generator. Downhole operation of the air/diesel DSG was started in June 1981 and was terminated in late February 1982. During this period two units were placed downhole with the first operating for about 20 days. It was removed, the support systems were slightly modified, and the second one was operated for 106 days. During this latter interval the generator operated for 70% of the time with surface air compressor problems the primary source of the down time. Thermal contact, as evidenced by a temperature increase in the production well casing gases, and an oil production increase were measured in one of the four wells in the air/diesel pattern. Reservoir scrubbing of carbon monoxide was observed, but no conclusive data on scrubbing of SO/sub x/ and NO/sub x/ were obtained. Corrosion of the DSG combustor walls and some other parts of the downhole package were noted. Metallurgical studies have been completed and recommendations made for other materials that are expected to better withstand the downhole combustion environment. 39 figures, 8 tables.

  11. Energy and exergy analysis of the Kalina cycle for use in concentrated solar power plants with direct steam generation

    DEFF Research Database (Denmark)

    Knudsen, Thomas; Clausen, Lasse Røngaard; Haglind, Fredrik

    2014-01-01

    In concentrated solar power plants using direct steam generation, the usage of a thermal storage unit based only on sensible heat may lead to large exergetic losses during charging and discharging, due to a poor matching of the temperature profiles. By the use of the Kalina cycle, in which...... evaporation and condensation takes place over a temperature range, the efficiency of the heat exchange processes can be improved, possibly resulting also in improved overall performance of the system. This paper is aimed at evaluating the prospect of using the Kalina cycle for concentrated solar power plants...... with direct steam generation. The following two scenarios were addressed using energy and exergy analysis: generating power using heat from only the receiver and using only stored heat. For each of these scenarios comparisons were made for mixture concentrations ranging from 0.1 mole fraction of ammonia to 0...

  12. An Isothermal Steam Expander for an Industrial Steam Supplying System

    OpenAIRE

    Chen-Kuang Lin; Guang-Jer Lai; Yoshiyuki Kobayashi; Masahiro Matsuo; Min-Chie Chiu

    2015-01-01

    Steam is an essential medium used in the industrial process. To ensure steam quality, small and middle scale boilers are often adopted. However, because a higher steam pressure (compared to the necessary steam pressure) is generated, the boiler’s steam pressure will be reduced via a pressure regulator before the steam is directed through the process. Unfortunately, pressure is somewhat wasted during the reducing process. Therefore, in order to promote energy efficiency, a pressure regulator i...

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Roussel, G. [AIB-Vincotte Nuclear, Brussels (Belgium)

    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.

  15. Integrity assurance of the secondary side of steam generator in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joung Soo; Kim, Hong Pyo; Lim, Yun Soo; Hwang, Seong Sik; Yi, Yong Sun; Kim, Dong Jin; Kim, Sung Soo; Jung, Man Gyo

    2005-09-15

    Residual stresses on the expansion transition regions of steam generator tubes expanded by explosive and hydraulic expansion techniques were measured using several different methods such as strain gauge, XRD, electrolytic polishing, and stress corrosion cracking methods. The SCC method was applied by measuring the cracking time using C-ring specimens to which precisely measured stress had been imposed and comparing the cracking time of the expanded tube specimens in order to estimate the magnitude of residual stress developed on the expansion transition regions. Axial residual stress on the outer surface of both, Inconel 600 and 690 tubes was measured to be mainly compressive, which can not induce circumferential ODSCC on the expansion transition regions. According to SCC test results, SCC was not observed to occur on the expansion transition regions of the expanded model specimen tubes, which means that the residual stresses developed on the expansion transition regions by the explosive and the hydraulic expansion methods are not big enough to induce SCC. However, sludge piled up on the top of tubesheet during operation of NPPs might change the stress state on the expansion transition regions, which can result in occurring SCC.

  16. Uncertainty analysis for probabilistic steam generators tube rupture in LBB applications

    Energy Technology Data Exchange (ETDEWEB)

    Durbec, V.; Pitner, P.; Pages, D. [Electricite de France, 78 - Chatou (France). Research and Development Div.; Riffard, T. [Electricite de France, 69 - Villeurbanne (France). Engineering and Construction Div.; Flesch, B. [Electricite de France, 92 - Paris la Defense (France). Generation and Transmission Div.

    1997-10-01

    Steam Generators (SG) of Pressurized Water Reactors have experienced world wide various types of tube degradations, mainly from stress corrosion cracking; because of this damage, primary-secondary leakage or tube rupture can occur. Safety against the risk of tube rupture is achieved through a combination of periodic in-service inspections (eddy current testing), surveillance of leaks during operation (leak before break concept) and tube plugging. In order to optimize the tube bundle SG maintenance, Electricite de France has developed a specific software named COMPROMIS. The model, based on probabilistic fracture mechanics makes it possible to quantify the influence of in service inspections and maintenance work on the risk of a SG Tube Rupture (SGTR), taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive examinations, crack initiation and propagation, critical sizes, leak before risk of break, etc...). This paper focuses on the leak rate calculation module and presents a sensitivity study of the influence of the leak before break on the conditional failure probability. (author) 8 refs.

  17. Ligament rupture and unstable burst behaviors of axial flaws in steam generator U-bends

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, Chi Bum, E-mail: bahn@pusan.ac.kr [Pusan National University, 2 Busandaehak-ro 63 beon-gil, Geumjeong-gu, Busan 609-735 (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering & Construction Co. Inc., Seongnam 463-870 (Korea, Republic of); Majumdar, Saurin [Argonne National Laboratory, Lemont, IL 60439 (United States)

    2015-11-15

    Highlights: • Ligament rupture and unstable burst pressure tests were conducted with U-bends. • In general, U-bends showed higher ligament rupture and burst pressures than straight tubes. • U-bend test data was bounded by 90% lower limit of the probabilistic models for straight tubes. • Prediction models for straight tubes could be conservatively applied to U-bends. - Abstract: Incidents of U-bend cracking in steam generator (SG) tubes have been reported, some of which have led to tube rupture. Experimental and analytical modeling efforts to determine the failure criteria of flawed SG U-bends are limited. To evaluate structural integrity of flawed U-bends, ligament rupture and unstable burst pressure tests were conducted on 57 and 152 mm bend radius U-bends with axial electrical discharge machining notches. In general, the ligament rupture and burst pressures of the U-bends were higher than those of straight tubes with similar notches. To quantitatively address the test data scatter issue, probabilistic models were introduced. All ligament rupture and burst pressures of U-bends were bounded by 90% lower limits of the probabilistic models for straight tubes. It was concluded that the prediction models for straight tubes could be applied to U-bends to conservatively evaluate the ligament rupture and burst pressures of U-bends with axial flaws.

  18. Impurities incorporation into magnetite scale formed on simulated steam generator tubing

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, K.; Yamaguchi, K.; Koike, M. [Kyushu Electric Power Co., Inc. (Japan); Kawamura, H.; Hirano, H. [Central Research Inst. of Electric Power Industry (Japan); Yamada, Y.; Nakamura, T. [The Kansai Electric Power Co., Inc. (Japan)

    2002-07-01

    From a viewpoint of ensuring the integrity of steam generators (SGs) tubing in PWR plants, the research was made into how impurities in the secondary coolant are incorporated into magnetite (Fe{sub 3}O{sub 4}) scale formed on the tube in a laboratory test. We experimented with a method to form Fe{sub 3}O{sub 4} scale on a tube under a boiling heat transfer condition in the laboratory test, simulating the conditions of SG in the actual PWR plants. Based on the scale formation method, we investigated the incorporation of sulfur (S) into the scale. S is known as the most common impurity solved in the secondary coolant and a dominant factor in making heat transfer crevice environment acidic. The effects of sodium (Na) and silicon (Si), solved in test solution with S, on the S incorporation into scale were also investigated. The test resulted in a double-layered scale being formed on the tube surface, with the outer scale being porous and the inner scale dense. It was revealed that the S incorporation into scales was affected by the S concentration in the solution and existence of other impurities, such as Na and Si. (authors)

  19. A method of cleaning the wash waters of steam-generators working on sulfurous fuel oils

    Energy Technology Data Exchange (ETDEWEB)

    Shishckenko, V.V.

    1980-12-30

    The method of cleaning the wash waters of steam generators can be used to treat vanadium-containing wash water of low temperature heating surfaces of the boilers, of electric heat stations, and other boilers. In order to increase the economic efficacy by lowering requirements in one type and by preventing the cementing of the surfaces of the heater, 40-45% of the water is completely heated to 32-35/sup 0/C (after the second reciprocal stage). It is mixed with sodium sulfate, and later with additional water and lime. It is passed through a layer of glauberite, and cooled to 2-10/sup 0/C. Subsequently, it is passed through a sodium sulfate layer and added to the return stage by means of sodium hydroxide. Spent regeneration solutions from the cationic hydrogen filters are used as the additional water. Lowering the concentration of calcium sulfate in the return-water prevents its crystallization in the device and supply-lines--which decreases the use of raw material and increases the reliability of the return system.

  20. Inducement of IGA/SCC in Inconel 600 steam generator tubing during unit outages

    Energy Technology Data Exchange (ETDEWEB)

    Durance, D.; Sedman, K. [Bruce Power, Tiverton, Ontario (Canada); Roberts, J. [CANTECH Associates Ltd., Burlington, Ontario (Canada); King, P. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada); Gorman, J. [Dominion Engineering, Reston, VA (United States); Allen, R. [Kinectrics, Inc., Toronto, Ontario (Canada)

    2008-07-01

    The degradation of Unit 4 SG tubing by IGA/SCC has limited both the operating period and end of life predictions for Unit 4 since restart in late 2003. The circumferential IGA/SCC has been most significant in SG4 with substantial increases in both initiation and growth rates from 2005 through the spring of 2007. A detailed review of the occurrence of circumferential OD IGA/SCC at the RTZ in the HL TTS region of Bruce 4 steam generator tubes has led a conclusion that it is probable that the IGA/SCC has been the result of attack by partially reduced sulfur species such as tetrathionates and thiosulfates during periods of low temperature exposure. It is believed that attack of this type has mostly likely occurred during startup evolutions following outages as the result the development of aggressive reduced sulfur species in the TTS region during periods when the boilers were fully drained for maintenance activities. The modification of outage practices to limit secondary side oxygen ingress in the spring of 2007 has apparently arrested the degradation and has had significant affects on the allowable operating interval and end of life predictions for the entire unit. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Nassersharif, Bahram (New Mexico State University, Las Cruces, NM); Caffey, Thurlow Washburn Howell; Jedlicka, Russell P. (New Mexico State University, Las Cruces, NM); 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.

  2. Risk analysis of heat recovery steam generator with semi quantitative risk based inspection API 581

    Science.gov (United States)

    Prayogo, Galang Sandy; Haryadi, Gunawan Dwi; Ismail, Rifky; Kim, Seon Jin

    2016-04-01

    Corrosion is a major problem that most often occurs in the power plant. Heat recovery steam generator (HRSG) is an equipment that has a high risk to the power plant. The impact of corrosion damage causing HRSG power plant stops operating. Furthermore, it could be threaten the safety of employees. The Risk Based Inspection (RBI) guidelines by the American Petroleum Institute (API) 58 has been used to risk analysis in the HRSG 1. By using this methodology, the risk that caused by unexpected failure as a function of the probability and consequence of failure can be estimated. This paper presented a case study relating to the risk analysis in the HRSG, starting with a summary of the basic principles and procedures of risk assessment and applying corrosion RBI for process industries. The risk level of each HRSG equipment were analyzed: HP superheater has a medium high risk (4C), HP evaporator has a medium-high risk (4C), and the HP economizer has a medium risk (3C). The results of the risk assessment using semi-quantitative method of standard API 581 based on the existing equipment at medium risk. In the fact, there is no critical problem in the equipment components. Damage mechanisms were prominent throughout the equipment is thinning mechanism. The evaluation of the risk approach was done with the aim of reducing risk by optimizing the risk assessment activities.

  3. Structural alterations, pore generation, and deacetylation of α- and β-chitin submitted to steam explosion.

    Science.gov (United States)

    Tan, Too Shen; Chin, Hui Yen; Tsai, Min-Lang; Liu, Chao-Lin

    2015-05-20

    The purpose of this study was to use an environmentally friendly steam explosion method to achieve α- and β-chitin structural alterations, pore generation, and deacetylation, enhancing the degree of deacetylation (DD) in chitin and extending its applications. The samples of α- and β-chitin possessing various moisture contents that were exploded at 9 kg/cm(2) exhibited higher DDs, lower densities, lower crystallinity and more porous structures compared to unexploded chitin. After explosion, β-chitin exhibited a larger expansion ratio, lower crystallinity and contained a larger proportion of small-sized particles compared to α-chitin. The highest DD values of exploded α- and β-chitin with 75% moisture content were 42.9% and 43.7%, respectively. The exploded chitin samples with lower moisture content exhibited lower DDs, densities, crystallinity indices, smaller particle sizes, and higher expansion ratios than the chitin samples with higher moisture content. The chitin samples with lower moisture content also contained larger and more numerous pores.

  4. Development and experimental validation of a computational model for a helically coiled steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Colorado, D.; Hernandez, J.A. [Centro de Investigacion en Ingenieria y Ciencia Aplicadas (CIICAp), Universidad Autonoma del Estado de Morelos (UAEM), Av. Universidad 1001, Col. Chamilpa, C.P. 62209 Cuernavaca, Morelos (Mexico); Papini, D.; Santini, L.; Ricotti, M.E. [Department of Energy, CeSNEF-Nuclear Engineering Division, Politecnico di, Milano, Via La Masa, 34, 20156, Milan (Italy)

    2011-04-15

    A computational model is developed to describe the thermo-fluid-dynamic behaviour of a helically coiled steam generator device working with water and widely adopted in the nuclear industry. The discretized governing equations are coupled using an implicit step by step method. The mathematical model includes: a subcooled liquid region, a two-phase flow region, and a superheated vapour region (according to the once-through nature of the heat exchanger). All the flow variables (enthalpies, temperatures, pressures, vapour qualities, velocities, heat fluxes, etc.), together with the thermo-physical properties, are evaluated at each point of the grid in which the domain is discretized. A full-scale experimental investigation carried out at SIET thermal-hydraulics labs in Piacenza (Italy), and aimed at characterizing the fluid-dynamic behaviour of two-phase flows in helically coiled tubes, is referenced in the present paper. Two-phase pressure drops data reduction allowed optimizing a suitable form of the friction factor multiplier required by momentum balance equation. Comparisons of the numerical simulations with a wide range of two-phase pressure drops measurements (experiments conducted both in diabatic and adiabatic conditions) are shown in order to validate the proposed model. (authors)

  5. ASCO steam generators operating experience. Safety criteria for defect management and effectiveness of preventive measures

    Energy Technology Data Exchange (ETDEWEB)

    Toribio, E.L. [Associacion Nuclear Asco AIE, Barcelona (Spain)

    1997-02-01

    ASCO NPP is a two W-PWR 930 Mwe Units. Each Unit is provided with three Westinghouse Model D3 steam generators which are of preheater type and Inconel 600 MA as tube material. The Secondary side was designed and erected with copper alloys. Unit I: 81.072 EFPH, and Unit II: 69.720 EFPH. The results of the Eddy Currents Inspections performed during the first refueling outage showed Denting at tube support plates and PWSCC at roll transition zone in Unit I and Denting in Unit II. Later inspections showed other types of damages, such as: (1) ODSCC at tube support plates intersections. (2) Circumferential cracks OD and ID at roll transition zone. (3) Wear at antivibration bars and preheater baffles level. Consequently, in order to limit the plugging rate, A.N. ASCO decided to license new plugging criteria in addition to the 40% depth criterion included in Technical Specification. The new licensing criteria and surveillance requirements, varying with tube zone, are explained in the paper.

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

    Energy Technology Data Exchange (ETDEWEB)

    Cizelj, L.; Sorsek, I. [Jozef Stefan Institute, Ljubljana (Slovenia); Riesch-Oppermann, H. [Forschungszentrum Karlsruhe (Germany)

    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.

  7. Recent safety issues concerning steam generators in France and their analysis by IRSN

    Energy Technology Data Exchange (ETDEWEB)

    Sollier, T.; Le Calvar, M.; Balestreri, F.; Mermaz, F. [Inst. de Radioprotection ed de Surete Nucleaaire (IRSN) (France)

    2009-07-01

    In France between 2004 and 2008, there were recurrent safety issues concerning the operation of Steam Generators (SGs). Among these issues, at least three are generic to the EDF Nuclear Power Plant (NPP) fleet: In 2004, 2005 and 2006, a total of three primary to secondary leaks occurred at Cruas NPP. The root cause of these leaks was a modification of the thermal-hydraulic condition of the SG due to a heavy build-up of oxide deposits at the flow holes of the quatrefoil-shaped Tube Support Plates (TSPs). The clogging of the TSPs, meant that the water/steam flow accelerated at the U-bend location and that tubes were subjected to high cycle fatigue near the uppermost TSPs due to flow-induced vibration. For each unscheduled outage, the origin of the leaks was a circumferential fatigue crack located at the upper edge of the uppermost TSP; In 2008, a primary to secondary leak occurred at Fessenheim NPP. The source of the leak was a circumferential crack located at the edge of the uppermost TSP at approximately the same location where cracks were found on Cruas Units. However, the SGs of Fessenheim Unit 2 have circular flow holes without significant flow section reduction due to oxide deposits. The root cause of the event was determined to be fluid-elastic instability in the U-Bend for a tube not supported by an Anti-Vibration Bar (AVB). The AVB position in the tube bundle deviated from the manufacturing design, something which affects a large number of SGs in France; In 2008, a plug failure was observed at Saint Alban NPP. A plug was propelled from the hot to the cold leg during the primary coolant circuit hydrotest. The plugging operation had been performed before the hydrotest. In this paper, IRSN presents its technical analysis of these events. It includes the SG secondary side water conditioning operation, the non-destructive testing methods in relation to the clogging-rate evaluation and tube integrity assessment, and the mechanical issues due to tube vibration

  8. Development of Design Criteria for Fluid Induced Structural Vibration in Steam Generators and Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Catton, Ivan; Dhir, Vijay K.; Alquaddoomi, O.S.; Mitra, Deepanjan; Adinolfi, Pierangelo

    2004-03-26

    OAK-B135 Flow-induced vibration in heat exchangers has been a major cause of concern in the nuclear industry for several decades. Many incidents of failure of heat exchangers due to apparent flow-induced vibration have been reported through the USNRC incident reporting system. Almost all heat exchangers have to deal with this problem during their operation. The phenomenon has been studied since the 1970s and the database of experimental studies on flow-induced vibration is constantly updated with new findings and improved design criteria for heat exchangers. In the nuclear industry, steam generators are often affected by this problem. However, flow-induced vibration is not limited to nuclear power plants, but to any type of heat exchanger used in many industrial applications such as chemical processing, refrigeration and air conditioning. Specifically, shell and tube type heat exchangers experience flow-induced vibration due to the high velocity flow over the tube banks. Flow-induced vibration in these heat exchangers leads to equipment breakdown and hence expensive repair and process shutdown. The goal of this research is to provide accurate measurements that can help modelers to validate their models using the measured experimental parameters and thereby develop better design criteria for avoiding fluid-elastic instability in heat exchangers. The research is divided between two primary experimental efforts, the first conducted using water alone (single phase) and the second using a mixture of air or steam and water as the working fluid (two phase). The outline of this report is as follows: After the introduction to fluid-elastic instability, the experimental apparatus constructed to conduct the experiments is described in Chapter 2 along with the measurement procedures. Chapter 3 presents results obtained on the tube array and the flow loop, as well as techniques used in data processing. The project performance is described and evaluated in Chapter 4 followed by

  9. Next Generation Nuclear Plant Steam Generator and Intermediate Heat Exchanger Materials Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright

    2010-09-01

    DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Today’s high-temperature alloys and associated ASME Codes for reactor applications are approved up to 760°C. However, some primary system components, such as the Intermediate Heat Exchanger (IHX) for the NGNP will require use of materials that can withstand higher temperatures. The thermal, environmental, and service life conditions of the NGNP will make selection and qualification of some high-temperature materials a significant challenge. Examples include materials for the core barrel and core internals, such as the control rod sleeves. The requirements of the materials for the IHX are among the most demanding. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. A number of solid solution strengthened nickel based alloys have been considered for

  10. Corrosion behaviour of a stream generator tube material in simulated steam generator feedwater containing chlorides and sulphates

    Energy Technology Data Exchange (ETDEWEB)

    Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Sirkiae, P.; Yliniemi, K. [VTT Manufacturing Technology, Espoo (Finland); Buddas, T.; Halin, M.; Tompuri, K. [Fortum Power and Heat Oy, Loviisa Power Plant (Finland)

    2002-07-01

    The goal of the present work has been to assess the effect of relatively high concentrations of anionic impurities (Cl{sup -}, SO{sub 4}{sup 2-}) on the corrosion behaviour of Ti-stabilised stainless steel SG tubes in simulated steam generator feed-water. The main observations of this work can be summarised as follows: Sulphate ions seem to be more aggressive than chloride ions towards the primary passive film on 08X18H10T stainless steel. The results may indicate that it is more important to have a low concentration of sulphate ions than of chloride ions in secondary side water when the effects of chemical conditions on tube degradation are considered. The presence of chloride ions seems to weaken the detrimental effect of sulphate ions on the stability of oxide films growing on 08X18H10T stainless steel. No localised corrosion features of 08X18H10T stainless steel were detected in the voltammetric and impedance measurements in solutions containing up to 5000 ppb sulphates, chlorides or both of the anions. (authors)

  11. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  12. Development of the advanced CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Na, Y. H.; Lee, S. Y.; Choi, J. H.; Lee, B. C.; Kim, S. N.; Jo, C. H.; Paik, J. S.; On, M. R.; Park, H. S.; Kim, S. R. [Korea Electric Power Co., Taejon (Korea, Republic of)

    1997-07-01

    The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs.

  13. CFX Analysis of the CANDU Moderator Thermal-Hydraulics in the Stern Lab. Test Facility

    Science.gov (United States)

    Kim, Hyoung Tae

    2014-06-01

    A numerical calculation with the commercial CFD code CFX is conducted for a test facility simulating the CANDU moderator thermal-hydraulics. Two kinds of moderator thermal-hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the CANDU moderator circulating vessel, which is called a Calandria, housing a matrix of horizontal rod bundles simulating the Calandria tubes. The first of these tests is the pressure drop measurement of a cross flow in the horizontal rod bundles. The other is the local temperature measurement on the cross section of the horizontal cylinder vessel simulating the Calandria. In the present study the full geometric details of the Calandria are incorporated in the grid generation of the computational domain to which the boundary conditions for each experiment are applied. The numerical solutions are reviewed and compared with the available test data.

  14. Experience of steam generator tube examination in the hot laboratory of EDF: analysis of recent events concerning the secondary side

    Energy Technology Data Exchange (ETDEWEB)

    Thebault, Y.; Bouvier, O. de; Boccanfuso, M.; Coquio, N.; Barbe, V.; Molinie, E. [EDF-DIN-CEIDRE (France)

    2011-07-01

    Until 2010, more than 60 steam generator (SG) tubes have been removed and analysed in the EDF hot laboratory of CEIDRE/Chinon. This article is particularly related to three recent events that lead to the extraction of several tubes dedicated to laboratory destructive examinations. The first event that constitutes a first occurrence on the EDF Park, concerns the detection of a circumferential crack on the external surface of a tube located at tube support plate elevation. After this observation, several tubes have been extracted from Bugey 3 and Fessenheim 2 nuclear power plants with steam generators equipped with 600 MA bundle. The other two events concern the consequences of chemical cleaning of the tube bundle steam generators. The examples chosen are from Cruas 4 et Chinon B2 units whose tubes were extracted following non destructive testing performed immediately after or at the completion of cycle following the chemical cleaning. In the case of Cruas 4, Eddy Current Testing (ET) were performed for requalification of steam Generators after chemical cleaning. They allowed the detection of an indication located at the bottom of tube for a large number of tubes; the ET signal was similar to that corresponding to 'deposit' corrosion. Moreover, inspections of Chinon-B2 SGs at the end of the operation cycle following the chemical cleaning, showed the presence of conductor deposits at the bottom of some tubes. The first part of this document presents the major results of laboratory examinations of the pulled tubes of Bugey 3 and Fessenheim 2 and their analysis. Hypothesis concerning damage mechanisms of the tubes are also proposed. The second part of the paper relates the results of the laboratory examinations of the pulled tubes of Cruas 4 and Chinon B 2 after chemical cleaning and their analysis. (authors)

  15. Effect of steam generator configuration in a loss of the RHR during mid-loop operation at PKL facility

    Energy Technology Data Exchange (ETDEWEB)

    Villanueva, J. F.; Carlos, S.; Martorell, S.; Sanchez, F. [Dpto. Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino Vera s/n, 46022 Valencia (Spain)

    2012-07-01

    The loss of the residual heat removal system in mid-loop conditions may occur with a non-negligible contribution to the plant risk, so the analysis of the accidental sequences and the actions to mitigate the accident are of great interest in shutdown conditions. In order to plan the appropriate measures to mitigate the accident is necessary to understand the thermal-hydraulic processes following the loss of the residual heat removal system during shutdown. Thus, transients of this kind have been simulated using best-estimate codes in different integral test facilities and compared with experimental data obtained in different facilities. In PKL (Primaerkreislauf-Versuchsanlage, primary coolant loop test facility) test facility different series of experiments have been undertaken to analyze the plant response in shutdown. In this context, the E3 and F2 series consist of analyzing the loss of the residual heat removal system with a reduced inventory in the primary system. In particular, the experiments were developed to investigate the influence of the steam generators secondary side configuration on the plant response, what involves the consideration of different number of steam generators filled with water and ready for activation, on the heat transfer mechanisms inside the steam generators U-tubes. This work presents the results of such experiments calculated using, RELAP5/Mod 3.3. (authors)

  16. A Differential-Algebraic Model for the Once-Through Steam Generator of MHTGR-Based Multimodular Nuclear Plants

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2015-01-01

    Full Text Available Small modular reactors (SMRs are those fission reactors whose electrical output power is no more than 300 MWe. SMRs usually have the inherent safety feature that can be applicable to power plants of any desired power rating by applying the multimodular operation scheme. Due to its strong inherent safety feature, the modular high temperature gas-cooled reactor (MHTGR, which uses helium as coolant and graphite as moderator and structural material, is a typical SMR for building the next generation of nuclear plants (NGNPs. The once-through steam generator (OTSG is the basis of realizing the multimodular scheme, and modeling of the OTSG is meaningful to study the dynamic behavior of the multimodular plants and to design the operation and control strategy. In this paper, based upon the conservation laws of mass, energy, and momentum, a new differential-algebraic model for the OTSGs of the MHTGR-based multimodular nuclear plants is given. This newly-built model can describe the dynamic behavior of the OTSG in both the cases of providing superheated steam and generating saturated steam. Numerical simulation results show the feasibility and satisfactory performance of this model. Moreover, this model has been applied to develop the real-time simulation software for the operation and regulation features of the world first underconstructed MHTGR-based commercial nuclear plant—HTR-PM.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Homogenisation method for the dynamic analysis of a complete nuclear steam generator with fluid-structure interaction

    Energy Technology Data Exchange (ETDEWEB)

    Sigrist, Jean-Francois [DCNS Propulsion-DI/STS, 44620 La Montagne (France)], E-mail: jean-francois.sigrist@dcn.fr; Broc, Daniel [CEA Saclay-DEMT/EMSI, 91191 Gif-sur-Yvette (France)

    2008-09-15

    The present paper deals with the dynamic analysis of a steam generator tube bundle with fluid-structure interaction modelling. As the coupled fluid-structure problem involves a huge number of degrees of freedom to account for the tube displacements and the fluid pressure evolutions, classical coupled method cannot be applied for industrial studies. In the present case, the three-dimensional fluid-structure problem is solved with an homogenisation method, which has been previously exposed and successfully validated for FSI modelling in a nuclear reactor [Sigrist, J.F., Broc, D., 2007a. Homogenisation method for the modal analysis of a nuclear reactor with internal structures modelling and fluid-structure interaction coupling. Nuclear Engineering and Design 237, 431-440]. Formulation of the homogenisation method for general two- and three-dimensional cases is exposed in the paper. Application to a simplified, however representative, model of an actual industrial nuclear component (steam generator) is proposed. The problem modelling, which includes tube bundle, primary and secondary fluids and pressure vessel, is performed with an engineering finite element code in which the homogenisation technique has been implemented. From the practical point of view, the analysis highlights the major fluid-structure interaction effects on the dynamic behaviour of the steam generator; from the theoretical point of view, the study demonstrates the efficiency of the homogenisation method for periodic fluid-structure problems modelling in industrial configurations.

  19. Experimental residual stress evaluation of hydraulic expansion transitions in Alloy 690 steam generator tubing

    Energy Technology Data Exchange (ETDEWEB)

    McGregor, R.; Doherty, P. [Babcock and Wilcox International, Cambridge, Ontario (Canada); Hornbach, D. [Lambda Research Inc., Cincinnati, OH (United States); Abdelsalam, U. [McMaster Univ., Hamilton, Ontario (Canada)

    1995-12-31

    Nuclear Steam Generator (SG) service reliability and longevity have been seriously affected worldwide by corrosion at the tube-to-tubesheet joint expansion. Current SG designs for new facilities and replacement projects enhance corrosion resistance through the use of advanced tubing materials and improved joint design and fabrication techniques. Here, transition zones of hydraulic expansions have undergone detailed experimental evaluation to define residual stress and cold-work distribution on and below the secondary-side surface. Using X-ray diffraction techniques, with supporting finite element analysis, variations are compared in tubing metallurgical condition, tube/pitch geometry, expansion pressure, and tube-to-hole clearance. Initial measurements to characterize the unexpanded tube reveal compressive stresses associated with a thin work-hardened layer on the outer surface of the tube. The gradient of cold-work was measured as 3% to 0% within .001 inch of the surface. The levels and character of residual stresses following hydraulic expansion are primarily dependent on this work-hardened surface layer and initial stress state that is unique to each tube fabrication process. Tensile stresses following expansion are less than 25% of the local yield stress and are found on the transition in a narrow circumferential band at the immediate tube surface (< .0002 inch/0.005 mm depth). The measurements otherwise indicate a predominance of compressive stresses on and below the secondary-side surface of the transition zone. Excellent resistance to SWSCC initiation is offered by the low levels of tensile stress and cold-work. Propagation of any possible cracking would be deterred by the compressive stress field that surrounds this small volume of tensile material.

  20. A Fundamental study of remedial technology development to prevent stress corrosion cracking of steam generator tubing

    Energy Technology Data Exchange (ETDEWEB)

    Park, In Gyu; Lee, Chang Soon [Sunmoon University, Asan (Korea)

    1998-04-01

    Most of the PWR Steam generators with tubes in Alloy 600 alloy are affected by Stress Corrosion Cracking, such as PWSCC(Primary Water Stress Corrosion Cracking) and ODSCC(Outside Diameter Stress Corrosion Cracking). This study was undertaken to establish the background for remedial technology development to prevent SCC. in the report are included the following topics: (1) General: (i) water chemistry related factors, (ii) Pourbaix(Potential-pH) Diagram, (iii) polarization plot, (iv) corrosion mode of Alloy 600, 690, and 800, (v) IGA/SCC growth rate, (vi) material suspetibility of IGA/SCC, (vii) carbon solubility of Alloy 600 (2) Microstructures of Alloy 600 MA, Alloy 600 TT, Alloy 600 SEN Alloy 690 TT(Optical, SEM, and TEM) (3) Influencing factors for PWSCC initiation rate of Alloy 600: (i) microstructure, (ii) water chemistry(B, Li), (iii) temperature, (iv) plastic deformation, (v) stress relief annealing (4) Influencing factors for PWSCC growth rate of Alloy 600: (i) water chemistry(B, Li), (ii) Scott Model, (iii) intergranular carbide, (iv) temperature, (v) hold time (5) Laboratory conditions for ODSCC initiation rate: 1% NaOH, 316 deg C; 1% NaOH, 343 deg C; 50% NaOH, 288 deg C; 10% NaOH, 302 deg C; 10% NaOH, 316 deg C; 50% NaOH, 343 deg C (6) Sludge effects for ODSCC initiation rate: CuO, Cr{sub 2}O{sub 3}, Fe{sub 3}O{sub 4} (7) Influencing factors for PWSCC growth rate of Alloy 600: (i) Caustic concentration effect, (ii) carbonate addition effect (8) Sulfate corrosion: (i) sulfate ratio and pH effect, (ii) wastage rate of Alloy 600 and Alloy 690 (9) Crevice corrosion: (i) experimental setup for crevice corrosion, (ii) organic effect, (iii) (Na{sub 2}SO{sub 4} + NaOH) effect (10) Remedial measures for SCC: (i) Inhibitors, (ii) ZnO effect. (author). 30 refs., 174 figs., 51 tabs.

  1. Development of countermeasure against scale deposition at steam generators of PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Domae, M.; Miyajima, K.; Hirano, H. [Surface Science Dept., Central Research Inst. of Electric Power Industry (Japan); Kushida, H. [Nuclear Power Engineering Dept., Tokyo Electric Power Environmental Engineering Co., Inc. (Japan)

    2002-07-01

    Scale deposition has occurred at steam generators of several PWRs. The scale deposition may lead to reduction of flow rate of coolant, deterioration of heat exchanging efficiency and so on. These phenomena affect plant operation performance. Thus, elucidation of the mechanism of the scale deposition and some effective countermeasure are required. In CRIEPI (Central Research Institute of Electric Power Industry), the scale deposition is studied from two aspects: fluid dynamics and water chemistry. Concerning the water chemistry, we think that electro-kinetic behavior of scale, that is, metal oxides is of great importance. The final goal of the water chemical approach is to evaluate electro-kinetic potential (zeta potential) of metal oxides such as magnetite (Fe{sub 3}O{sub 4}) and hematite (Fe{sub 2}O{sub 3}), and to develop some countermeasure of the scale deposition based on the electro-kinetic data. As a first step, the zeta potential of 25 {mu}m Fe{sub 3}O{sub 4} particles was measured by the streaming potential method at room temperature, and effect of dispersant addition was studied. The dispersants examined were poly-acrylic acid (PAA, M{sub w} {proportional_to} 25,000) and polyvinylpyrrolidone (PVP, M{sub w} {proportional_to} 40,000). It has been found that the addition of PAA of more than 10 ppm lowers the zeta potentials by 5 - 15 mV in whole pH range, and that the addition of PVP of more than 10 ppm reduces absolute value of the zeta potentials. (authors)

  2. Study of thermal influence on tubes due to sodium-water reactions in LMFBR steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Hamada, H.; Kurihara, A.; Nishimura, M. [Japan Nuclear Cycle Development Institute, Ibaraki (Japan)

    2004-07-01

    A study of thermal influence on heat-transfer tubes in sodium-water reactions is carried out to evaluate the tube rupture due to overheating in the water leak accident of an LMFBR steam generator (SG). By assuming the sodium-water reaction jet to be a two-phase flow that consists of sodium and hydrogen, the heat-transfer characteristics are examined and a simple model of effective heat-transfer coefficient (HTC) is proposed for the safety evaluation of the SG. Comparison of the model with experimental data leads to the following conclusions: An upper limit exists in the HTC between reaction jet and tube wall, and it is equivalent in approximation to the HTC of single-phase sodium flow. The HTC can be written in simple form as functions of the HTC of single-phase sodium flow, void fraction and temperatures of sodium, hydrogen and tube wall. Hydrogen provides negligible heating effect, so that the apparent HTC would decrease with increase of the hydrogen temperature that can readily surpass that of sodium. The outer-surface temperature of tube wall would not rise so high beyond the temperature of sodium that is excellent in heat-transfer characteristics, even if tube wall is exposed to the high-temperature hydrogen. The transient heat conduction analysis with the mean value of the data can appropriately evaluate the outer-surface temperature of tube wall by the metallographic observation, while the analysis with the maximum value can conservatively evaluate the tube wall temperature. (authors)

  3. Evaluation of machine learning tools for inspection of steam generator tube structures using pulsed eddy current

    Science.gov (United States)

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

    2017-02-01

    Degradation of nuclear steam generator (SG) tubes and support structures can result in a loss of reactor efficiency. Regular in-service inspection, by conventional eddy current testing (ECT), permits detection of cracks, measurement of wall loss, and identification of other SG tube degradation modes. However, ECT is challenged by overlapping degradation modes such as might occur for SG tube fretting accompanied by tube off-set within a corroding ferromagnetic support structure. Pulsed eddy current (PEC) is an emerging technology examined here for inspection of Alloy-800 SG tubes and associated carbon steel drilled support structures. Support structure hole size was varied to simulate uniform corrosion, while SG tube was off-set relative to hole axis. PEC measurements were performed using a single driver with an 8 pick-up coil configuration in the presence of flat-bottom rectangular frets as an overlapping degradation mode. A modified principal component analysis (MPCA) was performed on the time-voltage data in order to reduce data dimensionality. The MPCA scores were then used to train a support vector machine (SVM) that simultaneously targeted four independent parameters associated with; support structure hole size, tube off-centering in two dimensions and fret depth. The support vector machine was trained, tested, and validated on experimental data. Results were compared with a previously developed artificial neural network (ANN) trained on the same data. Estimates of tube position showed comparable results between the two machine learning tools. However, the ANN produced better estimates of hole inner diameter and fret depth. The better results from ANN analysis was attributed to challenges associated with the SVM when non-constant variance is present in the data.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    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.

  5. The travesty of discarding used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ottensmeyer, P. [Univ. of Toronto, Toronto, Ontario (Canada)

    2016-09-15

    The current plan worldwide for virtually all used nuclear fuels is costly deep burial to attempt to isolate their long-term radiotoxicity permanently. Alternatively Canada's 50,000 tons spent CANDU fuel, of which only 0.74% of the heavy atoms have been fissioned to extract their energy, could supply 130 times more non-carbon energy using proven economical recycling and fast-neutron technologies. The result in this country alone would currently be the creation of $74 trillion of reliable electricity on demand without greenhouse gas emissions. It would avoid adding 475 billion tons CO{sub 2} to the atmosphere compared to the use of coal, to mitigate climate change. Worldwide recycling of stored spent nuclear fuel and replenishing with depleted uranium in fast-neutron reactors could avoid emitting over 20 trillion tons CO{sub 2}, or over six times the current total atmospheric CO{sub 2} content. As added bonus the long-term radiotoxicity of the used CANDU fuel is effectively eliminated, making a long-term deep geological repository unnecessary. Even the shorter-lived radioisotope fission products become valuable stable atoms and minerals that would fetch $3 million per ton. Such an alternative is certainly worth pursuing. (author)

  6. 百万千瓦级核电汽轮发电机组选型%The Selecion of the speed of 1000MW Nuclear Steam Turbine Generator

    Institute of Scientific and Technical Information of China (English)

    王雪松

    2001-01-01

    This paper is to show how to select the speed of 1000MW nuclear steam turbine generator forour country's next nuclear power plants in accordance with the developing trend of the nuclear steam turbine generator abroad as well as a comprehensive analysis and comparison of full speed nuclear steam turbine generator and half speed steam turbine generator at 1000MW.%通过对国外核汽轮发电机组发展趋势的分析和对百万千瓦全转速与半转速机组的综合分析比较,简要阐述广东继岭澳一期电站工程后百万千瓦级核电站汽轮发电机组的选型问题。

  7. Study of the vibrations induced by two-phase flow in steam generator: measurement of void fraction in a two-phase flow; Etude des vibrations induites dans les tubes de generateurs de vapeur: mesure du taux de vide dans un ecoulement diphasique

    Energy Technology Data Exchange (ETDEWEB)

    Sivault, S

    1998-07-01

    Two-phase flow can trigger vibration phenomena that are not well predicted by models like the homogeneous model. Concerning the steam generator of a Candu type reactor, these vibrations may lead to the failure of tubes. The coupling between thermo-hydraulic and vibration phenomena requires models that treat sliding between liquid and vapor phases. The purpose of this work is to study a series of experiments performed in a freon loop. These experiments simulate a two-phase flow through a bundle of tubes. Most estimations of vibratory parameters are based on the assumption of a uniform distribution of the void fraction. An optic probe has been used to measure the void fraction. The first part of this study is devoted to the processing of the response spectra given by the probe. The second part presents an estimation of the void fraction given by different models, a comparison between experimental and theoretical results allows to discuss their validity range. (A.C.) 6 refs.

  8. Evaluation of Efficiency in Steam Generator of C3 Power Plant at Cap Des Biches in Dakar

    Directory of Open Access Journals (Sweden)

    A. Kane

    2012-07-01

    Full Text Available The aim of the present study is to determine the efficiency of the power plant boiler steam C3 Cape deer and to evaluate the impact of it. We chose to calculate the return by the empirical formula of Martin which is based on two important parameters which are the temperature at the exit of the chimney and the ambient temperature. The calculation of these efficiencies allowed us to make comparative studies with data from the manufacturer and we have detected anomalies. On this basis we made a number of recommendations for improvement of the groups in normal, the application will lead to an optimization of the real exploitation of groups and a performance improvement. We gave various reasons for poor performance of steam generators and recommendations that can be used both in production efficiency on compliance with operating instructions. Solutions have been proposed after diagnosis with particular emphasis on compliance with operating instructions and maintenance schedule.

  9. Lead-induced stress-corrosion cracking of alloy 600 in plausible steam generator crevice environments

    Energy Technology Data Exchange (ETDEWEB)

    Wright, M.D. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Manolescu, A. [Ontario Hydro Technologies, Toronto, Ontario (Canada); Mirzai, M. [Ontario Hydro, Toronto, Ontario (Canada)

    1999-03-01

    Laboratory stress-corrosion cracking (SCC) test environments were developed to simulate crevice chemistries representative of Bruce Nuclear Generating Station A (BNPD A) steam generators (SGs); these test environments were used to determine the susceptibility of Alloy 600 to lead-induced SCC under plausible SG conditions. Test environments were based on plant SG hideout return data and analysis of removed tubes and deposits. Deviations from the normal near-neutral crevice pH environment were considered to simulate possible faulted excursion crevice chemistry and to bound the postulated crevice pH range of 3 to 9 (at temperature). The effect of lead contamination up to 1000 ppm, but with an emphasis on the 100- to 500-ppm range, was determined. SCC susceptibility was investigated using constant extension rate tensile (CERT) tests and encapsulated C-ring tests. CERT tests were performed at 305 degrees C on tubing representative of BNPD A SG U-bends. The C-ring test method allowed a wider test matrix, covering 3 temperatures (280 degrees C, 304 degrees C and 315 degrees C), 3 strain levels (0.2%, 2% and 4%), and tubing representative of U-bends plus tubing given a simulated stress relief to represent material at the tube sheet. The results of this test program confirmed that in the absence of lead contamination, cracking does not occur in these concentrated, 3.3 to 8.9 pH range, crevice environments. Also, it appears that the concentrated crevice environments suppress lead-induced cracking relative to that seen in all-volatile-treatment (AVT) water. For the (static) C-ring tests, lead-induced SCC was only produced in the near-neutral crevice environment and was more severe at 500 ppm than at 100 ppm PbO. This trend was also observed in CERT tests, but some cracking-grain boundary attack occurred in acidic (pH 3.3) and alkaline (pH 8.9) environments. The C-ring tests indicated that a certain amount of resistance to cracking was imparted by simulated stress relief of

  10. Optimization and implementation study of plutonium disposition using existing CANDU Reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Since early 1994, the Department of Energy has been sponsoring studies aimed at evaluating the merits of disposing of surplus US weapons plutonium as Mixed Oxide (MOX) fuel in existing commercial Canadian Pressurized Heavy Water reactors, known as CANDU`s. The first report, submitted to DOE in July, 1994 (the 1994 Executive Summary is attached), identified practical and safe options for the consumption of 50 to 100 tons of plutonium in 25 years in some of the existing CANDU reactors operating the Bruce A generating station, on Lake Huron, about 300 km north east of Detroit. By designing the fuel and nuclear performance to operate within existing experience and operating/performance envelope, and by utilizing existing fuel fabrication and transportation facilities and methods, a low cost, low risk method for long term plutonium disposition was developed. In December, 1995, in response to evolving Mission Requirements, the DOE requested a further study of the CANDU option with emphasis on more rapid disposition of the plutonium, and retaining the early start and low risk features of the earlier work. This report is the result of that additional work.

  11. Speciation of iodine (I-127) in the natural environment around Canadian CANDU sites

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, S.J.; Kotzer, T.G.; Chant, L.A

    2001-06-01

    In Canada, very little data is available regarding the concentrations and chemical speciation of iodine in the environment proximal and distal to CANDU Nuclear Power Generating Stations (NPGS). In the immediate vicinity of CANDU reactors, the short-lived iodine isotope {sup 131}I (t{sub 1/2} = 8.04 d), which is produced from fission reactions, is generally below detection and yields little information about the environmental cycling of iodine. Conversely, the fission product {sup 129}I has a long half-life (t{sub 1/2} = 1.57x10{sup 7} y) and has had other anthropogenic inputs (weapons testing, nuclear fuel reprocessing) other than CANDU over the past 50 years. As a result, the concentrations of stable iodine ({sup 127}I) have been used as a proxy. In this study, a sampling system was developed and tested at AECL's Chalk River Laboratories (CRL) to collect and measure the particulate and gaseous inorganic and organic fractions of stable iodine ({sup 127}I) in air and associated organic and inorganic reservoirs. Air, vegetation and soil samples were collected at CRL, and at Canadian CANDU Nuclear Power Generating Stations (NPGS) at OPG's (Ontario Power Generation) Pickering (PNGS) and Darlington NPGS (DNGS) in Ontario, as well as at NB Power's Pt. Lepreau NPGS in New Brunswick. The concentrations of particulate and inorganic iodine in air at CRL were extremely low, and were often found to be below detection. The concentrations are believed to be at this level because the sediments in the CRL area are glacial fluvial and devoid of marine ionic species, and the local atmospheric conditions at the sampling site are very humid. Concentrations of a gaseous organic species were comparable to worldwide levels. The concentrations of particulate and inorganic iodine in air were also found to be low at PNGS and DNGS, which may be attributed to reservoir effects of the large freshwater lakes in southern Ontario, which might serve to dilute the atmospheric iodine

  12. Advanced CANDU reactor pre-licensing progress

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.K.; West, J.; Snell, V.G.; Ion, R.; Archinoff, G.; Xu, C. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca

    2005-07-01

    The Advanced CANDU Reactor (ACR) is an evolutionary advancement of the current CANDU 6 reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The Canadian Nuclear Safety Commission (CNSC) staff are currently reviewing the ACR design to determine whether, in their opinion, there are any fundamental barriers that would prevent the licensing of the design in Canada. This CNSC licensability review will not constitute a licence, but is expected to reduce regulatory risk. The CNSC pre-licensing review started in September 2003, and was focused on identifying topics and issues for ACR-700 that will require a more detailed review. CNSC staff reviewed about 120 reports, and issued to AECL 65 packages of questions and comments. Currently CNSC staff is reviewing AECL responses to all packages of comments. AECL has recently refocused the design efforts to the ACR-1000, which is a larger version of the ACR design. During the remainder of the pre-licensing review, the CNSC review will be focused on the ACR-1000. AECL Technologies Inc. (AECLT), a wholly-owned US subsidiary of AECL, is engaged in a pre-application process for the ACR-700 with the US Nuclear Regulatory Commission (USNRC) to identify and resolve major issues prior to entering a formal process to obtain standard design certification. To date, the USNRC has produced a Pre-Application Safety Assessment Report (PASAR), which contains their reviews of key focus topics. During the remainder of the pre-application phase, AECLT will address the issues identified in the PASAR. Pursuant to the bilateral agreement between AECL and the Chinese nuclear regulator, the National Nuclear Safety Administration (NNSA) and its Nuclear Safety Center (NSC), NNSA/NSC are reviewing the ACR in seven focus areas. The review started in September 2004, and will take three years. The main objective of the review is to determine how the ACR complies

  13. Thermal gain of CHP steam generator plants and heat supply systems

    Science.gov (United States)

    Ziganshina, S. K.; Kudinov, A. A.

    2016-08-01

    Heating calculation of the surface condensate heat recovery unit (HRU) installed behind the BKZ-420-140 NGM boiler resulting in determination of HRU heat output according to fire gas value parameters at the heat recovery unit inlet and its outlet, heated water quantity, combustion efficiency per boiler as a result of installation of HRU, and steam condensate discharge from combustion products at its cooling below condensing point and HRU heat exchange area has been performed. Inspection results of Samara CHP BKZ-420-140 NGM power boilers and field tests of the surface condensate heat recovery unit (HRU) made on the bimetal calorifier base KCk-4-11 (KSk-4-11) installed behind station no. 2 Ulyanovsk CHP-3 DE-10-14 GM boiler were the basis of calculation. Integration of the surface condensation heat recovery unit behind a steam boiler rendered it possible to increase combustion efficiency and simultaneously decrease nitrogen oxide content in exit gases. Influence of the blowing air moisture content, the excess-air coefficient in exit gases, and exit gases temperature at the HRU outlet on steam condensate amount discharge from combustion products at its cooling below condensing point has been analyzed. The steam condensate from HRU gases is offered as heat system make-up water after degasification. The cost-effectiveness analysis of HRU installation behind the Samara CHP BKZ-420-140 NGM steam boiler with consideration of heat energy and chemically purified water economy has been performed. Calculation data for boilers with different heat output has been generalized.

  14. Inspection of ferromagnetic support structures from within alloy 800 steam generator tubes using pulsed eddy current

    Science.gov (United States)

    Buck, Jeremy Andrew

    Nondestructive testing is a critical aspect of component lifetime management. Nuclear steam generator (SG) tubes are the thinnest barrier between irradiated primary heat transport system and the secondary heat transport system, whose components are not rated for large radiation fields. Conventional eddy current testing (ECT) and ultrasonic testing are currently employed for inspecting SG tubes, with the former doing most inspections due to speed and reliability based on an understanding of how flaws affect coil impedance parameters when conductors are subjected to harmonically induced currents. However, when multiple degradation modes are present simultaneously near ferromagnetic materials, such as tube fretting, support structure corrosion, and magnetite fouling, ECT reliability decreases. Pulsed eddy current (PEC), which induces transient eddy currents via square wave excitation, has been considered in this thesis to simultaneously examine SG tube and support structure conditions. An array probe consisting of a central driver, coaxial with the tube, and an array of 8 sensing coils, was used in this thesis to perform laboratory measurements. The probe was delivered from the inner diameter (ID) of the SG tube, where support hole diameter, tube frets, and 2D off-centering were varied. When considering two variables simultaneously, scores obtained from a modified principal components analysis (MPCA) were sufficient for parameter extraction. In the case of hole ID variation with two dimensional tube off-centering (three parameters), multiple linear regression (MLR) of the MPCA scores provided good estimates of parameters. However, once a fourth variable, outer diameter tube frets, was introduced, MLR proved insufficient. Artificial neural networks (ANNs) were investigated in order to perform pattern recognition on the MPCA scores to simultaneously extract the four measurement parameters from the data. All models throughout this thesis were created and validated using

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

    Science.gov (United States)

    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.

  16. Lead-induced SCC of alloy 600 in plausible steam generator crevice environments

    Energy Technology Data Exchange (ETDEWEB)

    Wright, M.D. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Manolescu, A. [Ontario Hydro Technologies, Toronto, Ontario (Canada); Mirzai, M. [Ontario Hydro, Toronto, Ontario (Canada)

    1998-07-01

    Laboratory stress corrosion cracking (SCC) test environments developed to simulate representative BNGS-A steam generator (SG) crevice chemistries have been used to determine the susceptibility of Alloy 600 to lead-induced SCC under plausible SG conditions. Test environments were based on plant SG hideout return data and analysis of removed tubes and deposits. Deviations from the normal near neutral crevice pH environment were considered to simulate possible faulted excursion crevice chemistry and to bound the postulated crevice pH range of 3-9 (at temperature). The effect of lead contamination up to 1000 ppm, but with an emphasis on the 100 to 500 ppm range, was determined. SCC susceptibility was investigated using constant extension rate tensile (CERT) tests and encapsulated C-ring tests. CERT tests were performed at 305 degrees C on tubing representative of BNGS-A SG U-bends. The C-ring test method allowed a wider test matrix covering three temperatures (280, 304 and 315 degrees C), three strain levels (0.2%, 2% and 4%) and tubing representative of U-bends plus tubing given a simulated stress relief to represent material at the tubesheet. The results of this test program confirmed that in the absence of lead contamination, cracking does not occur in these concentrated, 3.3 to 8.9 pH range, crevice environments. Also, it appears that the concentrated crevice environments suppress lead-induced cracking relative to that seen in all-volatile-treatment (AVT) water. For the (static) C-ring tests, lead-induced SCC was only produced in the near-neutral crevice environment and was more severe at 500 ppm than 100 ppm PbO. This trend was also observed in CERT tests but some cracking/grain boundary attack occurred in acidic (pH 3.3) and alkaline (pH 8.9) environments. The C-ring tests indicated that a certain amount of resistance to cracking was imparted by simulated stress relief of the tubing. This heat treatment, confirmed to have resulted in sensitization, promoted

  17. Leak detection capability in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Azer, N.; Barber, D.H.; Boucher, P.J. [and others

    1997-04-01

    This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis.

  18. An Isothermal Steam Expander for an Industrial Steam Supplying System

    Directory of Open Access Journals (Sweden)

    Chen-Kuang Lin

    2015-01-01

    Full Text Available Steam is an essential medium used in the industrial process. To ensure steam quality, small and middle scale boilers are often adopted. However, because a higher steam pressure (compared to the necessary steam pressure is generated, the boiler’s steam pressure will be reduced via a pressure regulator before the steam is directed through the process. Unfortunately, pressure is somewhat wasted during the reducing process. Therefore, in order to promote energy efficiency, a pressure regulator is replaced by a steam expander. With this steam expander, the pressure will be transformed into mechanical energy and extracted during the expansion process. A new type of isothermal steam expander for an industrial steam supplying system will be presented in the paper. The isothermal steam expander will improve the energy efficiency of a traditional steam expander by replacing the isentropic process with an isothermal expansion process. With this, steam condensation will decrease, energy will increase, and steam quality will be improved. Moreover, the mathematical model of the isothermal steam expander will be established by using the Schmidt theory, the same principle used to analyze Stirling engines. Consequently, by verifying the correctness of the theoretical model for the isothermal steam expander using experimental data, a prototype of 100 c.c. isothermal steam expander is constructed.

  19. Development of an Integrity Assessment Procedure for CANDU Pressure Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Han Sub [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The pressure tubes used in a CANDU reactor are made from Zr-2.5Nb. During service the pressure tubes operate at temperatures between about 150 and 310 .deg. C, and with variable coolant pressures up to 11MPa corresponding to hoop stress of up to 130MPa. The maximum flux of fast neutrons (E>1MeV) from the fuel is about 4X10{sup 17}nm{sup -2}{sub s}{sup -1}. The pressure tubes are exposed to very severe degradation environment. The aging degradation of the pressure tubes are summarized as below. - Geometric deformation; axial elongation, diametric creep, and wall thinning. - Deuterium uptake; some fraction of the deuterium generated by the corrosion of pressure tubes is absorbed into the pressure tubes. Total equivalent hydrogen content in the pressure tube is the sum of the initial hydrogen content before operation and the deuterium uptake during operation. High concentration of hydrogen inside the pressure tubes makes the metal susceptible to Delayed Hydride Cracking. The DHC is a degradation mechanism of prime importance for CANDU pressure tubes. Mechanical properties, in particular fracture toughness, are deteriorated by high concentration of dissolved hydrogen. - Flaws; volumetric flaws are generated during operation. Wear scars by debris fretting, and bearing pad fretting are common. These volumetric flaws can be a site of crack initiation by fatigue or DHC. Cracks can propagate by DHC or fatigue crack propagation if conditions are met. - Material properties degradation; mechanical properties are affected by neutron irradiation. Yield strength and tensile strength are increased, and fracture toughness is deteriorated. The susceptibility to DHC is also affected. The integrity assessment of the pressure tube is a procedure to determine if the risk of pressure tube failure is controlled to maintain acceptably low. CSA N285.4 and 285.8 are two important guidelines regarding the integrity of pressure tubes. N285.4 is to guide in-service inspection, and N285

  20. Cost efficiency of Japanese steam power generation companies: A Bayesian comparison of random and fixed frontier models

    Energy Technology Data Exchange (ETDEWEB)

    Assaf, A. George [Isenberg School of Management, University of Massachusetts-Amherst, 90 Campus Center Way, Amherst 01002 (United States); Barros, Carlos Pestana [Instituto Superior de Economia e Gestao, Technical University of Lisbon, Rua Miguel Lupi, 20, 1249-078 Lisbon (Portugal); Managi, Shunsuke [Graduate School of Environmental Studies, Tohoku University, 6-6-20 Aramaki-Aza Aoba, Aoba-Ku, Sendai 980-8579 (Japan)

    2011-04-15

    This study analyses and compares the cost efficiency of Japanese steam power generation companies using the fixed and random Bayesian frontier models. We show that it is essential to account for heterogeneity in modelling the performance of energy companies. Results from the model estimation also indicate that restricting CO{sub 2} emissions can lead to a decrease in total cost. The study finally discusses the efficiency variations between the energy companies under analysis, and elaborates on the managerial and policy implications of the results. (author)

  1. Boiler and steam generator corrosion: Fossil fuel power plants. (Latest citations from the NTIS bibliographic database). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-11-01

    The bibliography contains citations concerning corrosion effects, mechanisms, detection, and inhibition in fossil fuel fired boilers. Fluidized bed combustors and coal gasification are included in the applications. The citations examine hot corrosion, thermal mechanical degradation, and intergranular oxidation corrosion studies performed on the water side and hot gas side of heat exchanger tubes and support structures. Coatings and treatment of material to inhibit corrosion are discussed. Corrosion affecting nuclear powered steam generators is examined in a separate bibliography. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  2. Steam generation unit in a simple version of biomass based small cogeneration unit

    Directory of Open Access Journals (Sweden)

    Sornek Krzysztof

    2014-01-01

    Full Text Available The organic Rankine cycle (ORC is a very promising process for the conversion of low or medium temperature heat to electricity in small and micro scale biomass powered systems. Classic ORC is analogous to Clausius–Rankine cycle in a steam power plant, but instead of water it uses low boiling, organic working fluids. Seeking energy and economical optimization of biomass-based ORC systems, we have proposed some modifications e.g. in low boiling fluid circuit construction. Due to the fact that the operation of a micro steam turbine is rather inefficient from the technical and economic point of view, a specially modified air compressor can be used as a steam piston engine. Such engine should be designed to work at low pressure of the working medium. Studies regarding the first version of the prototype installation were focused on the confirmation of applicability of a straw boiler in the prototype ORC power system. The results of the previous studies and the studies described in the paper (on the new cogeneration unit confirmed the high potential of the developed solution. Of course, many further studies have to be carried out.

  3. Boxberg III-2 x 500 MW units: Refurbishing and environmental protection measures on the 815 T/H steam generator of works II in Boxberg Power Station

    Energy Technology Data Exchange (ETDEWEB)

    Cossman, R.; Fritz, M.; Bauchmueller, R. [L& C Steinmueller GmbH, Gummersbach (Germany)

    1995-12-01

    The object of the upgrading measures on the steam generators is: (1) To comply with the requirements of the German antipollution law, which imposes a permissible NO{sub x} content in the flue gas of less than 200 Mg/m{sup 3} STP and a CO content of less than 250 Mg/m{sup 3} STP. (2) To increase the boiler efficiency and availability and the efficiency of the water/steam cycle.

  4. Types and analysis of defects in welding junctions of the header to steam generator shells on power-generating units with VVER-1000

    Science.gov (United States)

    Ozhigov, L. S.; Voevodin, V. N.; Mitrofanov, A. S.; Vasilenko, R. L.

    2016-10-01

    Investigation objects were metal templates, which were cut during the repair of welding junction no. 111 (header to the steam generator shell) on a power-generating unit with VVER-1000 of the South-Ukraine NPP, and substances of mud depositions collected from walls of this junction. Investigations were carried out using metallography, optical microscopy, and scanning electron microscopy with energy dispersion microanalysis by an MMO-1600-AT metallurgical microscope and a JEOL JSM-7001F scanning electron microscope with the Shottky cathode. As a result of investigations in corrosion pits and mud depositions in the area of welding junction no. 111, iron and copper-enriched particles were revealed. It is shown that, when contacting with the steel header surface, these particles can form microgalvanic cells causing reactions of iron dissolution and the pit corrosion of metal. Nearby corrosion pits in metal are microcracks, which can be effect of the stress state of metal under corrosion pits along with revealed effects of twinning. The hypothesis is expressed that pitting corrosion of metal occurred during the first operation period of the power-generating unit in the ammonia water chemistry conditions (WCC). The formation of corrosion pits and nucleating cracks from them was stopped with the further operation under morpholine WCC. The absence of macrocracks in metal of templates verifies that, during operation, welding junction no. 111 operated under load conditions not exceeding the permissible ones by design requirements. The durability of the welding junction of the header to the steam generator shell significantly depends on the technological schedule of chemical cleaning and steam generator shut-down cooling.

  5. Review of Dissimilar Metal Welding for the NGNP Helical-Coil Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    John N. DuPont

    2010-03-01

    The U.S. Department of Energy (DOE) is currently funding research and development of a new high temperature gas cooled reactor (HTGR) that is capable of providing high temperature process heat for industry. The steam generator of the HTGR will consist of an evaporator economizer section in the lower portion and a finishing superheater section in the upper portion. Alloy 800H is expected to be used for the superheater section, and 2.25Cr 1Mo steel is expected to be used for the evaporator economizer section. Dissimilar metal welds (DMW) will be needed to join these two materials. It is well known that failure of DMWs can occur well below the expected creep life of either base metal and well below the design life of the plant. The failure time depends on a wide range of factors related to service conditions, welding parameters, and alloys involved in the DMW. The overall objective of this report is to review factors associated with premature failure of DMWs operating at elevated temperatures and identify methods for extending the life of the 2.25Cr 1Mo steel to alloy 800H welds required in the new HTGR. Information is provided on a variety of topics pertinent to DMW failures, including microstructural evolution, failure mechanisms, creep rupture properties, aging behavior, remaining life estimation techniques, effect of environment on creep rupture properties, best practices, and research in progress to improve DMW performance. The microstructure of DMWs in the as welded condition consists of a sharp chemical concentration gradient across the fusion line that separates the ferritic and austenitic alloys. Upon cooling from the weld thermal cycle, a band of martensite forms within this concentration gradient due to high hardenability and the relatively rapid cooling rates associated with welding. Upon aging, during post weld heat treatment (PWHT), and/or during high temperature service, C diffuses down the chemical potential gradient from the ferritic 2.25Cr 1Mo steel

  6. Application of Spherical Steam Accumulator in the Saturated Steam Power Generation Technology of Steel Plants%球形蒸汽蓄能器在钢铁厂饱和蒸汽发电技术中的应用

    Institute of Scientific and Technical Information of China (English)

    杨学友; 祝百东; 周春丽

    2014-01-01

    介绍了某钢铁公司回收厂区富余饱和蒸汽进行发电的方案,着重介绍了球形蒸汽蓄能器的原理及特点,并与卧式筒形蓄热器进行对比。球形蒸汽蓄能器具有占地少、投资省、系统简化、蒸汽含水率低等优点,推荐在钢铁厂饱和蒸汽回收利用中广泛采用球形蒸汽蓄能器。%The plan for recovering surplus steam in a steel plant to generate power is introduced focusing on the principle and characteristics of spherical steam accumulator, which is compared with horizontal cylinder-type accumulator. Spherical steam accumulator has the advantages of less land occupation, smaller investment, simplified system and lower water con-tent in the steam, deserving to be widely promoted in applications of saturated steam recovery and utilization in steel plants.

  7. Radiological Characteristics of decommissioning waste from a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahmed, Rizwan; Heo, Gyun Young [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

    2011-11-15

    The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be 1.04 x 10{sup 16} Bq, 2.09 x 10{sup 3} W, 5.31 x 10{sup 14} m{sup 3}-water, 4.69 x 10{sup 5} kg, and 7.38 x 10{sup 1} m{sup 3}, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

  8. FFT Analysis on Coupling Effect of Axial and Torsional Vibrations in Circular Cross Section Beam of Steam Turbine Generators

    Directory of Open Access Journals (Sweden)

    Xiang Xu

    2013-09-01

    Full Text Available This paper presents a novel method to nonlinearly investigate the dynamics of the coupled axial and torsional vibrations in the circular cross section beam of the steam turbine generator using the FFT analysis. Firstly, the coupled axial and torsional vibrations of a beam are proved by equivalent law of shearing stress and different boundary conditions. Then, a nonlinear mathematical model of the coupled axial and torsional vibrations is established by the Galerkin method. Lastly, the fast Fourier transform (FFT is employed to investigate the coupled effect of the beam vibration. A practical calculation example is calculated numerically and the coupled mechanism of the beam’s axial and torsional vibrations is analyzed in detail. The analysis results show that the frequencies of the coupled response would be existed in some special orders and the coupled response frequencies are smaller than the single vibration. Since for the first time the coupled mechanism of the beam’s axial and torsional vibrations is theoretically analyzed, the findings in this work may provide directive reference for practical engineering problems in design of steam turbine generators.

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

    Energy Technology Data Exchange (ETDEWEB)

    Krause, T. W.; Babbar, V. K.; Underhill, P. R. [Department of Physics, Royal Military College of Canada, Kingston, ON (Canada)

    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.

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

    Science.gov (United States)

    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.

  11. The influence of manufacturing processes on the microstructure, grain boundary characteristics and SCC behavior of Alloy 690 steam generator tubing

    Energy Technology Data Exchange (ETDEWEB)

    Sarver, J.M. [Babcock and Wilcox, Alliance, OH (United States). Research and Development Division; Doherty, P.E.; Doyle, D.M. [Babcock and Wilcox International Division, Cambridge, Ontario (Canada); Palumbo, G. [Ontario Hydro Technologies, Toronto, Ontario (Canada)

    1995-12-31

    Thermally treated Alloy 690 is the tubing material of choice for replacement steam generators in the United States. Throughout the world, it is manufactured using different melting and thermomechanical processing methods. The influence of different processing steps on the intergranular stress corrosion cracking (IGSCC) behavior of Alloy 690 has not been thoroughly evaluated. Evaluations were performed on Alloy 690 steam generator tubing produced using several different melting practices and thermomechanical processing procedures. The evaluations included extensive microstructural examinations as well as constant extension rate (CERT) tests. The CERT test results indicated that the thermally treated Alloy 690 tubing which was subjected to higher annealing temperatures displayed the highest degree of resistance to stress corrosion cracking (SCC). Examination of the microstructures indicated that the microstructural changes which are produced by increased annealing temperatures are subtle. In an attempt to further elucidate and quantify the effect of manufacturing processes on corrosion behavior, grain boundary character distribution (GBCD) measurements were performed on the same materials which were CERT tested. Analysis of GBCDs of the samples used in this study indicate that Alloy 690 exhibits a significantly larger fraction of special boundaries as compared to Alloy 600 and Alloy 800, regardless of the processing history of the tubing. Preliminary results indicate that a correlation may exist between processing method, GBCD`s and degree of IGSCC exhibited by the thermally treated samples examined in this study.

  12. Development of a Multi-dimensional Analysis Methodology for Sodium-water Reaction Phenomena in a SFR Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Eoh, Jae Hyuk; Kim, Se Yun; Choi, Byung Seon; Kim, Seong O

    2006-03-15

    A sodium-water reaction (SWR) has been considered as one of the most important issues to be resolved for designing the steam generator and the related systems of a sodium-cooled fast reactor (SFR). In this study, the reasonable SWR analysis at the vicinity of the reaction zone was made by using the appropriate multi-dimensional thermo-hydraulic simulation with detailed chemical reaction model. Based on the investigation results for the capabilities of commercial CFD codes, CFX version 5.7 and its EDM chemical model were selected to simulate the chemically reacting flow with various phases and components. As results of this study, a reasonable methodology for the SWR analysis including detailed information for each phase and the components was provided with the two-dimensional fields of velocity, temperature, mass fraction, etc. In order to investigate the subsequent tube rupture behavior at the reaction zone, the effective method to evaluate the subsequent tube rupture behavior is also proposed by using the relationship between the wastage rate and the eroded tube thickness and is set up by considering allowable stress intensity for the conventional tube material. It is expected that the results of this study will contribute to improve the reliability of a steam generator for GEN-IV SFR in the future.

  13. Doppler method leak detection for LMFBR steam generators. Pt. 1. Experimental results of bubble detection using small models

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Hiromichi [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab

    1999-05-01

    To prevent the expansion of the tube damage and to maintain structural integrity in the steam generators (SGs) of fast breeder reactors (FBRs), it is necessary to detect precisely and immediately the leakage of water from heat transfer tubes. Therefore, an active acoustic method was developed. Previous studies have revealed that in practical steam generators the active acoustic method can detect bubbles of 10 l/s within 10 seconds. To prevent the expansion of damage to neighboring tubes, it is necessary to detect smaller leakages of water from the heat transfer tubes. The Doppler method is designed to detect small leakages and to find the source of the leak before damage spreads to neighboring tubes. To evaluate the relationship between the detection sensitivity of the Doppler method and the bubble volume and bubble size, the structural shapes and bubble flow conditions were investigated experimentally, using a small structural model. The results show that the Doppler method can detect the bubbles under bubble flow conditions, and it is sensitive enough to detect small leakages within a short time. The doppler method thus has strong potential for the detection of water leakage in SGs. (author)

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

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

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

  15. Steam generation unit in a simple version of biomass based small cogeneration unit

    OpenAIRE

    Sornek Krzysztof; Filipowicz Mariusz; Szubel Mateusz; Bożek Estera; Izdebski Krzysztof

    2014-01-01

    The organic Rankine cycle (ORC) is a very promising process for the conversion of low or medium temperature heat to electricity in small and micro scale biomass powered systems. Classic ORC is analogous to Clausius–Rankine cycle in a steam power plant, but instead of water it uses low boiling, organic working fluids. Seeking energy and economical optimization of biomass-based ORC systems, we have proposed some modifications e.g. in low boiling fluid circuit construction. Due to the fact that ...

  16. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)

    1995-12-31

    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

  17. The Recovery and Power Generation Technology for Low Pressure Exhaust Steam%低压放散蒸汽回收发电技术

    Institute of Scientific and Technical Information of China (English)

    戴海波

    2015-01-01

    对钢铁企业低压饱和蒸汽管网运行情况进行研究,分析总结低压低温饱和的间断性放散蒸汽放散规律,通过放散蒸汽的有效回收,并利用螺杆膨胀机发电技术,形成低压饱和蒸汽资源化和再利用的项目方案。%The operation state of low pressure saturated steam pipeline networks of steel enterprises was investigated and the pattern of intermittent exhausting of low-pressure low-temperature saturated steam was analyzed. Through effective recovery of the exhaust steam and adopting of the power generation technology of screw expander, a project plan for recy-cling and utilization of low pressure saturated steam was drawn up.

  18. Hydrogen Generation from Catalytic Steam Reforming of Acetic Acid by Ni/Attapulgite Catalysts

    Directory of Open Access Journals (Sweden)

    Yishuang Wang

    2016-11-01

    Full Text Available In this research, catalytic steam reforming of acetic acid derived from the aqueous portion of bio-oil for hydrogen production was investigated using different Ni/ATC (Attapulgite Clay catalysts prepared by precipitation, impregnation and mechanical blending methods. The fresh and reduced catalysts were characterized by XRD, N2 adsorption–desorption, TEM and temperature program reduction (H2-TPR. The comprehensive results demonstrated that the interaction between active metallic Ni and ATC carrier was significantly improved in Ni/ATC catalyst prepared by precipitation method, from which the mean of Ni particle size was the smallest (~13 nm, resulting in the highest metal dispersion (7.5%. The catalytic performance of the catalysts was evaluated by the process of steam reforming of acetic acid in a fixed-bed reactor under atmospheric pressure at two different temperatures: 550 °C and 650 °C. The test results showed the Ni/ATC prepared by way of precipitation method (PM-Ni/ATC achieved the highest H2 yield of ~82% and a little lower acetic acid conversion efficiency of ~85% than that of Ni/ATC prepared by way of impregnation method (IM-Ni/ATC (~95%. In addition, the deactivation catalysts after reaction for 4 h were analyzed by XRD, TGA-DTG and TEM, which demonstrated the catalyst deactivation was not caused by the amount of carbon deposition, but owed to the significant agglomeration and sintering of Ni particles in the carrier.

  19. Simulation Model for Stochastic Analysis and Performance Evaluation of Steam Generator System of a Thermal Power Plant

    Directory of Open Access Journals (Sweden)

    Yogesh Vora,

    2011-06-01

    Full Text Available This paper presents the stochastic analysis and performance evaluation of turbo generator system of a thermal plant by making the use of performance evaluation using probabilistic approach. The steam generator system of thermal power plant under the research study consists mainly sub-systems boiler, super heater and reheaterarranged in series with two feasible states: working and failed. Failure and repair rates for all the sub-systems are assumed to be constant. Initially transition diagram representing the operational behavior is drawn and then problem formulation is done using Markov approach. Based on the data collection and its analysis for thermal Power Plant, Performance matrix for each subsystem is also developed. Then from these results, availability matrices and graphs of failure and repair rates for maximum availability of each system is analysed and then condition based maintenance decisions are decided.

  20. Co-generation on steam industrial systems with disks turbines; Co-geracao em sistemas industriais de vapor com turbinas de discos

    Energy Technology Data Exchange (ETDEWEB)

    Lezsovits, Ferenc [Universidad de Tecnologia y Economia de Budapest (Hungary)

    2010-03-15

    The disk turbine, also called Tesla turbine, being of simple construction and low cost, can be used as steam pressure reduction on industrial systems, generating simultaneously electric power, becoming the co-generation even at lower levels. Can be used for various operational parameters and mass flux ratios.This paper analyses the advantages and disadvantages of the turbines under various operation conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Yoshihisa Shindo; Kazuo Haga [Japan Nuclear Energy Safety Organization (JNES) Kamiya-cho MT Bldg., 4-3-20 Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

    2005-07-01

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

  2. Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-08-15

    The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

  3. J-resistance curves for Inconel 690 and Incoloy 800 nuclear steam generators tubes at room temperature and at 300 °C

    Science.gov (United States)

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

    2017-04-01

    The structural integrity of steam generator tubes is a relevant issue concerning nuclear plant safety. In the present work, J-resistance curves of Inconel 690 and Incoloy 800 nuclear steam generator tubes with circumferential and longitudinal through wall cracks were obtained at room temperature and 300 °C using recently developed non-standard specimens' geometries. It was found that Incoloy 800 tubes exhibited higher J-resistance curves than Inconel 690 for both crack orientations. For both materials, circumferential cracks resulted into higher fracture resistance than longitudinal cracks, indicating a certain degree of texture anisotropy introduced by the tube fabrication process. From a practical point of view, temperature effects have found to be negligible in all cases. The results obtained in the present work provide a general framework for further application to structural integrity assessments of cracked tubes in a variety of nuclear steam generator designs.

  4. Deposits on the secondary side of the steam generators: Causes and consequences; Depositos en el lado secundario de los generadores de vapor: causas y consecuencias

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Briceno, D.; Yague, C.; Fernandez Diaz, M.; Gomez Mancebo, B.; Fernandez Saavedra, R.

    2013-07-01

    Despite improvements in the control of the chemistry of the secondary circuit of the PWR type reactor, large amounts of corrosion products are incorporated in the steam generators and deposited on the surface of the tubes, on the flow holes the support plates and the tubes sheet. Magnetite accumulated on the tube sheet undergoes consolidation giving rise to what is known as hard sludge, which reduces heat transfer and induces a temperature rise in the tubes which will favor the possible processes degradation. Hard sludge accumulation has been identified as a necessary condition for degradation by denting of the tubes of steam generators found in Almaraz and Asco NPP: In this article, we discuss the causes of the accumulation of deposits and their effects on the degradation of the tubes of the steam generators. (Author)

  5. Risk Analysis of steam generation at the 5th refinery of South Pars Gas Company using HAZOP procedures

    Directory of Open Access Journals (Sweden)

    Vahid Aziznezhad

    2016-06-01

    Full Text Available considering HSE management program in its strategic framework, and employing integrated methods and tools commonly employed in a chemical process plant, a hazard risk assessments at the steam generation unit was investigated in an Iranian South Pars gas refinery, using PHA-PRO6 software. Recommendations were made to improve safety and prevent possible hazards, taking into account several years of production experiences, past maintenances and reported accidents and near misses. Implicating safety improvements by adopting recommendation yielded from HAZOP studies is expected to reduce system risks and enhance its reliability. The main process hazards identified include water flow fluctuation, pump failures and excess pressure leading to pipe failures, condensation at compensating steam outlets, leakage, substandard maintenance practices, deviations in control of pH, oxygen content, corrosion, condensed water entrapment, were identified as some of the major sources of risks involved. These were considered in constitution of the risk matrix, from which main recommendations and execution time needed are produced. Using the available techniques and standards as well as installing flow control devices could ensure maintaining a normal operating pressure which will in turn, reduce pump stoppage and the unit being out of service due to water shortage. Starting-up procedure for the unit was also revised based on the results of this study. The need for regular (annual inspection of instrumentation controls especially solenoids and check valves were also emphasized. To avoid reverse flow of water into the boilers, installation of a check valve before HV 213 was recommended.

  6. Automatic control of plants of direct steam generation with cylinder-parabolic solar collectors; Control automatico de plantas de generacion directa de vapor con colectores solares cilindro-parabolicos

    Energy Technology Data Exchange (ETDEWEB)

    Valenzuela Gutierrez, L.

    2008-07-01

    The main objective of this dissertation has been the contributions to the operation in automatic mode of a new generation of direct steam generation solar plants with parabolic-trough collectors. The dissertation starts introducing the parabolic-trough collectors solar thermal technology for the generation of process steam or steam for a Rankine cycle in the case of power generation generation, which is currently the most developed and commercialized technology. Presently, the parabolic-trough collectors technology is based on the configuration known as heat-exchanger system, based in the use of a heat transfer fluid in the solar field which is heated during the recirculation through the absorber tubes of the solar collectors, transferring later on the that thermal energy to a heat-exchanger for steam generation. Direct steam generation in the absorber tubes has always been shown as an ideal pathway to reduce generation cost by 15% and increase conversion efficiency by 20% (DISS, 1999). (Author)

  7. Neutronics-thermalhydraulics coupling in a CANDU SCWR

    Science.gov (United States)

    Adouki, Pierre

    In order to implement new nuclear technologies as a solution to the growing demand for energy, 10 countries agreed on a framework for international cooperation in 2002, to form the Generation IV International Forum (GIF). The goal of the GIF is to design the next generation of nuclear reactors that would be cost effective and would enhance safety. This forum has proposed several types of Generation IV reactors including the Supercritical Water-Cooled Reactor (SCWR). The SCWR comes in two main configurations: pressure vessel SCWR and pressure tube SCWR (PT-SCWR). In this study, the CANDU SCWR (a PT-SCWR) is considered. This reactor is oriented vertically and contains 336 channels with a length of 5 m. The target coolant inlet and outlet temperatures are 350 Celsius and 625 Celsius, respectively. The coolant flows downwards, and the reactor power is 2540 MWth. Various fuel designs have been considered in order not to exceed the linear element rating. However, the dependency between the core power and thermalhydraulics parameters results in the necessity to use a neutronics/thermalhydaulics coupling scheme to determine the core power and the thermalhydraulics parameters. The core power obtained has a power peaking factor of 1.4. The bundle power distribution for all channels has a peak at the third bundle from the inlet, but the value of this peak increases with the channel power. The heat-transfer coefficient and the specific-heat capacity have a peak at the same location in a channel, and this location shifts toward the inlet as the channel power increases. The exit coolant temperature increases with the channel power, while the exit coolant density and pressure decrease with the channel power. Also, higher channel powers lead to higher fuel and cladding temperatures. Moreover, as the coupling method is applied, the effective multiplication factor and the values of thermalhydaulics parameters oscillate as they converge.

  8. A study for good regulatin of the CANDU's in Korea. Development of safety regulatory requirement for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Ki; Shin, Y. K.; Kim, J. S.; Yu, Y. J.; Lee, Y. J. [Ajou Univ., Suwon (Korea, Republic of)

    2001-03-15

    The objective of project is to derive the policy recommendations to improve the efficiency of CANDU plants regulation. These policy recommendations will eventually contribute to the upgrading of Korean nuclear regulatory system and safety enhancement. During the first phase of this 2 years study, following research activities were done. On-site survey and analysis on CANDU plants regulation. Review on CANDU plants regulating experiences and current constraints. Review and analysis on the new Canadian regulatory approach.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-06-03

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

  10. Development of a 1D thermal-hydraulic analysis code for once-through steam generator in SMRs using straight tubes

    Energy Technology Data Exchange (ETDEWEB)

    Park, Youngjae; Kim, Iljin; Kim, Hyungdae [Kyung Hee University, Yongin (Korea, Republic of)

    2015-10-15

    Diverse integral/small-modular reactors (SMRs) have been developed. Once-through steam generator (OTSG) which generates superheated steam without steam separator and dryer was used in the SMRs to reduce volume of steam generator. It would be possible to design a new steam generator with best estimate thermal-hydraulic codes such as RELAP and MARS. However, it is not convenience to use the general purpose thermal-hydraulic analysis code to design a specific component of nuclear power plants. A widely used simulation tool for thermal-hydraulic analysis of drum-type steam generators is ATHOS, which allows 3D analysis. On the other hand, a simple 1D thermal-hydraulic analysis code might be accurate enough for the conceptual design of OTSG. In this study, thermal-hydraulic analysis code for conceptual design of OTSG was developed using 1D homogeneous equilibrium model (HEM). A benchmark calculation was also conducted to verify and validate the prediction accuracy of the developed code by comparing with the analysis results with MARS. Finally, conceptual design of OTSG was conducted by the developed code. A simple 1D thermal-hydraulic analysis code was developed for the purpose of conceptual design OTSG for SMRs. A set of benchmark calculations was conducted to verify and validate the analysis accuracy of the developed code by comparing results obtained with a best-estimated thermal-hydraulic analysis code, MARS. Finally, analysis of two different OTSG design concepts with superheating and recirculation was demonstrated using the developed code.

  11. Fuel management simulation for CANFLEX-RU in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fuel management simulations have been performed for CANFLEX-09% RU fuel in the CANDU 6 reactor. In this study, the bi-directional 4-bundle shift fuelling scheme was assumed. The lattice cell and time-average calculation were carried out. The refuelling simulation calculations were performed for 600 full power days. Time-averaged results show good axial power profile with the CANFLEX-RU fuel. During the simulation period, the maximum channel and bundle power were maintained below the licensing limit of CANDU 6 reactor. 7 refs., 4 figs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  13. Analysis of two-phase flow instability in helical tube steam generator in high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Yu; Lv, Xuefeng; Wang, Shengfei; Niu, Fenglei; Tian, Li [North China Electric Power Univ., Beijing (Switzerland)

    2012-03-15

    The steam generator composed of multi-helical tubes is used in high temperature gas cooled reactors and two-phase flow instability should be avoided in design. And density-wave oscillation which is mainly due to flow, density and the relationship between the pressure drop delays and feedback effects is one of the two-phase flow instability phenomena easily to occur. Here drift-flux model is used to simulate the performance of the fluid in the secondary side and frequency domain and time domain methods are used to evaluate whether the density-wave oscillation will happen or not. Several operating conditions with nominal power from 15% to 30% are calculated in this paper. The results of the two methods are in accordance, flow instability will occur when power is less than 20% nominal power, which is also according with the result of the experiments well.

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

    Science.gov (United States)

    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.

  15. Localization of defects in steam generator tubes using a multi-coil eddy current probe dedicated to high speed inspection

    Energy Technology Data Exchange (ETDEWEB)

    Joubert, P.-Y.; Le Bihan, Y.; Placko, D. [Ecole Normale Superieure de Cachan (France). Laboratoire d' Electricite Signaux et Robotique

    2002-07-01

    Steam generator (SG) tubing of pressurized water reactor in nuclear plants must be rapidly and accurately checked in order to detect defects in their early stages. In this paper, the authors present a multi-coil eddy current (EC) probe allowing both high speed inspection and circumferential localization of defects in the tube wall. A method of multi-coil EC signal processing, based on a continuous wavelet transform combined with a maximum likelihood diagnosis, is elaborated in order to enhance the detection performances and to provide automatic localization of defects. The inspection of SG tube samples shows good localization performances for defects as small as 10% deep, 15 mm long and 100 {mu}m wide outer diameter notches, of both circumferential and axial orientations. (author)

  16. Failure problems in superheater spacers of steam generators; Problematica de fallas en espaciadores de sobrecalentadores de generadores de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Chacon Nava, Jose G.; Martinez Villafane, Alberto [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Fuentes Samaniego, Raul [Universidad Autonoma de Nuevo Leon (Mexico); Mojica Calderon, Cecilio [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1988-12-31

    In this article the general aspects of the steam generator superheater fixed spacers failures are analyzed, emphasis is made on the influence several aspects such as the operation of the unit have, the appropriate execution of welds and the selection of binding materials. Likewise several recommendations are made to bring the failures to a minimum. [Espanol] En este articulo se analizan aspectos generales de fallas en espaciadores fijos de sobrecalentadores de generadores de vapor, y se hace hincapie en la influencia que tienen diversos aspectos tales como la operacion de la unidad, la adecuada ejecucion de soldaduras y la seleccion del material de aporte. Asimismo, se proponen algunas recomendaciones para reducir al minimo las fallas.

  17. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. A review of CANDU feeder wall thinning

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Han Sub [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Flow Accelerated Corrosion is an active degradation mechanism of CANDU feeder. The tight bend downstream to Grayloc weld connection, close to reactor face, suffers significant wall thinning by FAC. Extensive in-service inspection of feeder wall thinning is very difficult because of the intense radiation field, complex geometry, and space restrictions. Development of a knowledge-based inspection program is important in order to guarantee that adequate wall thickness is maintained throughout the whole life of feeder. Research results and plant experiences are reviewed, and the plant inspection databases from Wolsong Units One to Four are analyzed in order to support developing such a knowledge-based inspection program. The initial thickness before wall thinning is highly non-uniform because of bending during manufacturing stage, and the thinning rate is non-uniform because of the mass transfer coefficient distributed non-uniformly depending on local hydraulics. It is obvious that the knowledge-based feeder inspection program should focus on both fastest thinning locations and thinnest locations. The feeder wall thinning rate is found to be correlated proportionately with QV of each channel. A statistical model is proposed to assess the remaining life of each feeder using the QV correlation and the measured thicknesses. W-1 feeder suffered significant thinning so that the shortest remaining life barely exceeded one year at the end of operation before replacement. W-2 feeder showed far slower thinning than W-1 feeder despite the faster coolant flow. It is believed that slower thinning in W-2 is because of higher chromium content in the carbon steel feeder material. The average Cr content of W-2 feeder is 0.051%, while that value is 0.02% for W-1 feeder. It is to be noted that FAC is reduced substantially even though the Cr content of W-2 feeder is still very low

  19. Direct generation of steam and electricity in a open cycle Rankine; Generacion directa de vapor y electricidad en un ciclo Rankine abierto

    Energy Technology Data Exchange (ETDEWEB)

    Lentz, Alvaro; Almanza, Rafael; Flores, Vicente [UNAM, Mexico, D.F. (Mexico)

    2000-07-01

    In this work the results of the experimental tests about steam and electricity generation are presented. This work carried out in the solar thermal power plant of the Institute of Engineering with direct steam generation in parabolic through. The global efficiency of the system is studied as for the conversion solar-electricity. The efficiency is determined and it describes the obtaining process of the main plant components, like they are, the solar steam generator, the steam motor and the electric generator. [Spanish] En este trabajo se presentan los resultados de las pruebas experimentales de la generacion de vapor y electricidad realizadas en la planta solar del Instituto de Ingenieria con generacion directa de vapor en concentradores de canal parabolico. Se estudia la eficiencia global del sistema en cuanto a la conversion de energia solar-electricidad. Se determina la eficiencia y describe el proceso de obtencion de la misma y de los principales componentes de la planta como son, el generador de vapor solar, el motor de pistones de vapor y el alternador electrico.

  20. GRUVAL for ET inspection of the steam generator tubes; GRUVAL para la inspeccion ET de los tubos de los Generadores de Vapor

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  1. Devices for the contamination containment employees in the steam generator inspection; Dispositivo para confinamiento de la contaminacion empleados en la inspeccion de generadores de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Bueno, A.; Largo Izquierdo, P.; Calleja Rubio, J. A.

    2010-07-01

    The process of induced current inspection of the tubes of the steam generator is a typical programmed inspections at each refueling outages of pressurized water in nuclear power plants. components inspection being quite active, interested in the program of continuous improvement, further optimize the inspection system.

  2. Development of CANDU ECCS performance evaluation methodology and guides

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kwang Hyun; Park, Kyung Soo; Chu, Won Ho [Korea Maritime Univ., Jinhae (Korea, Republic of)

    2003-03-15

    The objectives of the present work are to carry out technical evaluation and review of CANDU safety analysis methods in order to assist development of performance evaluation methods and review guides for CANDU ECCS. The applicability of PWR ECCS analysis models are examined and it suggests that unique data or models for CANDU are required for the following phenomena: break characteristics and flow, frictional pressure drop, post-CHF heat transfer correlations, core flow distribution during blowdown, containment pressure, and reflux rate. For safety analysis of CANDU, conservative analysis or best estimate analysis can be used. The main advantage of BE analysis is a more realistic prediction of margins to acceptance criteria. The expectation is that margins demonstrated with BE methods would be larger that when a conservative approach is applied. Some outstanding safety analysis issues can be resolved by demonstration that accident consequences are more benign than previously predicted. Success criteria for analysis and review of Large LOCA can be developed by top-down approach. The highest-level success criteria can be extracted from C-6 and from them, the lower level criteria can be developed step-by-step, in a logical fashion. The overall objectives for analysis and review are to verify radiological consequences and frequency are met.

  3. Overview of activities on CANDU fuel in Argentina

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez, L.; Valesi, J., E-mail: lalvarez@cnea.gov.ar [National Commission on Atomic Energy, Fuel Engineering Department (Argentina)

    2013-07-01

    This paper gives an outline of activities on CANDU fuel in Argentina. It discusses the nuclear activities and electricity production in Argentina, evolution of the activities in fuel engineering, fuel fabrication, fuel performance at Embalse nuclear power plant and spent fuel storage options.

  4. Effect of DUPIC cycle on CANDU reactor safety parameters

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M. A. [Atomic Energy Authority, ETRR-2, Cairo (Egypt); Badawi, Alya [Dept. of Nuclear and Radiation Engineering, Alexandria University, Alexandria (Egypt)

    2016-10-15

    Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO{sub 2} enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

  5. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  6. Estimation of CANDU spent fuel disposal canister lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Kook, Dong Hak; Lee, Min Soo; Hwang, Yong Soo; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    Active nuclear energy utilization causes significant spent fuel accumulation problem. The cumulative amount of spent fuel is about 10,083 ton as of Dec. 2008, and is expected to increase up to 19,000 ton by 2020. Of those, CANDU spent fuels account for more than 60% of the total amounts. CANDU spent fuels had been stored in dry concrete silos since 1991 and during the past 15 years, 300 silos were constructed and {approx}3,200 ton of spent fuels are stored now. Another dry storage facility MACSTOR /KN-400 will store new-coming CANDU spent fuels from 2009. But, after intermediate storage ends, all CANDU spent fuels have to be disposed within multi-layer metallic canister which is composed of cast iron inside and copper outside. Canister lifetime estimation, therefore, is very important for the final disposal safety analysis. The most significant factor of lifetime is copper corrosion, and Y. S. Hwang developed a corrosion model in order to predict the general corrosion effect on copper canister lifetime during the final disposal period. This research applied his model to KURT1 where many disposal researches are being performed actively and the results shows safe margin of the copper canister for the very long-term disposal.

  7. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  8. Production of HBR from bromine and steam for off-peak electrolytic hydrogen generation

    Energy Technology Data Exchange (ETDEWEB)

    Schlief, R.E.; Hanrahan, R.J.; Stoy, M.A. [Univ. of Florida, Gainesville, FL (United States)] [and others

    1995-09-01

    Progress is reported on the development of a renewable energy source based solar-electrolytic system for production of hydrogen and oxygen. It employs water, bromine, solar energy and supplemental electrical power. The concept is being developed by Solar Reactor Technologies, Inc., (SRT), with the U.S. Department of Energy (DOE). An overview of the nature and objectives of this program is provided here, and technical progress made during the first (three-month) performance period of the Phase I work effort is reported. The SRT concept entails (1) absorption of concentrated solar radiation by bromine vapor Br{sub 2(g)} in a high-temperature reactor producing Br{sub (g)} atoms, (2) reaction of Br{sub (g)} with water yielding hydrogen bromide (HBr), and (3) electrolysis of stored hydrogen bromide for production of H{sub 2(g)} and recovery of Br{sub 2(I)}. Incorporation of solar radiation in the primary photochemical step (1) reduces by 50 - 70% the electrical power required to split water. The SRT concept is very attractive from an economic viewpoint as well. The reversible fuel cell, employed in the SRT electrolysis concept is capitalized via its use in load leveling by the utility. A 1 kW solar reactor was designed and constructed during the first three-month performance period by SRT personnel at the University of Florida, Gainesville. It was employed in taking survey data of the reaction between bromine and steam at temperatures between 900 and 1300 K. This reaction was run under purely thermal conditions, i.e. in the absence of solar photons. The experimental data are reported and interpreted employing concomitant thermodynamic calculations. The anticipated improvement is discussed briefly as well as the effect of a photochemical boost to the reaction. The amount of this enhancement will be studied in the next three month performance period.

  9. Current safety issues of CANDU licensing

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. [University of Ottawa, Ottawa (Canada); Natalizio, A. [ENSAC Associates, Ontario (Canada)

    1994-01-15

    As requested by Korea Institute of Nuclear Safety(KINS), the status of five generic licensing issues has been examined and their potential impact on a new plant that would be constructed in Canada has been evaluated. The results and conclusions of this evaluation are summarized as follows: steam explosion in calandria, hydrogen explosion in containment, use of PSA in reactor licensing, human factors, safety critical software.

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

    Energy Technology Data Exchange (ETDEWEB)

    Belle R. Upadhyaya; J. Wesley Hines

    2004-09-27

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

  11. Procedure of calculation of the spatial distribution of temperatures and heat fluxes in the steam generator of a nuclear power installation with an RBEC fast-neutron reactor

    Science.gov (United States)

    Frolov, A. A.; Sedov, A. A.

    2016-08-01

    A method for combined 3D/1D-modeling of thermohydraulics of a once-through steam generator (SG) based on the joint analysis of three-dimensional thermo- and hydrodynamics of a single-phase heating coolant in the intertube space and one-dimensional thermohydraulics of steam-generating channels (tubes) with the use of well-known friction and heat-transfer correlations under various boiling conditions is discussed. This method allows one to determine the spatial distribution of temperatures and heat fluxes of heat-exchange surfaces of SGs with a single-phase heating coolant in the intertube space and with steam generation within tubes. The method was applied in the analytical investigation of typical operation of a once-through SG of a nuclear power installation with an RBEC fast-neutron heavy-metal reactor that is being designed by Kurchatov Institute in collaboration with OKB GIDROPRESS and Leipunsky Institute of Physics and Power Engineering. Flow pattern and temperature fields were obtained for the heavy-metal heating coolant in the intertube space. Nonuniformities of heating of the steam-water coolant in different heat-exchange tubes and nonuniformities in the distribution of heat fluxes at SG heat-exchange surfaces were revealed.

  12. Use of Coke Oven Gas during shutdown period of Direct Reduced Iron Plant for Steam Generation: Experimentation and Optimization

    Directory of Open Access Journals (Sweden)

    Harish C. Dalai

    2015-11-01

    Full Text Available Development of any country is largely based on its magnitude of industrial growth. Steel industries in India took a leading role in the world after mid-sixties. During the last five decades, the steel industries all over the world made considerable developments in new methods for reducing iron ore directly to metallic iron for use as commercial scrap substitute in the manufacture of steels. BSL has been producing steel and other products over last two decades and the above process is used in BSL for one decade. Direct reduction is a process, which extracts high metallic solid iron by removing oxygen from iron ore or any other iron oxides without passing through molten stage, i.e. solid state reduction. The product, so formed is known as directly reduced tron (DRI or Sponge Iron. Due to shortage of raw material or any other problem in DRI sections, the DRI section undergoes a temporary short down. During the shutdown period the equipments tend to be idle. To keep the equipment ready for operation as well as to maintain the plants economy an alternate methodology is used in this research. Various operations are: CO gas from Coke Oven, DRI, Steam Generation, and Power Generation.

  13. Design and Development of a Robotic Crawler for CANDU Fuel Channel Inspection

    Science.gov (United States)

    Shukla, Shivam

    For the design of a new robotic crawler drive unit for CANDU fuel channel inspection, a complete design and screening process was done in order to fulfil the objective of this research. A brief explanation of CANDU reactors is provided along with a discussion of the inspection systems that are currently in use. A study of some existing inspection systems is presented which was used for the development of the new robotic crawler design. A number of concepts were generated which underwent a screening process with the help of two design tools. With the help of these tools, a concept was chosen as the final design and details of it are presented. To demonstrate a proof-of-concept, the physical prototype of the robotic crawler was manufactured and assembled. A speed controller was implemented in the final design of the robotic crawler. A set of test procedures were performed on the final design and the results are discussed. Some improvements that can be done on the final design of the robotic crawler are also discussed in the final section of this thesis.

  14. Power Efficiency of Steam Turbine Generator Switching into Thermal Circuit of Small and Medium Boiler Houses

    Directory of Open Access Journals (Sweden)

    R. I. Yesman

    2007-01-01

    Full Text Available The paper is devoted to the solution of the problem concerning power saving on the basis of small power-and-heat-supply plants.Power efficiency of power turbine generator switching into thermal circuit of small and medium boiler houses is justified in the paper.

  15. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Young; Park, Kun Chul [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2003-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SDS-1, SDS2, ECCS, and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  16. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Jae Young; Bang, Kwang Hyun [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2001-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SOS-1, SOS-2, ECCS and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  17. Remedial actions to improve the availability in steam generators; Acciones correctivas para mejorar la disponibilidad en generadores de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Torres Toledano, J. Gerardo; Porras Loaiza, G. Lizbeth; Sanchez Hernandez, Laura E.; Salinas B, Victor M.; Nebart G, Jesus [Instituto de Investigaciones Electricas, Temixco, Morelos (Mexico)

    2001-07-01

    In this article the results of the analysis of recurrent faults of four Generating Units of Thermoelectric Power Plants (UGT) of the Commission Federal de Electricidad (CFE) are presented. In three of the four cases, the studies focused to identify and to correct the root causes of recurrent faults in steam generator tubes, the analyzed units are: U-2 of the Thermoelectric Complex Pdte. Adolfo Lopez Mateos (CTPALM) of Tuxpan, Veracruz; U-3 of the Salamanca Thermoelectric Power Plant (TPP), and U-1 and 2 of the Thermoelectric Power Plant (TPP) Felipe Carrillo Puerto of Valladolid, Yucatan. In the other case the combinations air insufficiency was analyzed as one of the causes of decrement of power of the Unit 1 of the TPP Gral. Manuel Alvarez Moreno of Manzanillo, Colima. The steam generators of Unit 2 of the CTPALM and Unit 3 of the TPP of Salamanca presented faults in the reheater (RH) bank. In this case the recommendations focused to the improvement in the operative conditions to attenuate the faults by corrosion at high temperature. Unlike the CTPALM, in the case of U-3 of Salamanca it was concluded that the redesign of the RH was necessary; for this purpose it was proposed to permute trajectories of hot tubes with the cold ones to make the steam temperature uniform at the RH outlet. In order to evaluate the problems that have appeared in the steam generator of Units 1 and 2 of the TPP Felipe Carrillo Puerto of Valladolid, Yucatan the faults of the tubes of water wall were analyzed and it was found that the main mechanism of overheat is due to the impact of the combustion flame. In the study of the Unit 1 of the TPP Gral. Manuel Alvarez Moreno of Manzanillo, Colima, a methodology was developed to quantify the contribution that the main equipment has in the decrement of the power of the unit; this methodology consists of applying a process of computational simulation based in the design conditions and of real operation. In agreement with the obtained results, it

  18. Differential geometry based model for eddy current inspection of U-bend sections in steam generator tubes

    Science.gov (United States)

    Mukherjee, Saptarshi; Rosell, Anders; Udpa, Lalita; Udpa, Satish; Tamburrino, Antonello

    2017-02-01

    The modeling of U-Bend segment in steam generator tubes for predicting eddy current probe signals from cracks, wear and pitting in this region poses challenges and is non-trivial. Meshing the geometry in the cartesian coordinate system might require a large number of elements to model the U-bend region. Also, since the lift-off distance between the probe and tube wall is usually very small, a very fine mesh is required near the probe region to accurately describe the eddy current field. This paper presents a U-bend model using differential geometry principles that exploit the result that Maxwell's equations are covariant with respect to changes of coordinates and independent of metrics. The equations remain unaltered in their form, regardless of the choice of the coordinates system, provided the field quantities are represented in the proper covariant and contravariant form. The complex shapes are mapped into simple straight sections, while small lift-off is mapped to larger values, thus reducing the intrinsic dimension of the mesh and stiffness matrix. In this contribution, the numerical implementation of the above approach will be discussed with regard to field and current distributions within the U-bend tube wall. For the sake of simplicity, a two dimensional test case will be considered. The approach is evaluated in terms of efficiency and accuracy by comparing the results with that obtained using a conventional FE model in cartesian coordinates.

  19. A prediction method for the general corrosion behavior of Alloy 690 steam generator tube using eddy current testing

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hee-Sang; Choi, Myung Sik; Lee, Deok Hyun; Hur, Do Haeng, E-mail: dhhur@kaeri.re.kr

    2016-02-15

    Highlights: • A corrosion test for the tubes with different levels of eddy current noise was conducted. • A relationship between the corrosion rate and the eddy current noise of tubes was explored. • Corrosion rate was closely correlated to the tube noise of a rotating pancake probe. • Corrosion rate was not related to the tube noise measured using a bobbin probe. - Abstract: The purpose of this work is to develop an eddy current testing method to predict the general corrosion behavior of Alloy 690 steam generator tubes. A corrosion test was conducted for tubes with different levels of eddy current noise in simulated primary water at 330 °C, and their corrosion behavior was correlated with the tube noise measured using bobbin and rotating probes. The corrosion behavior was closely correlated with the tube noise measured using a rotating probe. However, there was no correlation between the corrosion behavior and the tube noise measured using a bobbin probe. The tube noise value measured using a rotating pancake coil probe is suggested to be a significant parameter in estimating the general corrosion behavior of tubes.

  20. Heat transfer characteristics of porous sludge deposits and their impact on the performance of commercial steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Kreider, M.A.; White, G.A.; Varrin, R.D.; Ouzts, P.J.

    1998-12-01

    Steam generator (SG) fouling, in the form of corrosion deposits on the secondary sides of SG tubes, has been known to occur in almost all commercial US nuclear PWR (pressurized water reactor) plants. The level of fouling, as measured by the quantity of corrosion products that form, varies widely from plant to plant. In addition, the effect of SG fouling, as measured by a decrease in effective heat-transfer coefficient, has also varied substantially among commercial US plants. While some have observed large decreases in heat transfer, others have noted little change in performance despite the presence of significant quantities of secondary corrosion layers on their SG tubes. This observation has led to considerable confusion about what role secondary deposits play in causing heat-transfer degradation in SGs. As will become clear later in this report, secondary deposits can have a wide range of effects on heat transfer, from highly resistive to slightly enhancing (reflected by negative fouling). These different behaviors are the result of differences in deposit thickness, composition, and morphology. The main focus of this report is an investigation of the effects of secondary deposits on SG thermal performance. This investigation includes compilation of detailed information on the properties of tube scale at five commercial US nuclear plants and corresponding information characterizing SG thermal performance at these plants.

  1. In situ generation of steam and alkaline surfactant for enhanced oil recovery using an exothermic water reactant (EWR)

    Science.gov (United States)

    Robertson, Eric P

    2011-05-24

    A method for oil recovery whereby an exothermic water reactant (EWR) encapsulated in a water soluble coating is placed in water and pumped into one or more oil wells in contact with an oil bearing formation. After the water carries the EWR to the bottom of the injection well, the water soluble coating dissolves and the EWR reacts with the water to produce heat, an alkali solution, and hydrogen. The heat from the EWR reaction generates steam, which is forced into the oil bearing formation where it condenses and transfers heat to the oil, elevating its temperature and decreasing the viscosity of the oil. The aqueous alkali solution mixes with the oil in the oil bearing formation and forms a surfactant that reduces the interfacial tension between the oil and water. The hydrogen may be used to react with the oil at these elevated temperatures to form lighter molecules, thus upgrading to a certain extent the oil in situ. As a result, the oil can flow more efficiently and easily through the oil bearing formation towards and into one or more production wells.

  2. Development of a water leak detector system for LMFBR steam generator. Pt. 1; Sound attenuation due to bubbles

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Hiromichi; Yoshida, Kazuo (Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.)

    1994-03-01

    In the steam generators (SG) of LMFBR, it is necessary to detect the leakage of water from tubes of heat exchanger as soon as leakage is occurred. The active acoustic detection method has drawn general interests owing to its short response time and reduction of the influence of background noise. In this paper, in order to study the applicability of active acoustic method for detection of water leakage in the SG, the sound attenuation characteristics due to bubbles are investigated under various bubble conditions and emitted sound conditions. Furthermore, using SG sector model, sound attenuation characteristics due to injection of bubbles are studied. As a result, it is clarified that the sound attenuation due to bubbles varies dependent upon size of bubbles, void fraction and thickness of bubble layer, that the attenuation of sound reaches maximum when bubbles resonate with the emitted frequency. The sound attenuation due to bubbles in the SG model attenuates immediately upon injection of bubbles, and sound attenuation depends upon bubble size as well as void fraction. (author).

  3. Liquid metal reactor KALIMER development - Study on the high temperature properties of the steam generator tubing for LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yong Soo; Kim, Soon Tae; Park, Hui Sang; Kim, Soo Han [Yonsei University, Seoul (Korea); Kim, Young Sik [Andong National University, Andong (Korea)

    1999-04-01

    This work dealt with the evaluation of super stainless steels for steam generator tubing of LMFBR. The experimental alloys were designed to simulate the elimination of alloying elements, in special, C and N. Regardless of carbon contents, super stainless steels showed the excellent properties (tensile properties and corrosion resistance) than those of 9Cr-1Mo steel. Nitrogen content has affected positively the ultimate tensile strength and yield strength by TT(Thermal Treatment), but the elongation was reduced by TT in case of nitrogen free alloy and the elongation was largely increased by TT in case of nitrogen bearing alloys. In acidic chloride environment, nitrogen has influenced a little on corrosion potential and critical current density, but largely on passive current density, especially, at high potential. However, the trend of corrosion potential and critical current density by nitrogen was similar to the results in acidic solutions, but passive current density was largely affected by nitrogen content of stainless steels. 29 refs., 24 figs., 8 tabs. (Author)

  4. Remedial actions to improve the availability in steam generators; Acciones correctivas para mejorar la disponibilidad en generadores de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Torres Toledano, J. Gerardo; Porras Loaiza, G. Lizbeth; Sanchez Hernandez, Laura E.; Salinas B, Victor M.; Nebart G, Jesus [Instituto de Investigaciones Electricas, Temixco, Morelos (Mexico)

    2001-07-01

    In this article the results of the analysis of recurrent faults of four Generating Units of Thermoelectric Power Plants (UGT) of the Commission Federal de Electricidad (CFE) are presented. In three of the four cases, the studies focused to identify and to correct the root causes of recurrent faults in steam generator tubes, the analyzed units are: U-2 of the Thermoelectric Complex Pdte. Adolfo Lopez Mateos (CTPALM) of Tuxpan, Veracruz; U-3 of the Salamanca Thermoelectric Power Plant (TPP), and U-1 and 2 of the Thermoelectric Power Plant (TPP) Felipe Carrillo Puerto of Valladolid, Yucatan. In the other case the combinations air insufficiency was analyzed as one of the causes of decrement of power of the Unit 1 of the TPP Gral. Manuel Alvarez Moreno of Manzanillo, Colima. The steam generators of Unit 2 of the CTPALM and Unit 3 of the TPP of Salamanca presented faults in the reheater (RH) bank. In this case the recommendations focused to the improvement in the operative conditions to attenuate the faults by corrosion at high temperature. Unlike the CTPALM, in the case of U-3 of Salamanca it was concluded that the redesign of the RH was necessary; for this purpose it was proposed to permute trajectories of hot tubes with the cold ones to make the steam temperature uniform at the RH outlet. In order to evaluate the problems that have appeared in the steam generator of Units 1 and 2 of the TPP Felipe Carrillo Puerto of Valladolid, Yucatan the faults of the tubes of water wall were analyzed and it was found that the main mechanism of overheat is due to the impact of the combustion flame. In the study of the Unit 1 of the TPP Gral. Manuel Alvarez Moreno of Manzanillo, Colima, a methodology was developed to quantify the contribution that the main equipment has in the decrement of the power of the unit; this methodology consists of applying a process of computational simulation based in the design conditions and of real operation. In agreement with the obtained results, it

  5. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

  6. Prediction and modeling of the two-dimensional separation characteristic of a steam generator at a nuclear power station with VVER-1000 reactors

    Science.gov (United States)

    Parchevsky, V. M.; Guryanova, V. V.

    2017-01-01

    A computational and experimental procedure for construction of the two-dimensional separation curve (TDSC) for a horizontal steam generator (SG) at a nuclear power station (NPS) with VVER-reactors. In contrast to the conventional one-dimensional curve describing the wetness of saturated steam generated in SG as a function of the boiler water level at one, usually rated, load, TDSC is a function of two variables, which are the level and the load of SGB that enables TDSC to be used for wetness control in a wide load range. The procedure is based on two types of experimental data obtained during rated load operation: the nonuniformity factor of the steam load at the outlet from the submerged perforated sheet (SPS) and the dependence of the mass water level in the vicinity of the "hot" header on the water level the "cold" end of SG. The TDSC prediction procedure is presented in the form of an algorithm using SG characteristics, such as steam load and water level as the input and giving the calculated steam wetness at the output. The zoneby-zone calculation method is used. The result is presented in an analytical form (as an empirical correlation) suitable for uploading into controllers or other controls. The predicted TDSC can be used during real-time operation for implementation of different wetness control scenarios (for example, if the effectiveness is a priority, then the minimum water level, minimum wetness, and maximum turbine efficiency should be maintained; if safety is a priority, then the maximum level at the allowable wetness and the maximum water inventory should be kept), for operation of NPS in controlling the frequency and power in a power system, at the design phase (as a part of the simulation complex for verification of design solutions), during construction and erection (in developing software for personnel training simulators), during commissioning tests (to reduce the duration and labor-intensity of experimental activities), and for training.

  7. A step towards closing the CANDU fuel cycle: an innovative scheme for reprocessing used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Collins, F.; Lister, D. [Univ. of New Brunswick, UNB Nuclear, Dept. of Chemical Engineering, Fredericton, New Brunswick (Canada)

    2011-07-01

    Disposal versus reprocessing costs for used CANDU fuel was recently discussed by Rozon and Lister in a report produced for the Nuclear Waste Management Organization (NWMO). Their study discussed the economic incentives for reprocessing, not for the recovery of fissile uranium but for the recovery of plutonium ash. A $370/kg break-even price of uranium was calculated, and their model was found to be very sensitive to the reprocessing costs of the chosen technology. Findings were consistent with earlier studies done by Harvard University. Various reprocessing technologies (most based on solvent extraction) have been in use for many decades, but there appears to be no conceptual engineering study available in the open literature for a spent fuel reprocessing facility - one that includes process flows, operating costs and economic analysis. A deeper engineering study of the design and economics of re-processing technologies has since been undertaken by the nuclear group at the University of New Brunswick. An improved fluorination process was developed and modeled using ASPEN process simulation software. This study examines the impact of chosen technology on the spent fuel re-processing costs. (author)

  8. Probabilistic fracture mechanics applied for DHC assessment in the cool-down transients for CANDU pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Radu, Vasile, E-mail: vasile.radu@nuclear.ro [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania); Roth, Maria [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania)

    2012-12-15

    For CANDU pressure tubes made from Zr-2.5%Nb alloy, the mechanism called delayed hydride cracking (DHC) is widely recognized as main mechanism responsible for crack initiation and propagation in the pipe wall. Generation of some blunt flaws at the inner pressure tube surface during refueling by fuel bundle bearing pad or by debris fretting, combined with hydrogen/deuterium up-take (20-40 ppm) from normal corrosion process with coolant, may lead to crack initiation and growth. The process is governed by hydrogen hysteresis of terminal solid solubility limits in Zirconium and the diffusion of hydrogen atoms in the stress gradient near to a stress spot (flaw). Creep and irradiation growth under normal operating conditions promote the specific mechanisms for Zirconium alloys, which result in circumferential expansion, accompanied by wall thinning and length increasing. These complicate damage mechanisms in the case of CANDU pressure tubes that are also are affected by irradiation environment in the reactor core. The structural integrity assessment of CANDU fuel channels is based on the technical requirements and methodology stated in the Canadian Standard N285.8. Usually it works with fracture mechanics principles in a deterministic manner. However, there are inherent uncertainties from the in-service inspection, which are associated with those from material properties determination; therefore a necessary conservatism in deterministic evaluation should be used. Probabilistic approach, based on fracture mechanics principle and appropriate limit state functions defined as fracture criteria, appears as a promising complementary way to evaluate structural integrity of CANDU pressure tubes. To perform this, one has to account for the uncertainties that are associated with the main parameters for pressure tube assessment, such as: flaws distribution and sizing, initial hydrogen concentration, fracture toughness, DHC rate and dimensional changes induced by long term

  9. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Directory of Open Access Journals (Sweden)

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  10. Hydrogen generation from steam reforming of ethanol in dielectric barrier discharge

    Institute of Scientific and Technical Information of China (English)

    Baowei Wang; Yijun Lü; Xu Zhang; Shuanghui Hu

    2011-01-01

    Dielectric barrier discharge(DBD)was used for the generation of hydrogen from ethanol reforming.Effects of reaction conditions,such as vaporization temperature,ethanol flow rate,water/ethanol ratio,and addition of oxygen,on the ethanol conversion and hydrogen yield,were studied.The results showed that the increase of ethanol flow rate decreased ethanol conversion and hydrogen yield,and high water/ethanol ratio and addition of oxygen were advantageous.Ethanol conversion and hydrogen yield increased with the vaporization room temperature up to the maximum at first,and then decreased slightly.The maximum hydrogen yield of 31.8% was obtained at an ethanol conversion of 88.4% under the optimum operation conditions of vaporization room temperature of 120℃,ethanol flux of 0.18 mL/min,water/ethanol ratio of 7.7 and oxygen volume concentration of 13.3%.

  11. Analysis of the financial impacts to the industrial energy user of using coal or municipal solid waste in a new process-steam-generating plant

    Energy Technology Data Exchange (ETDEWEB)

    1983-10-01

    An analysis is presented of the financial impacts to the industrial energy user of using either coal or MSW in a new process-steam-generating plant. The results of the analysis indicate that the use of coal or solid waste, rather than oil, in a new energy production plant represents an attractive investment. The financial analysis is based on replacing an existing oil-fired plant with a new plant financed via 100-percent debt. The analysis was structured to cover a range of steam demands, different plant ownership and operating structures, and the tax benefits available to these types of plants. Information is also provided on the types of technologies that would be appropriate given the assumed steam demands. In addition, information is provided on available tax benefits in light of recent tax law changes. Nine options for new coal and MSW plants were analyzed, reflecting a matching of technology and energy output to various process steam demands, as well as different ownership and operating structures.

  12. Scenarios for the transmutation of actinides in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, Bronwyn, E-mail: hylandb@aecl.ca [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Gihm, Brian, E-mail: gihmb@aecl.ca [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

    2011-12-15

    With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100-1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

  13. Algorithm for the calculation of a steam generator efficiency; Algoritmo para el calculo de la eficiencia de un generador de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Franco, David; Ambriz, Juan Jose; Romero Paredes, Hernando [Universidad Autonoma Metropolitana-Iztapalapa, Mexico, D. F. (Mexico)

    1994-12-31

    The efficiency calculation of steam generators is not always simple. The purpose of this paper is to propose an algorithm for the calculation of steam generators efficiency, easy to understand and carry out, in the form of a series of steps to be followed. It takes as starting point that the person in charge of applying these calculations has knowledge of the combustion processes and thermodynamic principles that rule such processes. [Espanol] El calculo de la eficiencia de los generadores de vapor no siempre es sencillo, el presente trabajo tiene como objetivo el de proponer un algoritmo de calculo de eficiencia de generadores de vapor, el cual sea facil de entender y de llevar a cabo, en forma de una serie de pasos a seguir. Se toma como punto de partida, que la persona encargada de aplicar estos calculos tenga el conocimiento de los procesos de combustion y principios termodinamicos que rigen tales procesos.

  14. Development of a water leak detection system for LMFBR steam generators. Pt. 2; General planning of sensor arrangement for active acoustic method

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Kazuo; Kumagai, Hiromichi (Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.)

    1994-04-01

    Development of a water leak detection system with short response time and high sensitivity for LMFBR steam generators is required to prevent failure propagation and to maintain structural integrity of steam generators. A new type of leak detection method, active acoustic method, which observes gas bubbles accompanying the leak using sonic waves is being developed. In this study, some series of experiments are carried out to investigate; (1) attenuation of sonic wave in a typical SG structure, (2) suitable method to attach waveguides to the SG shell, and (3) possibility of reflex method. Furthermore, a reference sensor arrangement for active acoustic method is selected based on the experimental results as the basis of future studies. (author).

  15. Assessment of DUPIC fuel compatibility with CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Choi, H. B.; Roh, G. H.; Jeong, C. J.; Rhee, B. W.; Choi, J. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    The compatibility of DUPIC fuel with the existing CANDU reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will require a quantitative analysis later. (author).

  16. An emergency water injection system (EWIS) for future CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre L.F. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: momarques@uol.com.br; Todreas, Neil E.; Driscoll, Michael J. [Massachusetts Inst.of Tech., Cambridge, MA (United States). Nuclear Engineering Dept.

    2000-07-01

    This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m{sup 2}: the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

  17. Advancement of safeguards inspection technology for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Sung; Park, W. S.; Cha, H. R.; Ham, Y. S.; Lee, Y. G.; Kim, K. P.; Hong, Y. D

    1999-04-01

    The objectives of this project are to develop both inspection technology and safeguards instruments, related to CANDU safeguards inspection, through international cooperation, so that those outcomes are to be applied in field inspections of national safeguards. Furthermore, those could contribute to the improvement of verification correctness of IAEA inspections. Considering the level of national inspection technology, it looked not possible to perform national inspections without the joint use of containment and surveillance equipment conjunction with the IAEA. In this connection, basic studies for the successful implementation of national inspections was performed, optimal structure of safeguards inspection was attained, and advancement of safeguards inspection technology was forwarded. The successful implementation of this project contributed to both the improvement of inspection technology on CANDU reactors and the implementation of national inspection to be performed according to the legal framework. In addition, it would be an opportunity to improve the ability of negotiating in equal shares in relation to the IAEA on the occasion of discussing or negotiating the safeguards issues concerned. Now that the national safeguards technology for CANDU reactors was developed, the safeguards criteria, procedure and instruments as to the other item facilities and fabrication facilities should be developed for the perfection of national inspections. It would be desirable that the recommendations proposed and concreted in this study, so as to both cope with the strengthened international safeguards and detect the undeclared nuclear activities, could be applied to national safeguards scheme. (author)

  18. The stress–strain state of the cracked welded joint between the header and the shell of PGV-1000M steam generator

    Directory of Open Access Journals (Sweden)

    S. M. Ban’ko

    2014-10-01

    Full Text Available The three-dimensional elastoplastic stress–strain state of the cracked welded joint between the “hot” header and the shell of PGV-1000M steam generator is numerically analyzed. The crack is located on the inside surface of the connector pipe, near the fillet. The effect of the loading history on the crack-tip stress-intensity factor is assessed.

  19. The chemistry and properties of organic boiler feedwater additives based on film-forming amines and their use in steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Hater, Wolfgang; Russchuetzky, Niels [BK Giulini GmbH, Duesseldorf (Germany); Olivet, David [BK Giulini, Barcelona (Spain)

    2009-02-15

    Film-forming amines have been successfully used for a number of decades to treat boiler feedwater, especially in industrial power plants. The results of recent studies of their properties and the results of operational trials should close the existing gaps in our knowledge of film-forming amines, so that this technology can be incorporated into the appropriate guidelines for the treatment of steam generators. (orig.)

  20. 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

    Energy Technology Data Exchange (ETDEWEB)

    Kurtz, R.J.; Heasler, P.G.; Baird, D.B. [Pacific Northwest Lab., Richland, WA (United States)

    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.

  1. 低品质蒸汽发电实践及节能分析%Practice and energy-saving analysis of power generation wi th saturated steam

    Institute of Scientific and Technical Information of China (English)

    许相波; 陈小芸

    2015-01-01

    Technical proposal of power generators with saturated steam in iron and steel enterprise was discussed , based on the generator used in Meishan Iron and Steel Company , characteristic parameters were introduced and energy -saving effect was mainly analyzed .It was shown that power generators with saturated steam could be adapted to the steam recovered by waste heat in iron company and energy-saving effect was significant .%论述了钢铁企业建设饱和蒸汽发电机组的技术方案,以梅钢应用案例为对象,介绍了发电设备的特性参数,重点分析了其节能效果。实践表明,饱和蒸汽发电机组能够适应钢铁企业余热回收的低品质蒸汽,节能效果明显。

  2. SAFIRE - a robotic inspection system for CANDU feeders

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, R. [OC Robotics, Bristol (United Kingdom)

    2011-07-01

    The condition of primary circuit feeder pipes in CANDU reactors is relevant to the commercial viability and plant life. One known wear mechanism is external fretting between feeder pipes and adjacent services or support structures, particularly within the Upper Feeder Cabinet (UFC). Fretting leads to wall thinning which must not exceed certain agreed limits. Chafe shields have been added to protect the feeder pipes. Regular inspections are required of the chafe shields, feeder pipes and other structures that may cause feeder damage. Historically, the dose received by inspectors conducting this work has been significant. For this reason Ontario Power Generation has invested in a remotely operated robot system to conduct visual inspections within the UFC. This system, called SAFIRE for 'Snake-Arm Feeder Inspection Robot Equipment' has been deployed at Pickering during 2010 and 2011 and has been used to inspect areas that are extremely difficult to inspect with existing manual techniques. The 2011 scope of work included inspection of a total of 660 feeder pipes in three UFC quadrants, in two reactors. The full scope was completed over a one-month period in Autumn 2011 in which SAFIRE was used during 23, twelve hour shifts. This included two periods each of 72 hours of continuous operation using multiple teams of operators. SAFIRE is remote controlled delivery system for multiple cameras to record still images and video. The main system elements include a snake-arm robot mounted on a mobile vehicle. It can be controlled from up to 500m away using a fibre/copper connection. The snake-arm is 2.2m long, 25mm wide and has 18 degrees of freedom. It is designed to snake between the rows of feeder pipes to inspect feeder/hanger interfaces, both above and below the feeder cabinet catwalks. Future upgrades offer the potential to add additional tools to increase functionality. This paper describes the SAFIRE development process from inception to operational experience

  3. Component Test Facility (Comtest) Phase 1 Engineering For 760°C (1400°F) Advanced Ultrasupercritical (A-USC) Steam Generator Development

    Energy Technology Data Exchange (ETDEWEB)

    Weitzel, Paul [Babcock & Wilcox Power Generation Group, Inc., Barberton, OH (United States)

    2016-05-13

    The Babcock & Wilcox Company (B&W) performed a Pre-Front End Engineering Design (Pre-FEED) of an A-USC steam superheater for a proposed component test program achieving 760°C (1400°F) steam temperature. This would lead to follow-on work in a Phase 2 and Phase 3 that would involve detail design, manufacturing, construction and operation of the ComTest. Phase 1 results have provided the engineering data necessary for proceeding to the next phase of ComTest. The steam generator superheater would subsequently supply the steam to an A-USC prototype intermediate pressure steam turbine. The ComTest program is important in that it will place functioning A-USC components in operation and in coordinated boiler and turbine service. It is also important to introduce the power plant operation and maintenance personnel to the level of skills required and provide the first background experience with hands-on training. The project will provide a means to exercise the complete supply chain events required in order to practice and perfect the process for A-USC power plant design, supply, manufacture, construction, commissioning, operation and maintenance. Representative participants will then be able to transfer knowledge and recommendations to the industry. ComTest is conceived in the manner of using a separate standalone plant facility that will not jeopardize the host facility or suffer from conflicting requirements in the host plant’s mission that could sacrifice the nickel alloy components and not achieve the testing goals. ComTest will utilize smaller quantities of the expensive materials and reduce the risk in the first operational practice for A-USC technology in the United States. Components at suitable scale in ComTest provide more assurance before putting them into practice in the full size A-USC demonstration plant.

  4. Evaluation of cracking in feedwater piping adjacent to the steam generators in Nine Pressurized Water Reactor Plants

    Energy Technology Data Exchange (ETDEWEB)

    Goldberg, A.; Streit, R.D.; Scott, R.G.

    1980-06-25

    Cracking in ASTM A106-B and A106-C feedwater piping was detected near the inlet to the steam generators in a number of pressurized water reactor plants. We received sections with cracks from nine of the plants with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Variations were observed in piping surface irregularities, corrosion-product, pit, and crack morphology, surface elmental and crystal structure analyses, and steel microstructures and mechanical properties. However, with but two exceptions, namely, arrest bands and major surface irregularities, we were unable to relate the extent of cracking to any of these factors. Tensile and fracture toughness (J/sub Ic/ and tearing modulus) properties were measured over a range of temperatures and strain rates. No unusual properties or microstructures were observed that could be related to the cracking problem. All crack surfaces contained thick oxide deposits and showed evidence of cyclic events in the form of arrest bands. Transmission electron microscopy revealed fatigue striations on replicas of cleaned crack surfaces from one plant and possibly from three others. Calculations based on the observed striation spacings gave a value of ..delta..sigma = 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses. Although surface irregularities and corrosion pits were sources for crack initiation and corrosion may have contributed to crack propagation, it is proposed that the overriding factor in the cracking problem is the presence of unforeseen cyclic loads.

  5. Regenerative superheated steam turbine cycles

    Science.gov (United States)

    Fuller, L. C.; Stovall, T. K.

    1980-01-01

    PRESTO computer program was developed to analyze performance of wide range of steam turbine cycles with special attention given to regenerative superheated steam turbine cycles. It can be used to model standard turbine cycles, including such features as process steam extraction, induction and feedwater heating by external sources, peaking, and high back pressure. Expansion line efficiencies, exhaust loss, leakages, mechanical losses, and generator losses are used to calculate cycle heat rate and generator output. Program provides power engineer with flexible aid for design and analysis of steam turbine systems.

  6. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  7. Application of Low-pressure Saturated Steam Power Generating Technology in Steel Enterprises%低压饱和蒸汽发电技术在钢铁企业的应用

    Institute of Scientific and Technical Information of China (English)

    刘颖; 马永锋

    2016-01-01

    介绍了钢铁企业饱和蒸汽发电主要的应用技术,并通过某钢厂的低压饱和蒸汽利用方案,对比了饱和蒸汽汽轮机和饱和蒸汽螺杆发电机应用的优缺点.%The main application technology of saturated steam power generating in steel enterprises is introduced. Taking the utilization program of low-pressure saturated steam pow-er generating of some steelmaker as an example, the advantages and disadvantages of saturat-ed steam turbine generator and saturated steam screw generator are compared.

  8. Modelling and simulation of a circulating fluidized-bed steam generator as an aid for process analysis and automation. Modellierung und Simulation eines ZWS-Dampferzeugers als Hilfsmittel zur Prozessanalyse und -automatisierung

    Energy Technology Data Exchange (ETDEWEB)

    Karbach, A.; Peters, R.; Schaub, G. (Lurgi GmbH, Frankfurt am Main (Germany, F.R.))

    1990-04-01

    This book deals with the development and application of mathematical model for the simulation of a steam generator with fluidized-bed combustion (coal combustion in the circulating fluidized-bed combustion). (orig./EF).

  9. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  10. Wet steam wetness measurement in a 10 MW steam turbine

    OpenAIRE

    Kolovratník Michal; Bartoš Ondřej

    2014-01-01

    The aim of this paper is to introduce a new design of the extinction probes developed for wet steam wetness measurement in steam turbines. This new generation of small sized extinction probes was developed at CTU in Prague. A data processing technique is presented together with yielded examples of the wetness distribution along the last blade of a 10MW steam turbine. The experimental measurement was done in cooperation with Doosan Škoda Power s.r.o.

  11. Wet steam wetness measurement in a 10 MW steam turbine

    Directory of Open Access Journals (Sweden)

    Kolovratník Michal

    2014-03-01

    Full Text Available The aim of this paper is to introduce a new design of the extinction probes developed for wet steam wetness measurement in steam turbines. This new generation of small sized extinction probes was developed at CTU in Prague. A data processing technique is presented together with yielded examples of the wetness distribution along the last blade of a 10MW steam turbine. The experimental measurement was done in cooperation with Doosan Škoda Power s.r.o.

  12. Transient analysis with high percentages steam generator tube plugging of Angra 1 nuclear power plant; Analise de transientes com altos percentuais de tamponamento dos tubos dos geradores de vapor de Angra 1

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Marcio Poubel; Martins Junior, Laercio Lucena; Vanni, Enio Antonio; Machado, Marcio Dornellas; Moreira, Francisco Jose [ELETRONUCLEAR, Rio de Janeiro, RJ (Brazil); Alvim, Antonio Carlos M. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    1999-11-01

    The present work is part of several analyses under development in ELETRONUCLEAR/ COPPE-UFRJ to evaluate impacts on licensing bases and operating conditions of an increase in steam generator tube plugging for Angra 1 NPP. Total loss of reactor coolant flow uncontrolled boron dilution transients were initially analysed. The final results indicated that were no impacts on FSAR established margins, in case of 24% steam generator tube plugging. (author) 3 refs., 4 figs., 4 tabs.

  13. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-03-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA analysis. There are three main area of model development, i.e. moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version.

  14. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Draft, development and optimization of a fuel cell system for residential power generation with steam reformer; Entwurf, Aufbau und Optimierung eines PEM-Brennstoffzellensystems zur Hausenergieversorgung mit Dampfreformer

    Energy Technology Data Exchange (ETDEWEB)

    Brandt, H.

    2006-05-17

    The first development cycle of a residential power generation system is described. A steam reformer was chosen to produce hydrogen out of natural gas. After carbon monoxide purification with a preferential oxidation (PrOx) unit the hydrogen rich reformat gas is feed to the anode of the PEM-fuel cell, where due to the internal reaction with air oxygen form the cathode side water, heat and electricity is produced. Due to an incomplete conversion the anode off gas contains hydrogen and residual methane, which is feed to the burner of the steam reformer to reduce the needed amount of external fuel to heat the steam reformer. To develop the system the components are separately investigated and optimized in their construction or operation to meet the system requirements. After steady state and dynamic characterization of the components they were coupled one after another to build the system. To operate the system a system control was developed to operate and characterize this complex system. After characterization the system was analyzed for further optimization. During the development of the system inventions like a water cooled PrOx, an independent fuel cell controller or a burner for anodic off gas recirculation were made. The work gives a look into the interactions between the components and allows to understand the problems by coupling such components. (orig.)

  17. Experimental Investigation on Rocket Steam Generator with High Flowrate%大流量火箭蒸汽发生器试验研究

    Institute of Scientific and Technical Information of China (English)

    赵宏; 张海栋; 孙小丽; 郑鑫

    2013-01-01

    In order to meet the demand of large flow rate steam supply,the rocket steam generator,u sing of liquid oxygen and alcohol,was developed and tested on the ground bench.The test results show that,the pressure build up time in combustion chamber is about two seconds,and typical steam parameters,such as pressure,flow rate,temperature and water vapor volume concentration,were 0.86MPa,40kg/s,254℃,80%,respectively.When total tested time exceeds thousand seconds,oxidation film formed on the injector panel can effectively prevent the spread of hot gas to the internal,and enhance the life duration of injector.As large flow steam source,the rocket steam generator may be applicable for altitude simulation fa cility.%为满足大流量高能引射工质发生装置研制的需要,在地面试验台上,对自行研制的大流量液氧酒精火箭蒸汽发生器进行了试验研究.试验结果表明,蒸汽发生器起动建压时间约2.0秒,主级段各性能参数平稳,蒸汽流量约40kg/s,蒸汽压力约0.86MPa,蒸汽温度约254℃,水蒸汽体积浓度80%.喷注器面板的氧化膜能够有效地阻止高温燃气向喷注面内部的扩散,延长喷注面寿命.大流量火箭蒸汽发生器可作为大型主动引射高空模拟试验台的蒸汽源.

  18. 利用低压饱和蒸汽发电的实例%Examples of Power Generation by Use of Low-pressure Saturated Steam

    Institute of Scientific and Technical Information of China (English)

    莫宾; 蔡鸣; 蒋洁

    2011-01-01

    Use of the low-pressure saturated steam which is the byproduct of the patent technology of "thermal phosphoric acid production of heat recovery",for power generation,the problem of excess low-pressure saturated steam in the thermal phosphoric acid production enterprises is solved.Via examples,the performance characteristics of generating units,power generation process and connection considerations with the external network are explained.With this technology,the electricity load in thermal phosphoric acid can be met,product energy consumption and cost can be reduced.It provides a new way for the comprehensive utilization of low-pressure saturated steam in thermal phosphoric acid production for the domestic enterprises.%利用"热法磷酸生产热能回收"专利技术副产的低压饱和蒸汽发电,解决了热法磷酸生产企业低压饱和蒸汽过剩的问题。通过实例,讲述了发电机组的性能特点、发电的工艺流程及与外网联接的注意事项。通过此技术的应用,基本能满足热法磷酸装置用电负荷,可降低产品的能耗和成本,达到节能减排的效果,为国内热法磷酸生产企业和其它行业低压饱和蒸汽的综合利用提供了一条新途径。

  19. Development of CANDU pressure tube integrity evaluation system

    Energy Technology Data Exchange (ETDEWEB)

    Kwac, S. L.; Kim, Y. J. [Sungkyunkwan Univ., Seoul (Korea, Republic of); Lee, J. S. [Kyonggi Univ., Suwon (Korea, Republic of); Park, Y. W. [KINS, Taejon (Korea, Republic of)

    1999-05-01

    The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw or contact with their calandria tubes is found during the periodic inspection, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to perform the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the integrity evaluation process. For this reason, an integrity evaluation system was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL. The evaluation procedure includes the crack growth calculation both by DHC and by fatigue. It also provides the prediction of fracture initiation, plastic collapse and leak-before-break(LBB), blister formation and blister growth. This system provides various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

  20. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building.

  1. In-reactor performance of pressure tubes in CANDU reactors

    Science.gov (United States)

    Rodgers, D. K.; Coleman, C. E.; Griffiths, M.; Bickel, G. A.; Theaker, J. R.; Muir, I.; Bahurmuz, A. A.; Lawrence, S. St.; Resta Levi, M.

    2008-12-01

    The pressure tubes in CANDU reactors have been operating for times up to about 25 years. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behaviour and discusses the factors controlling the behaviour of these components in currently operating CANDU reactors. The mechanical properties (such as ultimate tensile strength, UTS, and fracture toughness), and delayed-hydride-cracking properties (crack growth rate Vc, and threshold stress intensity factor, KIH) change with irradiation; the former reach a limiting value at a fluence of Pressure tubes exhibit elongation and diametral expansion. The deformation behaviour is a function of operating conditions and material properties that vary from tube-to-tube and as a function of axial location. Semi-empirical predictive models have been developed to describe the deformation response of average tubes as a function of operating conditions. For corrosion and, more importantly deuterium pickup, semi-empirical predictive models have also been developed to represent the behaviour of an average tube. The effect of material variability on corrosion behaviour is less well defined compared with other properties. Improvements in manufacturing have increased fracture resistance by minimising trace elements, especially H and Cl, and reduced variability by tightening controls on forming parameters, especially hot-working temperatures.

  2. Procurement and supply of CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bazeley, E.G. [E.G. Bazeley and Associates, Whitby, Ontario (Canada)

    2002-11-01

    In 1955 a decision was made to proceed with construction of a Nuclear Power Demonstration Station (NPD) near Rolfton, Ontario. This project, headed by Atomic Energy of Canada with major involvement of private industry, was the genesis for the development of nuclear electric generation in Canada. This paper reviews one aspect of the Canadian program: the evolution of fuel procurement and supply, which in itself has been a remarkable Canadian achievement. (author)

  3. Acceptance test for 900 MWe PWR unit replacement steam generators; Essai de reception des generateurs de vapeur de remplacement des tranches REP 900

    Energy Technology Data Exchange (ETDEWEB)

    Gourguechon, B.

    1993-12-31

    During the first half of 1994, the Gravelines 1 steam generators will be replaced (SG replacement procedure). The new SG`s differ from the former components notably by the alloy used for the tube bundle, in this case, the high chromium content Inconel 690. So, from this standpoint, they are to be considered as PWR 900 replacement SG first models and their thermal efficiency has consequently to be assessed. This will provide an opportunity of ensuring that the performance of the components delivered is in compliance with requirements and of making the necessary provisions if significant deviations are observed. The EFMT branch, which has been in charge of the instrumentation and acceptance of the different SG first models since the first PWR plants were commissioned, will be responsible for the acceptance tests and the ultimate validation of a performance assessment procedure applicable to the future replacement steam generators. The methods and tests proposed for SG expert appraisal are based on consideration of the importance of primary measurement quality for satisfactory SG assessment and of the new test facilities with which the 900 and 1 300 PWR plants are gradually being equipped. These facilities provide an on-site computer environment for tests compatible with the tools (PATTERN, etc.) used at EFMT and in other departments. This test is the first of this kind performed by EFMT and the test facility of a nuclear power plant. (author). 6 figs.

  4. Doppler method leak detection for LMFBR steam generators. Pt. 2. Detection characteristics of bubble in-water using large scale SG model

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Hiromichi [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab

    2000-06-01

    To prevent the expansion of tube damage and to maintain structural integrity in the steam generators (SGs) of a fast breeder reactor (FBR), it is necessary to detect precisely and immediately the leakage of water from heat transfer tubes. Therefore, an active acoustic method was developed. Previous studies have revealed that, in practical steam generators, the active acoustic method can detect bubbles of 10 l/s within 10 seconds. However to prevent the expansion of damage to neighboring tubes, it is necessary to detect smaller leakages of water from the heat transfer tubes. The Doppler method is designed to detect small leakages and to find the source of a leak before damage spreads to neighboring tubes. The detection sensitivity of the Doppler method and the influence of background noise were investigated experimentally. In-water experiments were performed using an SG full-sector model that simulates actual SGs. The results show that the Doppler method can detect bubbles of 0.1 l/s (equivalent to a water leak rate of about 0.1 g/s) within a few seconds and that the background noise has little effect on water leak detection performance. The Doppler method thus has great potential for the detection of water leakage in SGs. (author)

  5. First experience with steam generator diagnostic systems and neuronal networks at RWE Energie AG; Erste Erfahrungen mit Diagnosesystemen und Neuronalen Netzen bei der RWE Energie AG

    Energy Technology Data Exchange (ETDEWEB)

    Moll, W.; Puetter, J.; Pollack, M. [RWE Energie AG, Essen (Germany)

    2000-07-01

    With the aid of diagnosis systems installed at three reference plants it is possible to support an economic and trouble-free operation of steam generator plants. The steam generator diagnostic systems analyse and assess momentary plant conditions by means of thermodynamic recalculation. They can deploy boiler cleaning devices in a targeted and optimised manner. Subsequent to the present optimisation runs for different coal charges, the control of boiler cleaning devices is to become automatic. Diagnostic systems and optimisation software on the basis of neural networks can additionally be used for optimisation of the entire process. (orig.) [German] Mit Hilfe der an drei Referenzanlagen nachgeruesteten Diagnosesysteme ist es moeglich, einen wirtschaftlichen und stoerungsfreien Betrieb der Dampferzeugeranlagen zu unterstuetzen. Die Kesseldiagnosesysteme analysieren und bewerten mittels einer thermodynamischen Nachrechnung den momentanen Anlagenzustand. Durch sie koennen die Kesselreinigungseinrichtungen gezielt und optimiert eingesetzt werden. Im Anschluss an die zur Zeit laufende Optimierungsphase fuer verschiedene Kohlequalitaeten wird die Ansteuerung der Kesselreinigungseinrichtungen automatisiert. Diagnosesysteme und Neuronale Netze koennen zur Optimierung des Gesamtprozesses ergaenzend eingesetzt werden. (orig.)

  6. Heat Recovery Steam Generators for Combined Typical Cycle Steam Purge%9F等级燃气蒸汽联合循环机组余热锅炉吹管方式的选择

    Institute of Scientific and Technical Information of China (English)

    许宏宇

    2009-01-01

    根据余热锅炉三压系统不同压力级选择吹管方式,提高蒸汽吹扫的效果.%According to the different ranks of pressurizing system for the HRSG,we can select the steam-line blowing way to improve the purging effect of steam.

  7. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Shill, R. [Atomic Energy Of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Chung, S.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2005-03-15

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. (author)

  8. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAP5.MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAP5 source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version. 6 refs., 15 figs., 7 tabs. (Author)

  9. CURRENT STATUS OF INTEGRITY ASSESSMENT BY SIPPING SYSTEM OF SPENT FUEL BUNDLES IRRADIATED IN CANDU REACTOR

    Directory of Open Access Journals (Sweden)

    JONG-YOUL PARK

    2014-12-01

    Full Text Available In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.

  10. Operating Experience of MACSTOR Modules at CANDU 6 Stations

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, Robert R. [Atomic Energy Canada Ltd., Chalk River (Canada)

    2005-11-15

    Over the last three decades, Atomic Energy of Canada Limited (AECL) has contributed to the technology development and implementation of dry spent fuel management facilities in Canada, Korea and Romania During that period, AECL has developed a number of concrete canister models and the MACSTOR200 module, a medium size air-cooled vault with a 228 MgU (Mega grams of Uranium) capacity. AECL's dry storage technologies were used for the construction of eight large-scale above ground dry storage facilities for CANDU spent fuel. As of 2005, those facilities have an installed capacity in excess of 5,000 MgU. Since 1995, the two newest dry storage installations built for CANDU 6 reactors at Gentilly 2 (Canada) and Cernavoda (Romania) used the MACSTOR 200 module. Seven such modules have been built at Gentilly 2 during the 1995 to 2004 period and one at Cernavoda in 2003. The construction and operating experience of those modules is reviewed in this paper. The MACSTOR 200 modules were initially designed for a 50-year service life, with recent units at Gentilly 2 licensed for a 100-year service life in a rural (non-maritime) climate. During the 1995-2005 period, six of the eight modules were loaded with fuel. Their operation has brought a significant amount of experience on loading operations, performance of fuel handling equipment, radiation shielding, heat transfer, monitoring of the two confinement boundaries and radiation dose to personnel. Heat dissipation performance of the MACSTOR 200 was initially licensed using values derived from full scale tests made at AECL's Whiteshell Research Laboratories, that were backed-up by temperature measurements made on the first two modules. Results and computer models developed for the MACSTOR 200 module are described. Korea Hydro and Nuclear Power (KHNP) and its subsidiary Nuclear Environment Technology Institute (NETEC), in collaboration with Hyundai Engineering Company Ltd. (HEC) and AECL, are developing a new dry storage

  11. Proceedings of the fifth international conference on CANDU fuel. V.1,2

    Energy Technology Data Exchange (ETDEWEB)

    Lau, J.H. [ed.

    1997-07-01

    The First International Conference on CANDU Fuel was held in Chalk River in 1986. The CANDU Fuel community has gathered every three years since. The papers presented include topics on international experience, CANFLEX fuel bundles, Fuel design, Fuel modelling, Manufacturing and Quality assurance, Fuel performance and Safety, Fuel cycles and Spent Fuel management. Volume One was published in advance of the conference and Volume Two was printed after the conference.

  12. Estimation of fouling factors in the design of steam generators; Estimacion de factores de ensuciamiento en el diseno de generadores de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Flores Archila, David Ivanhoe

    1997-12-31

    This thesis proposes an alternative to correct design problems the evaluation of fouling factors of thermal energy transfer components of a power station generator. That alternative involves the development of a subroutine to apply fundamental empirical thermal transfer correlations, written in FORTRAN language and extends a principal program for steam generator thermal analysis, developed by Instituto de Investigaciones Electricas. [Espanol] Esta tesis propone una alternativa, basada en el calculo de factores de ensuciamiento de elementos de transferencia de energia termica de un generador de vapor, como mecanismo para corregir el diseno deficiente de areas de calefaccion. Esta alternativa involucra el desarrollo de una subrutina, escrita en FORTRAN, la cual realiza el calculo de los factores de ensuciamiento empleando correlaciones fundamentadas en la teoria de los procesos de transferencia de calor y forma parte de un programa principal para el analisis termico de generadores de vapor desarrollado en el Instituto de Investigaciones Electricas.

  13. 烧结生产线的饱和蒸汽发电系统%Power generation system with saturated steam in sintering line

    Institute of Scientific and Technical Information of China (English)

    张青枝; 陈恩鉴; 胡巍

    2013-01-01

    针对烧结线余热电站运行中短时间停机造成的问题,提出了一种烧结线饱和蒸汽发电系统,该系统可以克服烧结线停机带来的发电量减少的问题,并与现有系统进行了比较.%According to the problems by sintering machine shutdown in short time during sintering waste heat power station operation,a sintering power generation system with saturated steam was put forward. This system can solve the problem of generated energy reduction by sintering machine shutdown. A comparison was made between this system and existing system.

  14. Development of a water leak detection system for LMFBR steam generator. Pt. 3. Experimental results for detection of bubbles using the SG sector model

    Energy Technology Data Exchange (ETDEWEB)

    Kumagai, Hiromichi; Yoshida, Kazuo [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

    1995-05-01

    In order to prevent the expansion of tube damages and to maintain structural safety in steam generators (SG) of liquid metal fast breeder reactor (LMFBR), it is necessary to detect precisely and immediately the leakage of water from tubes of heat exchangers. The active acoustic detection method, which detects the sound attenuation due to bubbles generated at the sodium-water reactions, has drawn general interests owing to its short response time and reduction of the influence of background noise. Sound attenuation is also subjected to structures such as heat transfer tubes and shrouds. Accordingly, it is necessary to evaluate the sound attenuation due to structures. However, studies in these respects are very few. In this paper, using the water bath and SG sector model, the attenuation characteristics of sounds due to flat plates and heat transfer tubes are investigated under various conditions and discussed. (author).

  15. Observer-based delay-independent control for steam valves of steam turbo-generator%基于观测器的汽轮发电机气门开度的时滞无关控制

    Institute of Scientific and Technical Information of China (English)

    孙妙平; 年晓红; 潘欢

    2012-01-01

    本文考虑了具有非线性关联作用的带中间再热器的汽轮发电机的数学模型,研究了气门开度基于观测状态的时滞无关分散控制器的设计问题.首先把非线性关联函数变换为子系统状态变量的二次有界不等式,然后通过构造适当的Lypunov泛函,并利用线性矩阵不等式(LMI)的处理方法,得到了使汽轮发电机组渐近稳定的LMI充分条件.此外,还提出了了控制器增益矩阵和观测器增益矩阵的求解算法..最后以两机无穷大母线系统为例进行了仿真分析,验证了该方法的有效性.%A delay-independent decentralized controller based on the state observer is proposed for the steam valve opening of a turbo-generator. The design of this controller is based on the mathematical model of the turbo generator with reheater, which involves nonlinear interconnection function. The nonlinear interconnection function is first converted into a bounded quadratic inequality of the subsystem states; and then, by constructing a proper Lyapunov function and applying the linear matrix inequalities (LMI) method, we develop for the turbo-generator the sufficient condition of asymptomatic stability and determine the gain matrices for the controller and observer. Simulation has been performed in a two-machine infinite-bus power system; results demonstrate the effectiveness of the proposed method.

  16. Electric waste gas purification of a waste wood fired steam generator. Final report. Elektrische Abgasreinigung eines 'Abfallholzbefeuerten Dampfkessels'. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Brinkmann, W.

    1984-11-01

    The crude gas dust content of smoke gases from waste wood fired steam generators is influenced considerably by the type and quality of the waste combustibles. When firing wood waste with a high proportion of fine particles and ash, the smoke gases normally have a higher crude gas content compared to clean wood waste combustibles consisting of coarse particles. It was necessary to provide documents for the design of a suitable smoke gas dedusting system by means of corresponding firing and measuring programmes. After having evaluated the documents provided, it became clear that the use of an electric filter system is the most suitable for smoke gas dedusting of steam boilers fired by wood waste combustibles in the broadest sense, in the form of particles as well as in a blowable form. After putting the system into operation, it was possible to prove that the pure gas dust content was certainly lower than requested in the 'TA air' in all operating stages. (orig.).

  17. Studi Numerik Karakteristik Aliran dan Perpindahan Panas Pada Heat Recovery Steam Generator di PT Gresik Gases and Power Indonesia (Linde Indonesia

    Directory of Open Access Journals (Sweden)

    Dhika Suryananda

    2012-09-01

    Full Text Available Pertumbuhan ekonomi berdampak pada meningkatnya kebutuhan energi, sehingga menuntut peningkatan efisiensi dari power plant sebagai salah satu produsen energi. Pada saat ini power plant yang memiliki efisiensi paling tinggi adalah combined cycle power plant. Pada sistem combined cycle tersebut terdapat komponen Heat Recovery Steam Generator (HRSG yang berfungsi untuk meningkatkan efisiensi dari power plant dengan  cara menggunakan sisa panas dari gas buang  (exhaust gas turbine dan digunakan untuk memproduksi uap (steam untuk proses selanjutnya. Penelitian ini dilakukan menggunakan metode numerik (CFD dengan software FLUENT 6.3.26. Pemodelan yang dilakukan pada penelitian ini adalah 3 dimensi, aliran steady, turbulence model yang dipakai Relizable k-ε model dengan reaksi pembakarannya menggunakan spesies transport. Mixture materials yang digunakan merupakan methane-air. Data yang digunakan dalam penelitian ini menggunakan data yang di ambil di PT. GRESIK GASES and POWER INDONESIA.. Hasil yang didapatkan pada simulasi ini adalah bentuk bodi seperti enlargement, contraction, dan elbow memiliki pengaruh yang sangat besar terhadap distribusi temperatur, terkanan, dan kecepatan pada HRSG. Error dari hasil simulasi numerik dan referensi CCR sebagai berikut pada secondary superheater sebesar 8 %, pada primary superheater sebesar 6%, pada evaporator sebesar 0.00008% dan yang terakhir pada economizer sebesar 92 % . Penyebab perbedaan antara numerik dengan data CCR  adalah kurang akuratnya proses simulasi dan simplifikasi dari jajaran heat exchanger terutama pada bagian economizer.

  18. Potential use of California lignite and other alternate fuel for enhanced oil recovery. Phase I and II. Final report. [As alternative fuels for steam generation in thermal EOR

    Energy Technology Data Exchange (ETDEWEB)

    Shelton, R.; Shimizu, A.; Briggs, A.

    1980-02-01

    The Nation's continued reliance on liquid fossil fuels and decreasing reserves of light oils gives increased impetus to improving the recovery of heavy oil. Thermal enhanced oil recovery EOR techniques, such as steam injection, have generally been the most effective for increasing heavy oil production. However, conventional steam generation consumes a large fraction of the produced oil. The substitution of alternate (solid) fuels would release much of this consumed oil to market. This two-part report focuses on two solid fuels available in California, the site of most thermal EOR - petroleum coke and lignite. Phase I, entitled Economic Analysis, shows detailed cost comparisons between the two candidate fuels and also with Western coal. The analysis includes fuels characterizations, process designs for several combustion systems, and a thorough evaluation of the technical and economic uncertainties. In Phase II, many technical parameters of petroleum coke combustion were measured in a pilot-plant fluidized bed. The results of the study showed that petroleum coke combustion for EOR is feasible and cost effective in a fluidized bed combustor.

  19. Simulation of thermal fluid dynamics in parabolic trough receiver tubes with direct steam generation using the computer code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, Alexander; Merk, Bruno [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Hirsch, Tobias; Pitz-Paal, Robert [DLR Deutsches Zentrum fuer Luft- und Raumfahrt e.V., Stuttgart (Germany). Inst. fuer Solarforschung

    2014-06-15

    In the present feasibility study the system code ATHLET, which originates from nuclear engineering, is applied to a parabolic trough test facility. A model of the DISS (DIrect Solar Steam) test facility at Plataforma Solar de Almeria in Spain is assembled and the results of the simulations are compared to measured data and the simulation results of the Modelica library 'DissDyn'. A profound comparison between ATHLET Mod 3.0 Cycle A and the 'DissDyn' library reveals the capabilities of these codes. The calculated mass and energy balance in the ATHLET simulations are in good agreement with the results of the measurements and confirm the applicability for thermodynamic simulations of DSG processes in principle. Supplementary, the capabilities of the 6-equation model with transient momentum balances in ATHLET are used to study the slip between liquid and gas phases and to investigate pressure wave oscillations after a sudden valve closure. (orig.)

  20. Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

    Science.gov (United States)

    Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-12-01

    A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integrated into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.

  1. Supplementary examination of alternative materials in a model steam generator: Volume 3, tube characterization by metallography and transmission electron microscopy: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Krupowicz, J.J.; Stubbins, J.F.; Mehler, M.

    1988-10-01

    The microstructural characteristics of current and candidate alloys for steam generator heat transfer tubing were determined utilizing a variety of techniques. Mill annealed heats of Alloys 690 and 800NG were examined as well as heats of Alloy 600 in the mill annealed, process stabilized, sensitized and thermally treated conditions. Characterization included optical microscopy, transmission electron microscopy and scanning transmission electron microscopy of these materials in their pre- and post-test conditions (i.e., archive and exposure for 11,328 hours to 621/degree/F primary temperature). The results were utilized to elicit comparisons of these materials to stress corrosion cracking resistance in sulfate faulted secondary environments. 83 figs., 8 tabs.

  2. ASME power test code ptc 4.1 for steam generators; Codigo de pruebas de potencia ASME ptc 4.1 para generadores de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Plauchu Alcantara, Jorge Alberto [Plauchu Consultores, Morelia, Michoacan (Mexico)

    2001-07-01

    This presentation is oriented towards those who in this subject have experience in the design and equipment specification, plant projects, factory and field testing, operation or result analyses. An important fraction of the national energy supply, approximately 13%, is applied to the steam generation in the different aspects of the industrial activity, in the electrical industry of public service and in the commercial and services sector. The development of the national programs of energy efficiency verifies this when dedicating to this use of the energy important projects, some of them with support of the USAID. The measurement of the energy utilization or the efficiency of steam generators (or boilers) is made applying some procedure agreed by the parts and the one of greater acceptance and best known in Mexico and internationally is the ASME Power Test Code PTC 4.1 for Steam Generators. The purpose and formality in the determination of efficiency and of steam generation capacity behavior, thermal basic regime or fulfillment of guarantees, radically changes the exigencies of strict attachment to the PTC 4.1 This definition will determine the importance of the test method selected, the deviations and convened exceptions, the influence of the precision and the measurement errors, the consideration of auxiliary equipment, etc. An interpretation or incorrect application of the Test Code has lead and will lead to results and nonreliable decisions. [Spanish] Esta exposicion se orienta a quienes en este tema cuenta con experiencia en diseno y especificacion de equipo, proyecto de planta, pruebas en fabrica y campo, operacion o analisis de resultados. Una fraccion importante de la oferta nacional de energia, 13% aproximadamente, se aplica a la generacion de vapor en diferentes giros de actividad industrial, en la industria electrica, de servicio publico y en el sector de servicios y comercial. El desarrollo de los programas nacionales de eficiencia energetica comprueba

  3. Corrosion detection and monitoring in steam generators by means of ultrasound; Deteccion y monitoreo de corrosion por medio de ultrasonido en generadores de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Chacon Nava, Jose G.; Calva, Mauricio [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Fuentes Samaniego, Raul [Universidad Autonoma de Nuevo Leon (Mexico); Peraza Garcia, Alejandro [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1987-12-31

    The tube and component failures in steam generators due to corrosion cause huge economical losses. In this article the internal corrosion processes (hydrogen attack) and high temperature corrosion are described, as well as the ultrasound techniques used for its detection. The importance of obtaining corrosion rates, which are fundamental parameters for the detection of the tube`s residual life. The purpose is to prevent possible failures that would diminish the power plant availability. [Espanol] Las fallas de tuberia en componentes de generadores de vapor debidas a corrosion ocasionan considerables perdidas economicas. En este articulo se describen los procesos de corrosion interna (ataque por hidrogeno) y corrosion en alta temperatura, asi como tecnicas de ultrasonido empleadas para su deteccion. Se destaca la importancia de obtener valores de velocidad de corrosion, que es un parametro fundamental para la determinacion de la vida residual de tuberias. El proposito es poder prevenir posibles fallas que disminuyan la disponibilidad de centrales termoelectricas.

  4. Potential for low fracture toughness and lamellar tearing on PWR steam generator and reactor coolant pump supports. Resolution of generic technical activity A-12

    Energy Technology Data Exchange (ETDEWEB)

    Snaider, R.P.; Hodge, J.M.; Levin, H.A.; Zudans, J.J.

    1979-10-01

    This report summarizes work performed by the Nuclear Regulatory Commission staff and its contractor, Sandia Laboratories, in the resolution of Generic Technical Activity A-12, ''Potential for Low Fracture Toughness and Lamellar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports.'' The report describes the technical issues, the technical studies performed by Sandia describes the technical issues, the technical studies performed by Sandia Laboratories, the NRC staff's technical positions based on these studies, and the staff's plan for implementing its technical positions. It also provides recommendations for further work. The complete technical input from Sandia Laboratories is appended to the report.

  5. General purpose steam table library :

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, John H.; Belcourt, Kenneth Noel; Nourgaliev, Robert

    2013-08-01

    Completion of the CASL L3 milestone THM.CFD.P7.04 provides a general purpose tabular interpolation library for material properties to support, in particular, standardized models for steam properties. The software consists of three parts, implementations of analytic steam models, a code to generate tables from those models, and an interpolation package to interface the tables to CFD codes such as Hydra-TH. Verification of the standard model is maintained through the entire train of routines. The performance of interpolation package exceeds that of freely available analytic implementation of the steam properties by over an order of magnitude.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Shill, R. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Chung, S.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2004-07-01

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea. The module provides the benefit of occupying significantly less area than the concrete canisters presently used. The modules are designed for a minimum service life of 50 years. During that period, the spent fuel bundles shall be safely stored. This imposes that failure of a fuel bundle element or unacceptable degradation of an existing defect (from reactor operation) does not occur during the dry storage period. The fuel bundles are stored in an air-filled fuel basket that releases 365 Watts on average and a maximum of 390 Watts when rare fuel loading conditions are postulated. In addition, specific accidental air flow cooling conditions are postulated that consist of 100% blockage of all air inlets on one side of the module. These conditions can generate a peak daily fuel temperature of up to 155{sup o}C during a reference hot summer day during the first year of operation. The fuel temperature decreases over the years and also fluctuates due to daily and seasonal temperature variations. At this temperature, fuel elements with intact Zircaloy sheathing will not experience damage. However, for the few fuel bundle elements that are non-leaktight (less than 1 per 37,000), some re-oxidation of UO{sub 2} into higher oxides such as U{sub 3}O{sub 7} / U{sub 4}O{sub 9} and U{sub 3}O{sub 8} will occur. This latter form of Uranium oxide is undesirable due to its lower density that results in a volumetric increase of the pellet that can overstress the fuel element sheathing. The level of fuel pellet

  8. Solar steam supply: Initial operation of a plant

    OpenAIRE

    Krüger, Dirk; Lichtenthäler, Niels; Dersch, Jürgen; Schenk, Heiko; Hennecke, Klaus; Anthrakidis, Anette; Rusack, Markus; Lokurlu, Ahmet; Saidi, Karim; Walder, Marcus; Fischer, Stephan; Wirth, Hans Peter

    2011-01-01

    This paper describes experiences in operating a parabolic trough collector field for process heat supply by direct steam generation. The solar steam generator has been running automatically since its start in 2010 except for a winter pause up to now, August 2011, without any malfunction. It has supplied steam at 4 bar absolute and 143°C to the main production steam line on sunny days. Direct steam generation has proven to be a viable technology to supply saturated steam to an industrial st...

  9. Integrated evolution of the medium power CANDU{sup MD} reactors; Evolution integree des reacteurs CANDU{sup MD} de moyenne puissance

    Energy Technology Data Exchange (ETDEWEB)

    Nuzzo, F. [AECL Accelerators, Kanata, ON (Canada)

    2002-07-01

    The aim of this document is the main improvements of the CANDU reactors in the economic, safety and performance domains. The presentation proposes also other applications as the hydrogen production, the freshening of water sea and the bituminous sands exploitation. (A.L.B.)

  10. An installation for steam conversion of gases

    Energy Technology Data Exchange (ETDEWEB)

    Tabata, K.; Matsumoto, I.

    1983-01-28

    An installation is proposed for steam conversion of a hydrocarbon gas in order to produce an inorganic gas which chiefly consists of H2 and CO in which the line for feeding the hydrocarbon gas has a steam generator which has a microcapillary structure made of sponge metal, inorganic heat resistant fibers of glass, Si02, Al203 or carbon, inorganic heat resistant fibers twisted into a fiber or a cord of multipore ceramic material; the installation is equipped with a heater which regulates the water temperature, in which the steam generator is submerged. The installation is designed for converting natural gas, C3H8, other hydrocarbon gases and vapors of liquid hydrocarbons (Uv) into H2 and CO. The design and disposition of the steam generator simplify the design of the device, eliminating the pump for feeding the steam and the device for premixing of the steam and hydrocarbon gas.

  11. Simulation of In-Core Dose Rates for an Offline CANDU Reactor

    Science.gov (United States)

    Gilbert, Jordan

    This thesis describes the development of a Monte Carlo simulation to predict the dose rates that will be encountered by a novel robotic inspection system for the pressure tubes of an offline CANDU reactor. Simulations were performed using the Monte Carlo N-Particle (MCNP) radiation transport code, version 6.1. The radiation fields within the reactor, even when shut down, are very high, and can cause significant damage to certain structural components and the electronics of the inspection system. Given that the robotic system will rely heavily on electronics, it is important to know the dose rates that will be encountered, in order to estimate the component lifetimes. The MCNP simulation was developed and benchmarked against information obtained from Ontario Power Generation and the Canadian Nuclear Laboratories. The benchmarking showed a good match between the simulated values and the expected values. This simulation, coupled with the accompanying user interface, represent a tool in dose field prediction that is currently unavailable. Predicted dose rates for a postulated inspection at 7 days after shutdown, with 2:5 cm of tungsten shielding around the key components, would survive for approximately 7 hours in core. This is anticipated to be enough time to perform an inspection and shows that the use of this tool can aid in designing the new inspection system.

  12. Current Status and Characterization of CANDU Spent Fuel for Geological Disposal System Design

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun; Lee, Seung Woo; Cha, Jeong Hun; Choi, Jong Won; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yang [SK Engineering and Construction, Seoul (Korea, Republic of)

    2008-06-15

    Inventories to be disposed of, reference turn up, and source terms for CANDU spent fuel were evaluated for geological disposal system design. The historical and projected inventory by 2040 is expected to be 14,600 MtU under the condition of 30-year lifetime for unit 1 and 40-year lifetime for other units in Wolsong site. As a result of statistical analysis for discharge burnup of the spent fuels generated by 2007, average and stand deviation revealed 6,987 MWD/MtU and 1,167, respectively. From this result, the reference burnup was determined as 8,100 MWD/MtU which covers 84% of spent fuels in total. Source terms such as nuclide concentration for a long-term safety analysis, decay heat, thermo-mechanical analysis, and radiation intensity and spectrum was characterized by using ORIGEN-ARP containing conservativeness in the aspect of decay heat up to several thousand years. The results from this study will be useful for the design of storage and disposal facilities.

  13. Adequacy of the SPARC9O code for the simulation of the capture of aerosols by break in steam generator; Adecuacion del codigo sparc90 para la simulacion de la captura de aerosoles por rotura en el generador de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Escriva, A.; Munoz-Cobo, J. L.; Berna, C.; Herranz, L. E.

    2011-07-01

    The SPARC9O (Suppression Pool Aerosol Removal Code)is intended to perform the calculation in terms of discharge of aerosols in swimming pools, is for this reason that it is working to achieve an analytical model that suits the conditions of discharge in a steam generator the code.

  14. Additional generation capacity or steam demand administration?; Capacidad adicional de generacion o administracion de demanda de vapor?

    Energy Technology Data Exchange (ETDEWEB)

    Plauchu Lima, Alberto; Plauchu Alcantara, Jorge Alberto [Plauchu Consultores, Morelia, Michoacan (Mexico)

    1999-07-01

    The authors have presented in seminars XVIII and XIX of ATPAE papers on subjects on Demand in Steam Systems and on the Supply of Consulting Services, respectively, and this paper is a contribution that retaking both subjects, tries to motivate the interest in generally underestimated areas of opportunity in energy programs. It is intention of this paper to invite the spokesmen to this Seminar and of the futures ones to deepen and to offer continuity and pursuit of an event to the nest one, on subjects of permanent interest and to incorporate contributions of new experiences, criteria and technologies. One exhorts to professionals and technicians interested in the rational and economic administration of the energy to deepen into the real conditions, abilities and capacities of the equipment in the steam power plants and to the study of better options of its use, adaptation and operation. In order to exemplify the previous matter two cases of industrial plants in which real and satisfactorily and economically solved problems, were very different from the ones that the proprietors of those facilities originally raised when soliciting the support of consulting services. Notice is made that in both cases capacity problems were contemplated, after the study and recommended actions it was demonstrated they handle their demands without augmenting the installed capacity, with avoided investments and increased efficiency. Finally the determinant is emphasized that can be the function of the consultant in the approach of his services and the direction towards orientation to right solutions. [Spanish] Los autores han expuesto en los seminarios XVIII y XIX de ATPAE trabajos sobre temas de Demanda en Sistemas de Vapor y Prestacion de Servicios de Consultoria, respectivamente y la presente es una contribucion que retomando ambos temas, pretende motivar el interes en areas de oportunidad generalmente subestimadas en programas de energia. Es intencion de este trabajo invitar a los

  15. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, M. K.; Lee, W. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicted with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applied for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented. 12 refs., 26 figs., 3 tabs. (Author)

  16. System of aid for the starting of the steam generator of a thermoelectric unit; Sistema de ayuda para el arranque del generador de vapor de una unidad termoelectrica

    Energy Technology Data Exchange (ETDEWEB)

    Quintero R, Agustin; Suarez C, Dionisio A; Aquino E, Juan C; Diaz H, Carlos A; Cruz T, Jorge A; Sanchez L, Jose A [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico)